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Sample records for floor tube cracking

  1. Fatigue cracking on a steam generator tube

    International Nuclear Information System (INIS)

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

    2015-01-01

    A circumferential fatigue crack was observed on a steam generator tube of the unit 2 of the Fessenheim plant. The results of destructive testing and the examination of the fracture surface show that the circumferential crack is linked to a large number of cycles with a very low stress intensity factor. Other aggravating factors like inter-granular corrosion have played a role in the initiating phase of fatigue cracking. The damage has been exacerbated by the lack of support of the tube at the level of the anti-vibration bars. (A.C.)

  2. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

    On April 28, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03, open-quote Circumferential Cracking of Steam Generator Tubes.close-quote GL 95-03 was issued to obtain information needed to verify licensee compliance with existing regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs). This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience. This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness

  3. Fracture of longitudinally cracked ductile tubes

    International Nuclear Information System (INIS)

    Larsson, H.; Bernard, J.

    1978-01-01

    Various bulging factor and plasticity correction factor formulations are discussed and a new plasticity correction factor leading to a simple failure law is proposed. Failure stresses predicted by the usual Linear Elastic Fracture Mechanics formula corrected for plasticity are shown to be identical with the Dowling and Townley two-criteria approach if the relevant parameters are chosen in a suitable manner. Burst tests on AISI 304 stainless steel tubes performed at the Joint Research Centre, Ispra are described. The strengthening effect of the sealing patch was taken into account by replacing the Folias bulging factor by a smaller empirical factor determined by Bernard and Henry from fatigue crack growth tests. A flow stress sigma and a toughness Ksub(c) were derived which apply to the prediction of the onset of stable crack growth in 304 stainless steel tubes at room temperature. For other ductile materials and temperatures tentative formulae are proposed. (author)

  4. The cracking of pressure tubes in the Pickering reactor

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.

    1978-01-01

    Small cracks in 17 of the 390 pressure tubes in Unit 3 of the 2056 MW (electrical) Pickering Generating Station and of 52 tubes in Unit 4, resulted in each of these units being out of service for many months. The cracks originated at areas of extremely high residual tensile stress produced by improper positioning of the rolling tool used during construction to join the pressure tube to its end-fitting. The mechanism of failure was delayed hydrogen cracking. (author)

  5. Delayed hydrogen cracking test design for pressure tubes

    International Nuclear Information System (INIS)

    Haddad, Roberto; Loberse, Antonio N.; Yawny, Alejandro A.; Riquelme, Pablo

    1999-01-01

    CANDU nuclear power stations pressure tubes of alloy Zr-2,5 % Nb present a cracking phenomenon known as delayed hydrogen cracking (DHC). This is a brittle fracture of zirconium hydrides that are developed by hydrogen due to aqueous corrosion on the metal surface. This hydrogen diffuses to the crack tip where brittle zirconium hydrides develops and promotes the crack propagation. A direct current potential decay (DCPD) technique has been developed to measure crack propagation rates on compact test (CT) samples machined from a non irradiated pressure tube. Those test samples were hydrogen charged by cathodic polarization in an acid solution and then pre cracked in a fatigue machine. This technique proved to be useful to measure crack propagation rates with at least 1% accuracy for DHC in pressure tubes. (author)

  6. Tomographic visualization of stress corrosion cracks in tubing

    International Nuclear Information System (INIS)

    Morris, R.A.; Kruger, R.P.; Wecksung, G.W.

    1979-06-01

    A feasibility study was conducted to determine the possibility of detecting and sizing cracks in reactor cooling water tubes using tomographic techniques. Due to time and financial constraints, only one tomographic reconstruction using the best technique available was made. The results indicate that tomographic reconstructions can, in fact, detect cracks in the tubing and might possibly be capable of measuring the depth of the cracks. Limits of detectability and sensitivity have not been determined but should be investigated in any future work

  7. Composite tube cracking in kraft recovery boilers: A state-of-the-art review

    Energy Technology Data Exchange (ETDEWEB)

    Singbeil, D.L.; Prescott, R. [Pulp and Paper Research Inst. of Canada, Vancouver, British Columbia (Canada); Keiser, J.R.; Swindeman, R.W. [Oak Ridge National Lab., TN (United States)

    1997-07-01

    Beginning in the mid-1960s, increasing energy costs in Finland and Sweden made energy recovery more critical to the cost-effective operation of a kraft pulp mill. Boiler designers responded to this need by raising the steam operating pressure, but almost immediately the wall tubes in these new boilers began to corrode rapidly. Test panels installed in the walls of the most severely corroding boiler identified austenitic stainless steel as sufficiently resistant to the new corrosive conditions, and discussions with Sandvik AB, a Swedish tube manufacturer, led to the suggestion that coextruded tubes be used for water wall service in kraft recovery boilers. Replacement of carbon steel by coextruded tubes has solved most of the corrosion problems experienced by carbon steel wall tubes, however, these tubes have not been problem-free. Beginning in early 1995, a multidisciplinary research program funded by the US Department of Energy was established to investigate the cause of cracking in coextruded tubes and to develop improved materials for use in water walls and floors of kraft recovery boilers. One portion of that program, a state-of-the-art review of public- and private-domain documents related to coextruded tube cracking in kraft recovery boilers is reported here. Sources of information that were consulted for this review include the following: tube manufacturers, boiler manufacturers, public-domain literature, companies operating kraft recovery boilers, consultants and failure analysis laboratories, and failure analyses conducted specifically for this project. Much of the information contained in this report involves cracking problems experienced in recovery boiler floors and those aspects of spout and air-port-opening cracking not readily attributable to thermal fatigue. 61 refs.

  8. Delayed hydrogen cracking of zirconium alloy pressure tubes

    International Nuclear Information System (INIS)

    Jackman, A.H.; Dunn, J.T.

    1976-10-01

    After several years of almost continuous service, Pickering Units 3 and 4 have both experienced long outages to replace cracked pressure tubes. This report summarizes the status of the investigation into the cause of the cracks as of May 1976. The basic cause of the cracking was the presence of very high residual tensile stresses in the pressure tubes due to improper rolling procedures. These residual stresses are being reduced to acceptable levels by local stress relieving techniques at Bruce G.S. and in future reactors improvements in rolling procedures and changes in pressure tube specifications will prevent a recurrence of this problem. (author)

  9. Plastic collapse behavior for thin tube with two parallel cracks

    International Nuclear Information System (INIS)

    Moon, Seong In; Chang, Yoon Suk; Kim, Young Jin; Lee, Jin Ho; Song, Myung Ho; Choi, Young Hwan; Kim, Joung Soo

    2004-01-01

    The current plugging criterion is known to be too conservative for some locations and types of defects. Many defects detected during in-service inspection take on the form of multiple cracks at the top of tube sheet but there is no reliable plugging criterion for the steam generator tubes with multiple cracks. Most of the previous studies on multiple cracks are confined to elastic analyses and only few studies have been done on the steam generator tubes failed by plastic collapse. Therefore, it is necessary to develop models which can be used to estimate the failure behavior of steam generator tubes with multiple cracks. The objective of this study is to verify the applicability of the optimum local failure prediction models proposed in the previous study. For this, plastic collapse tests are performed with the tube specimens containing two parallel through-wall cracks. The plastic collapse load of the steam generator tubes containing two parallel through-wall cracks are also estimated by using the proposed optimum global failure model and the applicability is investigated by comparing the estimated results with the experimental results. Also, the interaction effect between two cracks was evaluated to explain the plastic collapse behavior

  10. Time-dependent crack growth in steam generator tube leakage

    International Nuclear Information System (INIS)

    Chung, H.D.; Lee, J.H.; Park, Y.W.; Choi, Y.H.

    2006-01-01

    In general, cracks found in steam generator tubes have semi-elliptical shapes and it is assumed to be rectangular shape for conservatism after crack penetration. Hence, the leak and crack growth behavior has not been clearly understood after the elliptical crack penetrates the tube wall. Several experimental results performed by Argonne Nation Laboratory exhibited time-dependent crack growth behavior of rectangular flaws as well as trapezoidal flaws under constant pressure. The crack growth faster than expected was observed in both cases, which is likely attributed to time-dependent crack growth accompanied by fatigue sources such as the interaction between active jet and crack. The stress intensity factor, K 1 , is necessary for the prediction of the observed fatigue crack growth behavior. However, no K 1 solution is available for a trapezoidal flaw. The objective of this study is to develop the stress intensity factor which can be used for the fatigue analysis of a trapezoidal crack. To simplify the analysis, the crack is assumed to be a symmetric trapezoidal shape. A new K 1 formula for axial trapezoidal through-wall cracks was proposed based on the FEM results. (author)

  11. Crack resistance increasing in epoxide-rubber coatings of NPP room floors

    International Nuclear Information System (INIS)

    Khorenzhenko, V.I.

    1986-01-01

    Problems of crack resistance increasing in epoxide-rubber coatings for the floors are considered. Exploitation experience of the floors in the special rooms of NPP is given. Perspectivity of application of the compositions described as the building materials for nuclear power stations is pointed out

  12. 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.

  13. SQUIRT, Seepage in Reactor Tube Cracks

    International Nuclear Information System (INIS)

    Paul, D.; Ghadiali, N.; Wilkowski, G.; Rahman, S.; Krishnaswamy, P.

    1997-01-01

    1 - Description of program or function: The SQUIRT software is designed to perform leakage rate and area of crack opening calculations for through-wall cracks in pipes. The fluid in the piping system is assumed to be water at either subcooled or saturated conditions. The development of the SQUIRT computer model enables licensing authorities and industry users to conduct the leak-rate evaluations for leak-before-break applications in a more efficient manner. 2 - Method of solution: The SQUIRT program uses a modified form of the Henry-Fauske model for the thermal-hydraulics analysis together with Elastic-Plastic Fracture Mechanics using GE/EPRI and LBB.ENG2 methods for crack opening analysis. 3 - Restrictions on the complexity of the problem: Squirt requires 512 KB of conventional memory and an organized structure. Software can only be executed from the main SQUIRT23 directory where the software was installed

  14. 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.)

  15. Plugging criteria for steam generator tubes with axial cracks near tube support plates

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2000-01-01

    Stress corrosion cracking with intergranular attack occurs on the secondary side of steam generator (SG) tubes where impurities concentrate due to boiling under restricted flow conditions. In the most of cases, it can be called ODSCC (Outer Diameter Stress Corrosion Cracking). The typical locations are areas near support plates, in sludge piles and at top of tubesheet crevices. Though it can also occur on free spans under the relatively thin deposits that build up on the tube surfaces. ODSCC near tube plate supports have been the cause of plugging of many tubes. Thus, studies on SG tubes plugging criteria related to this degradation mechanism are presented in this paper. Th purpose is to avoid unnecessary tube plugging from either safety or reliability standpoint. Based on these studies some conclusions on the plugging criteria and on the difficulties to apply them are addressed. (author)

  16. 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

  17. 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

  18. 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

  19. Controlled chloride cracking of austenitic stainless steel tube samples

    OpenAIRE

    Raseroka, M.S.; Pistorius, P.C.

    2009-01-01

    An experimental rig has been constructed to produce chloride stress corrosion cracks in Type 304L stainless steel tube samples. The samples are to be used to test possible in situ repair methods in future work. The factor which influences the time to failure most strongly is the sample temperature; the distribution of cracks within the sample is affected by local temperature variations and by the position of the water line. Low-frequency oscillations in stress, caused by the on-off temperatur...

  20. Development of delayed hydride cracking resistant-pressure tube

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Kim, S. S.; Yim, K. S

    2000-10-01

    For the first time, we demonstrate that the pattern of nucleation and growth of a DHC crack is governed by the precipitation of hydrides so that the DHC velocity and K{sub IH} are determined by an angle of the cracking plane and the hydride habit plane 10.7. Since texture controls the distribution of the 10.7 habit plane in Zr-2.5Nb pressure tube, we draw a conclusion that a textural change in Zr-2.5Nb tube from a strong tangential texture to the radial texture shall increase the threshold stress intensity factor, K{sub IH}, and decrease the delayed hydride cracking velocity. This conclusion is also verified by a complimentary experiment showing a linear dependence of DHCV and K{sub IH} with an increase in the basal component in the cracking plane. On the basis of the study on the DHC mechanism and the effect of manufacturing processes on the properties of Zr-2.5Nb tube, we have established a manufacturing procedure to make pressure tubes with improved DHC resistance. The main features of the established manufacturing process consist in the two step-cold pilgering process and the intermediate heat treatment in the {alpha} + {beta} phase for Zr-2.5Nb alloy and in the {alpha} phase for Zr-1Nb-1.2Sn-0.4Fe alloy. The manufacturing of DHC resistant-pressure tubes of Zr-2.5Nb and Zr-1N-1.2Sn-0.4Fe was made in the ChMP zirconium plant in Russia under a joint research with Drs. Nikulina and Markelov in VNIINM (Russia). Zr-2.5Nb pressure tube made with the established manufacturing process has met all the specification requirements put by KAERI. Chracterization tests have been jointly conducted by VNIINM and KAERI. As expected, the Zr-2.5Nb tube made with the established procedure has improved DHC resistance compared to that of CANDU Zr-2.5Nb pressure tube used currently. The measured DHC velocity of the Zr-2.5Nb tube meets the target value (DHCV <5x10{sup -8} m/s) and its other properties also were equivalent to those of the CANDU Zr-2.5Nb tube used currently. The Zr-1Nb-1

  1. Iodine induced stress corrosion cracking of zircaloy cladding tubes

    International Nuclear Information System (INIS)

    Brunisholz, L.; Lemaignan, C.

    1984-01-01

    Iodine is considered as one of the major fission products responsible for PCI failure of Zry cladding by stress corrosion cracking (SCC). Usual analysis of SCC involves both initiation and growth as sequential processes. In order to analyse initiation and growth independently and to be able to apply the procedures of fracture mechanics to the design of cladding, with respect to SCC, stress corrosion tests of Zry cladding tubes were undertaken with a small fatigue crack (approx. 200 μm) induced in the inner wall of each tube before pressurization. Details are given on the techniques used to induce the fatigue crack, the pressurization test procedure and the results obtained on stress releaved or recrystallized Zry 4 tubings. It is shown that the Ksub(ISCC) values obtained during these experiments are in good agreement with those obtained from large DCB fracture mechanics samples. Conclusions will be drawn on the applicability of linear elastic fracture mechanics (LEFM) to cladding design and related safety analysis. The work now underway is aimed at obtaining better understanding of the initiation step. It includes the irradiation of Zry samples with heavy ions to simulate the effect of recoil fragments implanted in the inner surface of the cladding, that could create a brittle layer of about 10 μm

  2. Transport of lead to crack tips in steam generator tubes

    International Nuclear Information System (INIS)

    Adler, G.D.; Marks, C.R.; Fruzzetti, K.

    2009-01-01

    The mechanisms by which lead is transported from its ultimate source to steam generator tubes and into cracks are not well understood and, to date, a comprehensive evaluation of possible mechanisms has not previously been performed. Specifically, local lead concentrations up to 20 wt. percent have been measured at crack tips, and it is not fully understood how lead concentrations of this magnitude occur, since lead concentrations in SG feedwater are typically quite low (on the order of a few parts per trillion). Additionally, there is evidence that at secondary side conditions, lead is essentially entirely adsorbed onto solid surfaces. Furthermore, if lead were present in the liquid phase, it would not be expected to be in a form that would facilitate concentration in a crevice (crack) by electrochemical means. There has previously been some speculation that lead transport to crack tips may occur through surface diffusion of adsorbed species. It has also been postulated that lead transport may occur via diffusion through the oxide layer along crack walls or via diffusion of lead out of the bulk Alloy 600 to grain boundaries exposed to secondary water by advancing cracks. However, there have been no critical evaluations of these hypotheses. With the current state of knowledge, it is difficult for utilities to determine whether additional efforts to further reduce the inventory of lead in the secondary system are justified. Furthermore, specific sources of lead that are especially likely to accelerate SCC cannot be identified (e.g., significant masses of lead are present in SG deposits, but it is not known if this lead can be transported to crack tips). The work presented in this paper quantitatively evaluates (based on the published literature, not new experimental work) a number of hypothesized lead transport mechanisms, including: Liquid phase diffusion; Electrochemically influenced diffusion of cations and anions; Bulk alloy diffusion; Surface diffusion; Solid

  3. Thermal fatigue crack growth on a thick wall tube containing a semi elliptical circumferential crack

    International Nuclear Information System (INIS)

    Deschanels, H.; Wakai, T.; Lacire, M.H.; Michel, B.

    2001-01-01

    In order to check the ability of the simplified assessment procedure (A16 guide) to predict fatigue crack growth, a benchmark problem was conducted. This work is carried out under the project ''agreement on the Exchange of Information and Collaboration in the field of Research and Development of Fast Breeder Reactor (FBR) between Europe (EU) and Japan''. Experimental work is conducted by PNC using Air cooled Thermal transient Test Facility (ATTF). Specimen is a thick wall tube containing a semi elliptical (3-D) circumferential crack and subjected to cyclic thermal transients. The constitutive material is the 304 austenitic stainless steel type SUS304. Due to thermal shock (650 C-300 C) the stress distribution through the wall is non-linear and well approximated using a 3 rd order polynomial. When comparing computations and tests data we observe a good agreement for the crack propagation in length. In crack depth, accurate results are obtained in the first part of the test, but on the later stage of the experiment the computations slightly underestimate the propagation (deep crack). In addition, we notice the importance of good evaluation of fracture mechanics parameters for non-linear stress distribution through the wall. At present A16 guide handbook gives stress intensity factor solutions for non-linear stress distribution through the wall. (author)

  4. Effect of Crack Tip Stresses on Delayed Hydride Cracking in Zr-2.5Nb Tubes

    International Nuclear Information System (INIS)

    Kim, Young Suk; Cheong, Yong Moo

    2007-01-01

    Delayed hydride cracking (DHC) tests have shown that the DHC velocity becomes faster in zirconium alloys with a higher yield stress. To account for this yield stress effect on the DHC velocity, they suggested a simple hypothesis that increased crack tip stresses due to a higher yield stress would raise the difference in hydrogen concentration between the crack tip and the bulk region and accordingly the DHC velocity. This hypothesis is also applied to account for a big leap in the DHC velocity of zirconium alloys after neutron irradiation. It should be noted that this is based on the old DHC models that the driving force for DHC is the stress gradient. Puls predicted that an increase in the yield stress of a cold worked Zr-2.5Nb tube due to neutron irradiation by about 300 MPa causes an increase of its DHC velocity by an order of magnitude or 2 to 3 times depending on the accommodation energy values. Recently, we proposed a new DHC model that a driving force for DHC is not the stress gradient but the concentration gradient arising from the stress-induced precipitation of hydrides at the crack tip. Our new DHC model and the supporting experimental results have demonstrated that the DHC velocity is governed primarily by hydrogen diffusion at below 300 .deg. C. Since hydrogen diffusion in Zr-2.5Nb tubes is dictated primarily by the distribution of the β-phase, the DHC velocity of the irradiated Zr-2.5Nb tube must be determined mainly by the distribution of the β-phase, not by the increased yield stress, which is in contrast with the hypothesis of the previous DHC models. In short, a controversy exists as to the effect on the DHC velocity of zirconium alloys of a change in the crack tip stresses by irradiation hardening or cold working or annealing. The aim of this study is to resolve this controversy and furthermore to prove the validity of our DHC model. To this end, we cited Pan et al.'s experiment where the delayed hydride cracking velocity, the tensile strengths

  5. Structural integrity assessment of steam generator tubes deteriorated through primary water stress corrosion cracking in transition region of tube expansion

    International Nuclear Information System (INIS)

    Silveira, Helvecio Carlos Klinke da

    2002-01-01

    In PWR plants, steam generator tube degradation has been one of the most important economical concerns, besides causing operational safety problems. In this work, a survey of steam generator tube degradation modes is done. Degradation mechanisms and influence factors are introduced and discussed. The importance of stress corrosion cracking, especially in transition region of tube expansion zone, is underlined. The actual steam generator tube plugging criteria are conservative. Proposed alternative criteria are introduced and discussed. Distinction is done to structural integrity assessment of defective tubes. Real data of tube defect indications of axial cracks in expansion transition zone due to primary water stress corrosion cracking are used in analysis. Results allow discussing application aspects of deterministic and probabilistic criteria on structural integrity assessment of tubes with defect indications. Applied models are specifics, but the application of concept may be extended to other steam generator tube degradation modes. (author)

  6. Delayed hydride cracking in Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Mieza, Juan I.; Domizzi, Gladys; Vigna, Gustavo L.

    2007-01-01

    Zr-2.5 Nb alloy from CANDU pressure tubes are prone to failure by hydrogen intake. One of the degradation mechanisms is delayed hydride cracking, which is characterized by the velocity of cracking. In this work, we study the effect of beta zirconium phase transformation over delayed hydride cracking velocity in Zr-2.5 Nb alloy from pressure tubes. Acoustic emission technique was used for cracking detection. (author) [es

  7. An in-tube radar for detecting cracks in metal tubing

    International Nuclear Information System (INIS)

    Caffey, Thurlow W. H.; Nassersharif, Bahram; Garcia, Gabe V.; Smith, Phillip R.; Jedlicka, Russell P.; Hensel, Edward C.

    2000-01-01

    A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique will be described for detection of defects using a continuous-wave radar device within metal tubing. The technique is 100% volumetric, and may find smaller defects, find them more rapidly, and find them less expensively than present methods. Because this project was started only recently, there is no demonstrated performance to report so far. However, the basic engineering concepts will be presented together with a description of the milestone tasks and dates

  8. Stress corrosion cracking in superheater and reheater austenitic tubing

    Energy Technology Data Exchange (ETDEWEB)

    Dooley, R. Barry [Structural Integrity Associates, Inc., Charlotte, NC (United States); Bursik, Albert [PowerPlant Chemistry GmbH, Neulussheim (Germany)

    2011-02-15

    University 101 courses are typically designed to help incoming first-year undergraduate students to adjust to the university, develop a better understanding of the college environment, and acquire essential academic success skills. Why are we offering a special Boiler and HRSG Tube Failures PPChem 101? The answer is simple, yet very conclusive: - There is a lack of knowledge on the identification of tube failure mechanisms and for the implementation of adequate counteractions in many power plants, particularly at industrial power and steam generators. - There is a lack of knowledge to prevent repeat tube failures. The vast majority of BTF/HTF have been, and continue to be, repeat failures. It is hoped that the information about the failure mechanisms of BTF supplied in this course will help to put plant engineers and chemists on the right track. The major goal of this course is the avoidance of repeat BTF. This eights lesson is focused on Stress Corrosion Cracking in Superheater and Reheater Austenitic Tubing. (orig.)

  9. Crack resistance curves determination of tube cladding material

    Energy Technology Data Exchange (ETDEWEB)

    Bertsch, J. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)]. E-mail: johannes.bertsch@psi.ch; Hoffelner, W. [Paul Scherrer Institut, CH-5232 Villigen PSI (Switzerland)

    2006-06-30

    Zirconium based alloys have been in use as fuel cladding material in light water reactors since many years. As claddings change their mechanical properties during service, it is essential for the assessment of mechanical integrity to provide parameters for potential rupture behaviour. Usually, fracture mechanics parameters like the fracture toughness K {sub IC} or, for high plastic strains, the J-integral based elastic-plastic fracture toughness J {sub IC} are employed. In claddings with a very small wall thickness the determination of toughness needs the extension of the J-concept beyond limits of standards. In the paper a new method based on the traditional J approach is presented. Crack resistance curves (J-R curves) were created for unirradiated thin walled Zircaloy-4 and aluminium cladding tube pieces at room temperature using the single sample method. The procedure of creating sharp fatigue starter cracks with respect to optical recording was optimized. It is shown that the chosen test method is appropriate for the determination of complete J-R curves including the values J {sub 0.2} (J at 0.2 mm crack length), J {sub m} (J corresponding to the maximum load) and the slope of the curve.

  10. A new repair criterion for steam generator tubes with axial cracks based on probabilistic integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Su; Oh, Chang-Kyun [KEPCO Engineering and Construction Company, Inc., 269, Hyeoksin-ro, Gimcheon, Gyeongsangbuk-do 39660 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, College of Engineering, Kyung Hee University, 1732 Deokyoungdaero, Giheung, Yongin, Gyeonggi 446-701 (Korea, Republic of)

    2017-03-15

    Highlights: • Probabilistic assessment was performed for axially cracked steam generator tubes. • The threshold crack sizes were determined based on burst pressures of the tubes. • A new repair criterion was suggested as a function of operation time. - Abstract: Steam generator is one of the major components in a nuclear power plant, and it consists of thousands of thin-walled tubes. The operating record of the steam generators has indicated that a number of axial cracks due to stress corrosion have been frequently detected in the tubes. Since the tubes are closely related to the safety and also the efficiency of a nuclear power plant, an establishment of the appropriate repair criterion for the defected tubes and its applications are necessary. The objective of this paper is to develop an accurate repair criterion for the tubes with axial cracks. To do this, a thorough review is performed on the key parameters affecting the tube integrity, and then the probabilistic integrity assessment is carried out by considering the various uncertainties. In addition, the sizes of critical crack are determined by comparing the burst pressure of the cracked tube with the required performance criterion. Based on this result, the new repair criterion for the axially cracked tubes is defined from the reasonably conservative value such that the required performance criterion in terms of the burst pressure is able to be met during the next operating period.

  11. Crack detection in oak flooring lamellae using ultrasound-excited thermography

    Science.gov (United States)

    Pahlberg, Tobias; Thurley, Matthew; Popovic, Djordje; Hagman, Olle

    2018-01-01

    Today, a large number of people are manually grading and detecting defects in wooden lamellae in the parquet flooring industry. This paper investigates the possibility of using the ensemble methods random forests and boosting to automatically detect cracks using ultrasound-excited thermography and a variety of predictor variables. When friction occurs in thin cracks, they become warm and thus visible to a thermographic camera. Several image processing techniques have been used to suppress the noise and enhance probable cracks in the images. The most successful predictor variables captured the upper part of the heat distribution, such as the maximum temperature, kurtosis and percentile values 92-100 of the edge pixels. The texture in the images was captured by Completed Local Binary Pattern histograms and cracks were also segmented by background suppression and thresholding. The classification accuracy was significantly improved from previous research through added image processing, introduction of more predictors, and by using automated machine learning. The best ensemble methods reach an average classification accuracy of 0.8, which is very close to the authors' own manual attempt at separating the images (0.83).

  12. Evaluation of a leaking crack in an irradiated CANDU pressure tube

    International Nuclear Information System (INIS)

    Coleman, C.E.; Simpson, L.A.

    1988-06-01

    Leak-before-break is used in CANDU reactors as part of the defence against rupture of the pressure tubes. Two important features of this technique are the action time available for detection of a leaking crack and the size of the leak allowing crack location. Support for continued reliance on leak-before-break is being obtained from experiments, on irradiated Zr-2.5 Nb pressure tubes attached to their end fittings, that simulate the behaviour of a leaking crack in a reactor. At reactor operating temperatures leaking cracks grow more slowly than dry cracks in the laboratory because they are cooled when pressurised water flashes to steam on their surface. These cracks remain stable till they are at least 70 mm long. From the results of these experiments the action time is at least 100 h. The leak rate increases rapidly when a through-wall crack extends a small amount, thus greatly assisting with crack location

  13. 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

  14. 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.

  15. Evaluation of the inner wall axial cracks of steam generator tubes by eddy current test

    International Nuclear Information System (INIS)

    Hur, Do Haeng; Choi, Myung Sik; Lee, Doek Hyun; Han, Jung Ho

    2001-01-01

    For the enhancement of ECT reliability on the primary water stress corrosion cracks of nuclear steam generator tubes, it is important to comprehend the signal characteristics on crack morphology and to select an appropriate probe type. In this paper, the sizing accuracy and the detectability for the inner wall axial cracks of tubes were quantitatively evaluated using the electric discharge machined notches and the corrosion cracks which were developed on the operating steam generator tubes. The difference of eddy current signal characteristics between pancake and axial coil were also investigated. The results obtained from this study provide a useful information for more precise evaluation on the inner wall axial cracks of steam generator tubes.

  16. Development of Evaluation Technology for Detection of Axial Crack at Eggcrate Intersection of Steam Generator Tube

    International Nuclear Information System (INIS)

    Choi, Myung Sik; Hur, Do Haeng; Kim, Kyung Mo; Han, Jung Ho; Lee, Deok Hyun; Song, Myung Ho

    2011-01-01

    The occurrence of outer diameter (OD) axial stress corrosion crack at egg crate intersection of steam generator tube in operating power plant is inspected primarily by the eddy current test using bobbin coil probe. Therefore, the characteristics of the bobbin coil signal from the axial crack at egg crate intersection of steam generator tube should be understood for the accurate and earlier detection of the crack. In this study, the mockup assembly simulating the steam generator tube with OD axial stress corrosion crack and tube support egg crate was manufactured, and the characteristics of bobbin coil eddy current signal was examined in order to extract the improved evaluation technique for the detection of the crack

  17. Top of tubesheet cracking in Bruce A NGS steam generator tubing - recent experience

    International Nuclear Information System (INIS)

    Clark, M.A.; Lepik, O.; Mirzai, M.; Thompson, I.

    1998-01-01

    During the Bruce A Nuclear Generating Station (BNGS-A) Unit 1 1997 planned outage, a dew point search method identified a leak in one steam generator(SG) tube. Subsequently, the tube was inspected with all available eddy current probes and removed for examination. The initial inspection results and metallurgical examination of the removed tube confirmed that the leak was due to intergranular attack/stress corrosion cracking (IGA/SCC) emanating from the secondary side of the tube at the top of the tubesheet location. Subsequently, eddy current and ultrasonic indications were found at the top of the tubesheet of other Alloy 600 SG tubes. To investigate the source of the indications and to validate the inspection probes, sections of 40 tubes with various levels of damage were removed. The metallurgical examination of the removed sections showed that both secondary side and primary side initiated, circumferential, stress corrosion cracking and intergranular attack occurred in the BNGS-A SG tubing. Significant degradation from both mechanisms was found, invariably located in the roll transition region of the top expansion joint between the tube and the tubesheet on the hot leg (304 degrees C) side of the tube. Various aspects of the failures and tube examinations are presented in this paper, including presentation of the cracking morphology, measured crack size distributions, and discussion of some factors possibly affecting the cracking. (author)

  18. 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

  19. Prediction of crack coalescence of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Abou-Hanna, Jeries; McGreevy, Timothy E.; Majumdar, Saurin

    2004-01-01

    Prediction of failure pressures of cracked steam generator tubes of nuclear power plants is an important ingredient in scheduling inspection and repair of tubes. Prediction is usually based on nondestructive evaluation (NDE) of cracks. NDE often reveals two neighboring cracks. If the cracks interact, the tube pressure under which the ligament between the two cracks fails could be much lower than the critical burst pressure of an individual equivalent crack. The ability to accurately predict the ligament failure pressure, called ''coalescence pressure,'' is important. The failure criterion was established by nonlinear finite element model (FEM) analyses of coalescence of two 100% through-wall collinear cracks. The ligament failure is precipitated by local instability of the ligament under plane strain conditions. As a result of this local instability, the ligament thickness in the radial direction decreases abruptly with pressure. Good correlation of FEM analysis results with experimental data obtained at Argonne National Laboratory's Energy Technology Division demonstrated that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for 100% through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments

  20. Study of scratch-induced stress corrosion cracking for steam generator tubes and scratch control

    International Nuclear Information System (INIS)

    Meng, F.; Xu, X.; Liu, X.; Wang, J.

    2014-01-01

    This paper introduces field cases for scratch-induced stress corrosion cracking (SISCC) of steam generator tubes in PWR and current studies in laboratories. According to analysis result of broke tubes, scratches caused intergranular stress corrosion cracking (IGSCC) with outburst. The effect of microstructure for nickel-base alloys, residual stresses caused by scratching process and water chemistry on SISCC and possible mechanism of SISCC are discussed. The result shows that scratch-induced microstructure evolution contributes to SISCC significantly. The causes of scratches during steam generator tubing manufacturing and installation process are stated and improved reliability with scratch control is highlighted for steam generator tubes in newly built nuclear power plants. (author)

  1. Study of scratch-induced stress corrosion cracking for steam generator tubes and scratch control

    Energy Technology Data Exchange (ETDEWEB)

    Meng, F.; Xu, X.; Liu, X. [Shanghai Nuclear Engineering Research and Design Institute, Shanghai (China); Wang, J. [Chinese Academy of Sciences, Institute of Metal Research, Shenyang (China)

    2014-07-01

    This paper introduces field cases for scratch-induced stress corrosion cracking (SISCC) of steam generator tubes in PWR and current studies in laboratories. According to analysis result of broke tubes, scratches caused intergranular stress corrosion cracking (IGSCC) with outburst. The effect of microstructure for nickel-base alloys, residual stresses caused by scratching process and water chemistry on SISCC and possible mechanism of SISCC are discussed. The result shows that scratch-induced microstructure evolution contributes to SISCC significantly. The causes of scratches during steam generator tubing manufacturing and installation process are stated and improved reliability with scratch control is highlighted for steam generator tubes in newly built nuclear power plants. (author)

  2. Evaluation of Eddy Current Signals from the Inner Wall Axial Cracks of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Choi, Myung Sik; Hur, Do Haeng; Lee, Doek Hyun; Han, Jung Ho; Park, Jung Am

    2001-01-01

    For the enhancement of ECT reliability on the primary water stress corrosion cracks of nuclear steam generator tubes, of which the occurrence is on the increase, it is important to comprehend the signal characteristics on crack morphology and to select an appropriate probe type. In this paper, the sizing accuracy and the detectability for the inner wall axial cracks of tubes were quantitatively evaluated using the following specimens: the electric discharge machined notches and the corrosion cracks which were developed on the operating steam generator tubes. The difference of eddy current signal characteristics between pancake and axial coil were also Investigated. The results obtained from this study provide a useful information for more precise evaluation on the inner wall axial tracks oi steam generator tubes

  3. Some engineering aspects of the investigation into the cracking of pressure tubes in the Pickering reactors

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Towgood, G.R.; Hunter, T.A.

    1976-01-01

    In August 1974, Pickering Unit 3 (514 MWe) was shutdown for a period of 8 months because of cracks in 17 of the 390 pressure tubes. The cracks were a result of incorrect installation procedures during construction. Improper positioning of the rolling tool used to join the Zr-2.5 wt% Nb pressure tube to the end fitting produced very high residual tensile stresses. High stresses in combination with periods with the tubes cold caused the cracking. Crack propagation was by fracture of hydrides which are brittle when cold. Subsequent investigation confirmed that properly rolled joints are not susceptible to such cracking. The resources of Canadian industry, Ontario Hydro and Atomic Energy of Canada were coordinated to find engineering solutions to the crack program. The defective tubes were removed from reactor, thoroughly examined to identify the cause of the cracks, and thoroughly tested to prove safety. Non-destructive techniques were quickly adopted for inspection of tubes in Pickering. Tools and procedures for retubing the 17 channels were prepared and Pickering Unit 3 was returned to service at the end of March 1975. (author)

  4. 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)

  5. Method of evaluation of stress corrosion cracking susceptibility of clad fuel tubes

    International Nuclear Information System (INIS)

    Takase, Iwao; Yoshida, Toshimi; Ikeda, Shinzo; Masaoka, Isao; Nakajima, Junjiro.

    1986-01-01

    Purpose: To determine, by an evaluation in out-pile test, the stress corrosion cracking susceptibility of clad fuel tubes in the reactor environment. Method: A plurality of electrodes are mounted in the circumferential direction on the entire surface of cladding tubes. Of the electrodes, electrodes at two adjacent places are used as measuring terminals and electrodes at another two places adjacent thereto are used as constant-current terminals. With a specific current flowing in the constant-current terminals, measurements are made of a potential difference between the terminals to be measured, and from a variation in the potential difference the depth of cracking of the cladding tube surface is presumed to determine the stress corrosion cracking susceptibility of the cladding tube. To check the entire surface of the cladding tube, the cladding tube is moved by each block in the circumferential direction by a contact changeover system, repeating the measurements of the potential difference. Contact type electrodes are secured with an insulator and held in uniform contact with the cladding tube by a spring. It is detachable by use of a locking system and movable as desired. Thus the stress corrosion cracking susceptibility can be determined without mounting the cladding tube through and also a fuel failure can be prevented. (Horiuchi, T.)

  6. Influence of flow stress choice on the plastic collapse estimation of axially cracked steam generator tubes

    International Nuclear Information System (INIS)

    Tonkovic, Zdenko; Skozrit, Ivica; Alfirevic, Ivo

    2008-01-01

    The influence of the choice of flow stress on the plastic collapse estimation of axially cracked steam generator (SG) tubes is considered. The plastic limit and collapse loads of thick-walled tubes with external axial semi-elliptical surface cracks are investigated by three-dimensional non-linear finite element (FE) analyses. The limit pressure solution as a function of the crack depth, length and tube geometry has been developed on the basis of extensive FE limit load analyses employing the elastic-perfectly plastic material behaviour and small strain theory. Unlike the existing solutions, the newly developed analytical approximation of the plastic limit pressure for thick-walled tubes is applicable to a wide range of crack dimensions. Further, the plastic collapse analysis with a real strain-hardening material model and a large deformation theory is performed and an analytical approximation for the estimation of the flow stress is proposed. Numerical results show that the flow stress, defined by some failure assessment diagram (FAD) methods, depends not only on the tube material, but also on the crack geometry. It is shown that the plastic collapse pressure results, in the case of deeper cracks obtained by using the flow stress as the average of the yield stress and the ultimate tensile strength, can become unsafe

  7. Coalescence model of two collinear cracks existing in steam generator tubes

    International Nuclear Information System (INIS)

    Moon, S.-I.; Chang, Y.-S.; Kim, Y.-J.; Park, Y.-W.; Song, M.-H.; Choi, Y.-H.; Lee, J.-H.

    2005-01-01

    The 40% of wall thickness criterion has been used as a plugging rule of steam generator tubes but it can be applicable just to a single-cracked tubes. In the previous studies preformed by the authors, a total of 10 local failure prediction models were introduced to estimate the coalescence load of two adjacent collinear through-wall cracks existing in thin plates, and the reaction force model and plastic zone contact model were selected as optimum models among them. The objective of this study is to verify the applicability of the proposed optimum local failure prediction models to the tubes with two collinear through-wall cracks. For this, a series of plastic collapse tests and finite element analyses were carried out using the tubes containing two collinear through-wall cracks. It has been shown that the proposed optimum failure models can predict the local failure behavior of two collinear through-wall cracks existing in tubes well. And a coalescence evaluation diagram was developed which can be used to determine whether the adjacent cracks detected by NED coalsece or not. (authors)

  8. 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.)

  9. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    Energy Technology Data Exchange (ETDEWEB)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H. [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH{sub T} was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle.

  10. Effects of Chemistry Parameters of Primary Water affecting Leakage of Steam Generator Tube Cracks

    International Nuclear Information System (INIS)

    Shin, D. M.; Cho, N. C.; Kang, Y. S.; Lee, K. H.

    2016-01-01

    Degradation of steam generator (SG) tubes can affect pressure boundary tightness. As a defense-in-depth measure, primary to secondary leak monitoring program for steam generators is implemented, and operation is allowed under leakage limits in nuclear power plants. Chemistry parameters that affect steam generator tube leakage due to primary water stress corrosion cracking (PWSCC) are investigated in this study. Tube sleeves were installed to inhibit leakage and improve tube integrity as a part of maintenance methods. Steam generators occurred small leak during operation have been replaced with new steam generators according to plant maintenance strategies. The correlations between steam generator leakage and chemistry parameters are presented. Effects of primary water chemistry parameters on leakage from tube cracks were investigated for the steam generators experiencing small leak. Unit A experienced small leakage from steam generator tubes in the end of operation cycle. It was concluded that increased solubility of oxides due to high pHT could make leakage paths, and low boron concentration lead to less blockage in cracks. Increased dissolved hydrogen may retard crack propagations, but it did not reduce leak rate of the leaking steam generator. In order to inhibit and reduce leakage, pH_T was controlled by servicing cation bed operation. The test results of decreasing pHT indicate low pHT can reduce leak rate of PWSCC cracks in the end of cycle

  11. Measurements of delayed hydride cracking propagation rate in the radial direction of Zircaloy-2 cladding tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan); Uchikoshi, H. [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer The delayed hydride cracking (DHC) velocity of Zircaloy-2 was measured. Black-Right-Pointing-Pointer The velocity followed the Arrhenius law up to 270 Degree-Sign C. Activation energy was 49 kJ/mol. Black-Right-Pointing-Pointer The threshold stress intensity factor for the DHC was from 4 to 6 MPa m{sup 1/2}. Black-Right-Pointing-Pointer An increase in material strength accelerated the DHC. Black-Right-Pointing-Pointer Precipitation and fracture of hydrides at a crack tip is responsible for the DHC. - Abstract: Delayed hydride cracking (DHC) tests of Zircaloy-2 cladding tubes were performed in the chamber of a scanning electron microscope (SEM) to directly observe the crack propagation and measure the crack velocity in the radial direction of the tubes. Pre-cracks were produced at the outer surfaces of the tubes. Hydrogen contents of the tubes were from 90 ppm to 130 ppm and test temperatures were from 225 Degree-Sign C to 300 Degree-Sign C. The crack velocity followed the Arrhenius law at temperatures lower than about 270 Degree-Sign C with apparent activation energy of about 49 kJ/mol. The upper temperature limit for DHC, above which DHC did not occur, was about 280 Degree-Sign C. The threshold stress intensity factor for the initiation of the crack propagation, K{sub IH}, was from about 4 MPa m{sup 1/2} to 6 MPa m{sup 1/2}, almost independent of temperature. An increase in 0.2% offset yield stress of the material accelerated the crack velocity and slightly decreased K{sub IH}. Detailed observations of crack tip movement showed that cracks propagated in an intermittent fashion and the propagation gradually approached the steady state as the crack depth increased. The SEM observations also showed that hydrides were formed at a crack tip and a number of micro-cracks were found in the hydrides. It was presumed from these observations that the repetition of precipitation and fracture of hydrides at the crack tip would be

  12. Depth-Sizing Technique for Crack Indications in Steam Generator Tubing

    International Nuclear Information System (INIS)

    Cho, Chan Hee; Lee, Hee Jeong; Kim, Hong Deok

    2009-01-01

    The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program

  13. Evaluation of crack propagation of alloy 600 tube in high temperature water, (1)

    International Nuclear Information System (INIS)

    Hirano, Hideo; Kawamura, H.; Kawamura, Kohji; Matsubara, Masaaki

    1990-01-01

    This report describes the analysis of stress intensity factors at cracks in alloy 600 steam generator tubes. Based on the results of the analysis, IGA/SCC tests were carried out to examine the effect of stress intensity and water quality on the crack propagation rate. The main test result are as follows: (1) Hoop stress was caused by the pressure difference between the internal and external surface of the steam generator tube. The calculated hoop stress was about 7 kg/mm 2 . In addition, the temperature difference between the internal and external surface caused thermal stress. The thermal stress was about 10 kg/mm 2 at the external surface and the one at the internal surface was about -10 kg/mm 2 . Total stress at the external and internal surface was 17 kg/mm 2 and -3 kg/mm 2 , respectively. (2) The stress intensity factor at the crack tip increased with increasing crack length. For a long crack, the stress intensity factor decreased with increasing crack number. However, for a short crack, the stress intensity factor decreased little with increasing crack number. (3) Under high stress-intensity conditions, i.e. 40∼50 kg·mm -3/2 , the IGA/SCC test showed that IGA/SCC propagated in AVT and AVT/boric-acid solution at 320degC and 350degC. However, the propagation rate was low. (author)

  14. Effect of layerwise structural inhomogeneity on stress- corrosion cracking of steel tubes

    Science.gov (United States)

    Perlovich, Yu A.; Krymskaya, O. A.; Isaenkova, M. G.; Morozov, N. S.; Fesenko, V. A.; Ryakhovskikh, I. V.; Esiev, T. S.

    2016-04-01

    Based on X-ray texture and structure analysis data of the material of main gas pipelines it was shown that the layerwise inhomogeneity of tubes is formed during their manufacturing. The degree of this inhomogeneity affects on the tendency of tubes to stress- corrosion cracking under exploitation. Samples of tubes were cut out from gas pipelines located under various operating conditions. Herewith the study was conducted both for sections with detected stress-corrosion defects and without them. Distributions along tube wall thickness for lattice parameters and half-width of X-ray lines were constructed. Crystallographic texture analysis of external and internal tube layers was also carried out. Obtained data testifies about considerable layerwise inhomogeneity of all samples. Despite the different nature of the texture inhomogeneity of gas pipeline tubes, the more inhomogeneous distribution of texture or structure features causes the increasing of resistance to stress- corrosion. The observed effect can be explained by saturation with interstitial impurities of the surface layer of the hot-rolled sheet and obtained therefrom tube. This results in rising of lattice parameters in the external layer of tube as compared to those in underlying metal. Thus, internal layers have a compressive effect on external layers in the rolling plane that prevents cracks opening at the tube surface. Moreover, the high mutual misorientation of grains within external and internal layers of tube results in the necessity to change the moving crack plane, so that the crack growth can be inhibited when reaching the layer with a modified texture.

  15. Stress corrosion cracking susceptibility of steam generator tubing on secondary side in restricted flow areas

    International Nuclear Information System (INIS)

    Fulger, M.; Lucan, D.; Radulescu, M.; Velciu, L.

    2003-01-01

    Nuclear steam generator tubes operate in high temperature water and on the secondary side in restricted flow areas many nonvolatile impurities accidentally introduced into circuit tend to concentrate. The concentration process leads to the formation of highly aggressive alkaline or acid solutions in crevices, and these solutions can cause stress corrosion cracking (SCC) on stressed tube materials. Even though alloy 800 has shown to be highly resistant to general corrosion in high temperature water, it has been found that the steam generator tubes may crack during service from the primary and/or secondary side. Stress corrosion cracking is still a serious problem occurring on outside tubes in operating steam generators. The purpose of this study was to evaluate the environmental factors affecting the stress corrosion cracking of steam generators tubing. The main test method was the exposure for 1000 hours into static autoclaves of plastically stressed C-rings of Incoloy 800 in caustic solutions (10% NaOH) and acidic chloride solutions because such environments may sometimes form accidentally in crevices on secondary side of tubes. Because the kinetics of corrosion of metals is indicated by anodic polarization curves, in this study, some stressed specimens were anodically polarized in caustic solutions in electrochemical cell, and other in chloride acidic solutions. The results presented as micrographs, potentiokinetic curves, and electrochemical parameters have been compared to establish the SCC behavior of Incoloy 800 in such concentrated environments. (authors)

  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. Delayed hydride cracking velocity and crack growth measurement using DCPD technique in Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Singh, R.N.; Kishore, R.; Roychaudhury, S.; Unnikrishnan, M.; Sinha, T.K.; De, P.K.; Banerjee, S.; Kumar, Santosh

    2000-12-01

    Nuclear structural materials have to perform under most demanding and exotic environmental conditions. Due to its unique properties dilute zirconium alloys are the only choice for in-core structural materials in water cooled nuclear reactors. Hydrogen related problems have been recognized as the life-limiting factor for the core components of Pressurized Heavy Water Reactors (PHWR). Delayed Hydride Cracking (Dhc) is one of them. In this study, Dhc crack growth has been monitored using Direct Current Potential Drop (Dcp) technique. Calibration curve between normalized Dcp output and normalized crack length was established at different test temperatures. Dhc velocity was measured along the axial direction of the Zirconium-2.5Niobium pressure tube material at 203 and 250 degree C. (author)

  18. Replacement of a cracked pressure tube in Bruce GS unit 2

    International Nuclear Information System (INIS)

    Dunn, J.T.

    1982-06-01

    In 1982 February, a primary heat transport system leak was detected in the annulus gas system by on-line instrumentation. The source of the leak was found to be a small axial crack in the pressure tube of fuel channel X-14. This fuel channel was removed and replaced by station maintenance staff, and the unit was returned to service five weeks after it had been shut down. The cracked pressure tube was sent to Chalk River Nuclear Laboratories for examination, and the crack was found to be very similar to those found in Pickering GS units 3 and 4 in 1974-75. It was caused by delayed hydride cracking during the period of high residual stress between the time of rolling and the pre-service stress relief

  19. Observations and insights into Pb-assisted stress corrosion cracking of alloy 600 steam generator tubes

    International Nuclear Information System (INIS)

    Thomas, L.; Bruemmer, Stephen M.

    2005-01-01

    Pb-assisted stress-corrosion cracking (PbSCC) of Alloy 600 steam-generator tubing in high-temperature-water service and laboratory tests were studied by analytical transmission electron microscopy of cross-sectioned samples. Examinations of pulled tubes from many pressurized water reactors revealed lead in cracks from 11 of 17 samples. Comparisons of the degraded intergranular structures with ones produced in simple laboratory tests with PbO in near-neutral AVT water showed that the PbSCC characteristics in service tubing could be reproduced without complex chemistries and heat-flow conditions that can occur during plant operation. Observations of intergranular and transgranular cracks promoted by Pb in the test samples also provided new insights into the mechanisms of PbSCC in mill-annealed and thermally treated Alloy 600

  20. A fracture mechanics model for iodine stress corrosion crack propagation in Zircaloy tubing

    International Nuclear Information System (INIS)

    Crescimanno, P.J.; Campbell, W.R.; Goldberg, I.

    1984-01-01

    A fracture mechanics model is presented for iodine-induced stress corrosion cracking in Zircaloy tubing. The model utilizes a power law to relate crack extension velocity to stress intensity factor, a hyperbolic tangent function for the influence of iodine concentration, and an exponential function for the influence of temperature and material strength. Comparisons of predicted to measured failure times show that predicted times are within a factor of two of the measured times for a majority of the specimens considered

  1. Optimization of crack detection in steam generator tubes using a punctual probe

    International Nuclear Information System (INIS)

    Levy, R.; Ferre, C.

    1985-01-01

    The existence of cracks at the upper end of the expanded zone of a steam generator tube is a recent problem. A differential pencil probe was used for the detection of those cracks with encouraging results. An optimization study has been necessary to solve the difficulties in the evaluation of defects, due to the design of the first probe; the result is a probe making possible a precise analysis of detected signals

  2. Workshop proceedings: U-bend tube cracking in steam generators

    Science.gov (United States)

    Shoemaker, C. E.

    1981-06-01

    A design to reduce the rate of tube failure in high pressure feedwater heaters, a number of failed drawn and stress relieved Monel 400 U-bend tubes removed from three high pressure feedwater heaters was examined. Steam extracted from the turbine is used to preheat the boiler feedwater in fossil fuel fired steam plants to improve thermal efficiency. This is accomplished in a series of heaters between the condenser hot well and the boiler. The heaters closest to the boiler handle water at high pressure and temperature. Because of the severe service conditions, high pressure feedwater heaters are frequently tubed with drawn and stress relieved Monel 400.

  3. Stress corrosion cracking of the tubing materials for nuclear steam generators in an environment containing lead

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, Uh Chul; Lee, Eun Hee; Hwang, Seong Sik

    2004-01-01

    Steam generator tube materials show a high susceptibility to stress corrosion cracking (SCC) in an environment containing lead species and some nuclear power plants currently have degradation problems associated with lead-induced stress corrosion cracking in a caustic solution. Effects of an applied potential on SCC is tested for middle-annealed Alloy 600 specimens since their corrosion potential can be changed when lead oxide coexists with other oxidizing species like copper oxide in the sludge. In addition, all the steam generator tubing materials used for nuclear power plants being operated and currently under construction in Korea are tested in a caustic solution with lead oxide. (author)

  4. Crack growth of throughwall flaw in Alloy 600 tube during leak testing

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Majumdar, Saurin

    2015-01-01

    Graphical abstract: - Highlights: • A series of leak testing was conducted at a constant pressure and room temperature. • The time-dependent increase in the leak rate was observed. • The fractography revealed slip offsets and crystallographic facets. • Time-dependent plasticity at the crack tip caused the slip offsets. • Fatigue by jet/structure interaction caused the crystallographic facets. - Abstract: We examined the issue of whether crack growth in a full thickness material can occur in a leaking crack. A series of leak tests was conducted at a room temperature and constant pressure (17.3 MPa) with Alloy 600 tube specimens containing a tight rectangular throughwall axial fatigue crack. To exclude a potential pulsation effect by a high pressure pump, the test water was pressurized by using high pressure nitrogen gas. Fractography showed that crack growth in the full thickness material can occur in the leaking crack by two mechanisms: time-dependent plasticity at the crack tip and fatigue induced by jet/structure interaction. The threshold leak rate at which the jet/structure interaction was triggered was between 1.3 and 3.3 L/min for the specific heat of the Alloy 600 tube tested

  5. Delayed hydride cracking in irradiated Zr-2.5 % Nb pressure tubes

    International Nuclear Information System (INIS)

    Cirimello, Pablo; Coronel, Pascual; Haddad, Roberto; Lafont, Claudio; Mizrahi, Rafael

    2003-01-01

    Pressure tubes in CANDU nuclear power plants are made of Zr-2.5 % Nb alloy, which is susceptible to a cracking process called Delayed Hydride Cracking (DHC). Measurement of DHC velocity on irradiated pressure tubes is essential to assure the validity of the Leak Before Break criterion. This work was performed on samples from two pressure tubes taken out of the Embalse NPP in 1995, belonging to fuel channels A-14 and L-12. DHC velocity in the axial direction was measured at 211 C degrees for samples taken from different axial positions, which allowed to study its dependence on fast neutron fluency and irradiation temperature. Non-irradiated material was also tested. It was found that DHC velocity results for the tested material were similar to those obtained for a great number of tubes irradiated in other CANDU plants. (author)

  6. Delayed hydride cracking and elastic properties of Excel, a candidate CANDU-SCWR pressure tube material

    International Nuclear Information System (INIS)

    Pan, Z.L.

    2010-01-01

    Excel, a Zr alloy which contains 3.5%Sn, 0.8%Nb and 0.8%Mo, shows high strength, good corrosion resistance, excellent creep-resistance and dimension stability and thus is selected as a candidate pressure tube material for CANDU-SCWR. In the present work, the delayed hydride cracking properties (K IH and the DHC growth rates), the hydrogen solubility and elastic modulus were measured in the irradiated and unirradiated Excel pressure tube material. (author)

  7. Observations on the influence of tube manufacturing technique on iodine stress corrosion cracking of unirradiated Zircaloy

    International Nuclear Information System (INIS)

    Syrett, B.C.; Cubicciotti, D.; Jones, R.L.

    1979-01-01

    Closed-end tube pressurization tests at 593 K were used to compare the susceptibilities to iodine stress corrosion cracking (SCC) of two lots of Zircaloy-2 tubing manufactured by different suppliers. Although both tubings were produced to exactly the same specifications in terms of dimensions, chemical composition, burst strength, and certain other properties, as-received specimens from the two lots exhibited markedly different behavior in iodine SCC tests. The tubing from one supplier had a lower SCC threshold stress and failed about 30 times more quickly than the tubing from the other supplier. However, it was found that this difference in SCC susceptibility was eliminated if the internal surfaces of the specimens were polished to a 3 μm finish prior to testing. These observations are discussed in terms of possible effects of surface or near-surface chacteristics of the tubing on SCC susceptibility

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  9. Failure assessment and evaluation of critical crack length for a fresh Zr-2 pressure tube of an Indian PHWR

    International Nuclear Information System (INIS)

    Krishnan, Suresh; Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1996-01-01

    Fracture analysis of Zr-2 pressure tubes having through wall axial crack was done using finite element method. The analysis was done for tubes in as received condition. During reactor operation the mechanical properties of Zr-2 undergo changes. The analysis is valid for pressure tubes of newly commissioned reactors. The main aim of the study was to determine critical crack length of pressure tubes in normal operating conditions. Elastic plastic fracture analysis was done for different crack lengths to determine applied J-integral values. Tearing modulus instability concept was used to evaluate critical crack length. One of the important parameter studied was, the effect of crack face pressure, which leaking fluid exert on the crack faces/lips of through wall axial crack. Its effect was found to be significant for pressure tubes. It increases the applied J-integral values. Approximate analytical solutions which takes into account the plasticity ahead of crack tip, are available and widely used. These formulae do not take into account the crack face pressure. Since, for the present situation the effect of crack face pressure is significant hence, detailed finite analysis was necessary. Detailed 3D finite element analysis gives an insight into the variation of J-integral values over the thickness of pressure tube. It was found that J values are maximum at the middle layer of the tube. A peak factor on J values was defined and evaluated as ratio of maximum J to average J across the thickness, crack opening area for each length was also evaluated. The knowledge of crack opening area is useful for leak before break studies. The failure assessment was also done using Central Electricity Generating Board (CEGB) R-6 method considering the ductile tearing. The reserve factors (or safety margins) for different crack lengths was evaluated using R-6 method. (author). 30 refs., 21 figs., 34 tabs

  10. Delayed hydride cracking in Zr-2.5% wt Nb pressure tubes

    International Nuclear Information System (INIS)

    Cirimello, Pablo; Haddad, Roberto; Domizzi, Gladys

    2003-01-01

    During service, pressure tubes of CANDU nuclear power reactor are prone to suffer crack growth by delayed hydride cracking (DHC). For a given H 2 plus D 2 concentration there is a critical temperature (T c ) below which DHC may occur. In this work, T c was measured for CCT specimens cut from Zr-2.5 Wt % Nb pressure tubes. Hydrogen was added to the specimens to get concentrations of 40, 59 and 72 ppm. It was found that T c is higher than the corresponding precipitation temperature. The axial crack velocity (V p ) was also measured. Decreasing temperature from T c makes V p increase until a maximum is attained at a temperature close to precipitation temperature. At lower temperatures, in the presence of precipitated hydrides, decreasing temperature implies lower velocities, following an Arrhenius law: Vp=Aexp(-Q/RT), with an activation energy Q= 66 KJ/mol K. (author)

  11. The influence of lead on stress corrosion cracking of steam generator tubing

    International Nuclear Information System (INIS)

    Ryan Curtis Wolfe

    2015-01-01

    Lead (Pb) is present at low concentrations on the secondary side of steam generators, but is known to accumulate in steam generator sludge and become concentrated in crevices and cracks. Pb is known to have played a role in the degradation of Alloy 600MA tubing, necessitating the replacement of those steam generators. There is new evidence which indicates that Pb has also played a role in the stress corrosion cracking (SCC) of Alloy 600TT. Furthermore. laboratory testing indicates that advanced tubing alloys such as Alloy 690TT and Alloy 800NG area also susceptible to this attack. In response to these vulnerabilities, utilities are attempting to manufacture tubing using processes which will impart optimal corrosion resistance, fabricate and operate SG's to minimize stress in the tubing, undertake efforts to identify and remove the sources of Pb, reduce the existing inventory of Pb using chemical or mechanical cleaning processes, and maintain rigorous chemistry controls. Research is warranted to qualify chemical methods to mitigate PbSCC that may be observed in service. This presentation will review work performed through the Electric Power Research Institute (EPRI) to address the issue of Pb-assisted stress corrosion cracking of steam generator tubing. (author)

  12. Application of eddy currents for identification of dimensional variations in PWR steam generator tubes and detection of stress corrosion cracks

    International Nuclear Information System (INIS)

    Comby, R.; Gourmelon, A.

    1985-01-01

    To avoid the risk of cracking on the secondary side of the roll expansion transition zone in steam generator (SG) tubes, tube profile at the upper face of the tube sheet must comply with specifications laid down by the manufacturer and EDF. EDF has developed an eddy current (EC) signal identification method, used for pre-service testing to detect any deviation in tube profile. Nevertheless, circumferential or longitudinal stress corrosion cracks (SCC), initiated on the primary side, have appeared on some SGs. A special rotating probe was used on these generators. The results of these checks have been correlated with metallurgical examination of the extracted tubes

  13. Enhancement of leak rate estimation model for corroded cracked thin tubes

    International Nuclear Information System (INIS)

    Chang, Y.S.; Jeong, J.U.; Kim, Y.J.; Hwang, S.S.; Kim, H.P.

    2010-01-01

    During the last couple of decades, lots of researches on structural integrity assessment and leak rate estimation have been carried out to prevent unanticipated catastrophic failures of pressure retaining nuclear components. However, from the standpoint of leakage integrity, there are still some arguments for predicting the leak rate of cracked components due primarily to uncertainties attached to various parameters in flow models. The purpose of present work is to suggest a leak rate estimation method for thin tubes with artificial cracks. In this context, 23 leak rate tests are carried out for laboratory generated stress corrosion cracked tube specimens subjected to internal pressure. Engineering equations to calculate crack opening displacements are developed from detailed three-dimensional elastic-plastic finite element analyses and then a simplified practical model is proposed based on the equations as well as test data. Verification of the proposed method is done through comparing leak rates and it will enable more reliable design and/or operation of thin tubes.

  14. The manufacturing of Stress Corrosion Crack (SCC) on Inconel 600 tube

    International Nuclear Information System (INIS)

    Bae, Seunggi; Bak, Jaewoong; Kim, Seongcheol; Lee, Sangyul; Lee, Boyoung

    2014-01-01

    The Stress Corrosion Crack (SCC), taken a center stage in recently accidents about nuclear power plants, is one of the environmentally induced cracking occurred when a metallic structure under tensile stress is exposed to corrosive environment. In this study, the SCC was manufactured in the simulated corrosive environmental conditions on Inconel 600 tube that widely applied in the nuclear power plants. The tensile stress which is one of the main factors to induce SCC was given by GTAW welding in the inner surface of the specimen. The corrosive environment was simulated by using the sodium hydroxide (NaOH) and sodium sulfide (Na 2 S). In this study, SCC was manufactured in the simulated corrosive environmental conditions with Inconel 600 tube that widely applied in the nuclear power plants. 1) The SCC was manufactured on Inconel 600 tube in simulated operational environments of nuclear power plants. In the experiment, the welding heat input which is enough to induce the cracking generated the SCC near the welding bead. So, in order to prevent the SCC, the residual stress on structure should be relaxed. 2) The branch-type cracking was detected

  15. Ultrasonic inspection of steam generator tubing for cracks, wall thinning and cross-sectional deformation

    International Nuclear Information System (INIS)

    Meyer, P.A.; Carodiskey, T.J.

    1988-01-01

    Periodic inspection of steam generator tubing is an important consideration in the efficient operation of a power generating facility. Since the operating life of these generators is finite, failures will occur. Due to the chemistry of the environment, thermal cycling, and other factors, flaws may develop that can cause rapid deterioration of the tubing while the overall performance of the unit may appear normal. In earlier presentation, the authors presented an ultrasonic bore-side array transducer which can be used with a conventional flaw detector instrument for the location of circumferential crack type defects on the outside tube surface. since that time, much additional experience has been gained on the performance of these probes. Probe performance has been characterized using fatigue crack samples and these results are reviewed. Probes have also been developed having 16 elements for use in larger diameter (25 mm) tubes. The bore-side array concept has been expanded to normal incidence tube well inspection allowing simultaneous wall thickness and eccentricity measurement which is very useful in the assessment of tube wastage and deformation. Preliminary data obtained in this area is presented

  16. Mitigation of caustic stress corrosion cracking of steam generator tube materials by blowdown -a case study

    International Nuclear Information System (INIS)

    Dutta, Anu; Patwegar, I.A.; Chaki, S.K.; Venkat Raj, V.

    2000-01-01

    The vertical U-tube steam generators are among the most important equipment in nuclear power plants as they form the vital link between the reactor and the turbogenerator. Over ∼ 35 years of operating experience of water cooled reactor has demonstrated that steam generator tubes are susceptible to various forms of degradation. This degradation leads to failure and outages of the power plant. A majority of these failures have been attributed to concentrated alkali attacks in the low flow areas such as crevices in the tube to tube sheet joints, baffle plate location and the areas of sludge deposits. Free hydroxides can be produced by improper maintenance of phosphate chemical control in the secondary side of the steam generators and also by the thermal decomposition of impurities present in the condenser cooling water which may leak into the feed water through the condenser tubes. The free hydroxides concentrate in the low flow areas. This buildup of free hydroxide in combination with residual stress leads to caustic stress corrosion cracking. In order to mitigate caustic stress corrosion cracking of Inconel 600 tubes, the trend is to avoid phosphate dosing. Instead All Volatile Treatment (AVT) for secondary water is used backed by full flow condensate polishing. Sodium hydroxide concentration is now being considered as the basis for steam generator blowdown. A methodology has been established for determining the blowdown requirement in order to mitigate caustic stress corrosion cracking in the secondary side of the vertical U-tube natural circulation steam generator. A case study has been carried out for zero solid treatment (AVT coupled with full flow condensate polishing plant) water chemistry. Only continuous blowdown schemes have been studied based on maximum caustic concentration permissible in the secondary side of the steam generator. The methodology established can also be used for deciding concentration of any other impurities

  17. Laboratory results of stress corrosion cracking of steam generator tubes in a complex environment - An update

    Energy Technology Data Exchange (ETDEWEB)

    Horner, Olivier; Pavageau, Ellen-Mary; Vaillant, Francois [EDF R and D, Materials and Mechanics of Components Department, 77818 Moret-sur-Loing (France); Bouvier, Odile de [EDF Nuclear Engineering Division, Centre d' Expertise et d' Inspection dans les Domaines de la Realisation et de l' Exploitation, 93206 Saint Denis (France)

    2004-07-01

    Stress corrosion cracking occurs in the flow-restricted areas on the secondary side of steam generator tubes of Pressured Water Reactors (PWR), where water pollutants are likely to concentrate. Chemical analyses carried out during the shutdowns gave some insight into the chemical composition of these areas, which has evolved during these last years (i.e. less sodium as pollutants). It has been modeled in laboratory by tests in two different typical environments: the sodium hydroxide and the sulfate environments. These models satisfactorily describe the secondary side corrosion of steam generator tubes for old plant units. Furthermore, a third typical environment - the complex environment - which corresponds to an All Volatile Treatment (AVT) environment containing alumina, silica, phosphate and acetic acid has been recently studied. This particular environment satisfactorily reproduces the composition of the deposits observed on the surface of the steam generator tubes as well as the degradation of the tubes. A review of the recent laboratory results obtained by considering the complex environment are presented here. Several tests have been carried out in order to study initiation and propagation of secondary side corrosion cracking for some selected materials in such an environment. 600 Thermally Treated (TT) alloy reveals to be less sensitive to secondary side corrosion cracking than 600 Mill Annealed (MA) alloy. Finally, the influence of some related factors like stress, temperature and environmental factors are discussed. (authors)

  18. Laboratory results of stress corrosion cracking of steam generator tubes in a complex environment - An update

    International Nuclear Information System (INIS)

    Horner, Olivier; Pavageau, Ellen-Mary; Vaillant, Francois; Bouvier, Odile de

    2004-01-01

    Stress corrosion cracking occurs in the flow-restricted areas on the secondary side of steam generator tubes of Pressured Water Reactors (PWR), where water pollutants are likely to concentrate. Chemical analyses carried out during the shutdowns gave some insight into the chemical composition of these areas, which has evolved during these last years (i.e. less sodium as pollutants). It has been modeled in laboratory by tests in two different typical environments: the sodium hydroxide and the sulfate environments. These models satisfactorily describe the secondary side corrosion of steam generator tubes for old plant units. Furthermore, a third typical environment - the complex environment - which corresponds to an All Volatile Treatment (AVT) environment containing alumina, silica, phosphate and acetic acid has been recently studied. This particular environment satisfactorily reproduces the composition of the deposits observed on the surface of the steam generator tubes as well as the degradation of the tubes. A review of the recent laboratory results obtained by considering the complex environment are presented here. Several tests have been carried out in order to study initiation and propagation of secondary side corrosion cracking for some selected materials in such an environment. 600 Thermally Treated (TT) alloy reveals to be less sensitive to secondary side corrosion cracking than 600 Mill Annealed (MA) alloy. Finally, the influence of some related factors like stress, temperature and environmental factors are discussed. (authors)

  19. Framatome experience and programs in relation to guide tube support pin cracking

    International Nuclear Information System (INIS)

    Benhamou, C.; Poitrenaud, P.

    1989-01-01

    Guide tube support pins installed in the upper internals of pressurized water reactors (PWR) have failed by stress corrosion cracking (SCC). Typical pin crack locations are in the first thread area, shank-to-shoulder transition, and at the end of the leaves. The support pins were made of Inconel X-750 with a solution treatment between 885 degree C (1625 degree F) and 1150 degree C (2100 degree F), followed by a single or double aging, depending on the material supplier. EDF and Framatome initiated an extensive program to address the concern for the potential of support pin cracking in 21 operating units. Short-term actions identified the cause of cracking as a combination of inherent high design stress and a material susceptible to stress corrosion cracking. Long-range objectives are to determine the relation between metallurgy and SCC resistance and to decrease the operating stress. Second-generation design improvements to increase SCC resistance included a revised heat treatment of solution annealing at 1093 degree C (2000 degree F) followed by aging at 704 degree C (1300 degree F), and use of a parabolic radius in the shank/shoulder area, and decreasing the installation torque. Third generation changes included an improved torquing procedure, polishing of crack-sensitive areas, and tighter dimensional control. Fourth-generation pin modifications required the use of Inconel X-750 water quenched from the solution-annealing temperature to improve resistance to SCC with thread rolling after aging. Stress corrosion cracking tests of Inconel X-750, 718, and A286 in a PWR environment were performed. Smooth tensile data on Inconel X-750 with the second-generation heat treatment allowed a life prediction of 80,000 hours or 11 years for a stress level about yield strength. The effect of grain size, grain boundary phases, and precipitate morphology on resistance to stress corrosion cracking were also evaluated

  20. Underclad crack development of steam generators tube sheets and reactor vessels nozzles in PWR plants

    International Nuclear Information System (INIS)

    Faure, F.; Bocquet, P.; Boudot, R.; Zacharie, G.

    1985-01-01

    Defects formed, before stress relieving treatment, under the coating of tube plates of steam generators and vessel pipes are cold cracks formed in the segregation zone during surface coating without pre- and postheating of the 2nd layers and eventually of the following coating layers. To solve this problem, the conditions of pre- and post-heating are reinforced and applied to all the coating layers. 13 refs [fr

  1. Analysis of the Causes of Cracks in the Bottom Floor of the Underground Garage of the Hefei Government Affairs Center by using 3D Finite-Element Analysis

    Directory of Open Access Journals (Sweden)

    ZHU Lei

    2015-09-01

    Full Text Available A three-dimensional finite-element software program is used in this study to analyze the causes of cracks in an underground garage. Numerous cracks, serious and regular alike, can be found in the underground garage of the Hefei Government Affairs Center. These cracks are mainly located around the central part of the bottom floor within a 44.6– 57.8 m radius. To explore the causes of the cracks, two attempts are made. On one hand, on-site crack detection and underground water monitoring are conducted. On the other hand, the finite-element software program ANSYS is adopted to establish a finite-element model for the floor–foundation and connecting beam–foundation soil systems of the underground garage. Furthermore, the influences of the underground foundation, underground water level, soil expansion, and Poisson ratio on the bottom floor are calculated and analyzed. On the basis of the calculation and monitoring results, the following conclusion can be made: underground water is the main cause of the bottom floor cracks because underground water exerts a pushing force from the bottom and causes the expansibility of expansive soil. The study aims to provide a theoretical basis for the treatment of cracking in the Hefei Government Affairs Center, and offer a reference for the design, construction, and maintenance of similar projects.

  2. Effect of the surface film electric resistance on eddy current detectability of surface cracks in Alloy 600 tubes

    International Nuclear Information System (INIS)

    Saario, T.; Paine, J.P.N.

    1995-01-01

    The most widely used technique for NDE of steam generator tubing is eddy current. This technique can reliably detect cracks grown in sodium hydroxide environment only at depths greater than 50% through wall. However, cracking caused by thiosulphate solutions have been detected and sized at shallower depths. The disparity has been proposed to be caused by the different electric resistance of the crack wall surface films and corrosion products in the cracks formed in different environments. This work was undertaken to clarify the role of surface film electric resistance on the disparity found in eddy current detectability of surface cracks in alloy 600 tubes. The proposed model explaining the above mentioned disparity is the following. The detectability of tightly closed cracks by the eddy current technique depends on the electric resistance of the surface films of the crack walls. The nature and resistance of the films which form on the crack walls during operation depends on the composition of the solution inside the crack and close to the crack location. During cooling down of the steam generator, because of contraction and loss of internal pressurization, the cracks are rather tightly closed so that exchange of electrolyte and thus changes in the film properties become difficult. As a result, the surface condition prevailing at high temperature is preserved. If the environment is such that the films formed on the crack walls under operating conditions have low electric resistance, eddy current technique will fail to indicate these cracks or will underestimate the size of these cracks. However, if the electric resistance of the films is high, a tightly closed crack will resemble an open crack and will be easily indicated and correctly sized by eddy current technique

  3. Evaluation of Fatigue Crack Initiation for Volumetric Flaw in Pressure Tube

    International Nuclear Information System (INIS)

    Choi, Sung Nam; Yoo, Hyun Joo

    2005-01-01

    CAN/CSA.N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA-N285.05-2005, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of ASME B and PV Sec. XI, 'Inservice Inspection of Nuclear Power Plant Components'. However, the evaluation methodology for a blunt volumetric flaw is described in CSA-N285.05-2005 code. The object of this paper is to address the fatigue crack initiation evaluation for the blunt volumetric flaw as it applies to the pressure tube at Wolsong NPP

  4. Conservatism of present plugging criteria on steam generator tubes and coalescence model of collinear through-wall axial cracks

    International Nuclear Information System (INIS)

    Lee, Jin Ho; Park, Youn Won; Song, Myung Ho; Kim, Young Jin; Moon, Seong In

    1999-01-01

    The steam generator tubing covers a major portion of the primary pressure-retaining boundary, so that very conservative approaches were taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever the cause was. However, it is reported that there is no safety problem even with thickness reductions greater than 40%. Recently, the plant specific plugging criteria are introduced in many countries by demonstrating that the cracked tube has a sufficient safety margin. One of the drawbacks of such criteria, even though not yet codified, is that it is developed based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is general. Their failure analyses have been, therefore, carried out using an idealized single crack to reduce complexity till now. The objective of this paper is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence criterion for twin through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we review the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence criterion, we perform finite element analysis

  5. Primary water stress corrosion cracking resistance of alloy X-750 for guide tube support pins

    International Nuclear Information System (INIS)

    Yonezawa, T.; Onimura, K.; Yonehana, M.; Fujitani, T.

    1990-01-01

    The authors have developed the maintenance free guide tube support pins for PWR, and conducted the three kinds of long time stress corrosion cracking tests in high temperature water, in order to verify the reliability of the maintenance free guide tube support pins. This paper describes the features of our maintenance free support pins and the results of long time stress corrosion cracking test for the maintenance free support pins. After exposure at 320 0 C in the simulated primary water of PWR for about 35,000 hours or at 360 0 C in the same chemistry water as the primary water for about 24,900 hours, no abnormal indication such as cracks was observed in all test support pins exposed 320 0 C and 360 0 C primary water, by ultrasonic inspection, and liquid penetrate test. From the above, it seems that our maintenance free support pins are keepable the soundness up to the end of plant life, in PWR plants

  6. Effect of the environment on a SG tube fatigue cracking at Fessenheim unit 2

    International Nuclear Information System (INIS)

    Duisabeau, L.; Fargeas, E.; Miloudi, S.; Leduc, A.; Hollner, S.; Thebault, Y.; Legras, L.; Mansour, C.

    2015-01-01

    In 2008, a primary-to-secondary leak was detected at TSP n8 level, on the tube R12C62 of Fessenheim unit 2 SG3. The leak was associated to a high cycle fatigue crack that was confirmed two years after, when the tube was pulled out for destructive examination. It revealed on the one hand a highly oxidized fracture surface and on the other hand, that the fatigue crack was initiated on small IGA (Intergranular Attack) piles located at the OD (Outside Diameter) surface of the alloy 600MA tube. In order to take into account a potential environmental effect on the fatigue limit of alloy 600MA in mechanical calculations implemented to establish the root cause failure analysis, several investigations were conducted to evaluate the environment at the tube/tubesheet interstice. To achieve this goal, a multi-scale analysis has been performed. It includes a global analysis of the corrosion damage of the SG, the SG chemistry monitoring, an evaluation of the pH in confined areas with MulteQ calculations based on hide out returns, as well as oxides characterization on the tube by Transmission Electronic Microscopy. All methods converge to a slightly neutral pH with pollutants such as copper, lead and sulfates leading to the conclusion that the fatigue limit of alloy 600MA has not been reduced by the chemical environment. All these chemical elements are known to affect in a certain extent the corrosion resistance of the alloy 600 in the secondary water. If all these pollutants can be detected during the global monitoring of the plant during operation or outage (blow down, hideout returns, feed water and sludge chemical analysis), transmission electronic microscopy offers a unique technique for better understanding how these pollutants may react in confined area, corroded area or free span oxides in the alloy 600 and thus for a better understanding of the corrosion mechanism of nickel based alloys in the secondary side

  7. Ultrasonic inspection of steam-generator tube axial cracking using Lamb wave

    International Nuclear Information System (INIS)

    Park, Jae Seok

    2007-02-01

    In this study, the interaction of Lamb wave propagating thin tube structure with finite vertical discontinuity was studied using both modal decomposition method (MDM) and experimental method. For MDM, a global matrix formulation and orthogonality of Lamb mode was employed to describe the boundary condition of finite vertical discontinuity of the tube and the mode conversion phenomenon respectively. The final form of governing equation by MDM was a linear matrix equation which could be solved using a simple matrix identity. The calculation result showed that, below the cut-off frequency, reflection amplitudes of both A0 and S0 Lamb mode increase as the depth of discontinuity increased beyond the threshold value. An experimental investigation was performed using a Hertzian-contact transducer and steam-generator tubes to verify the calculation results by MDM. A0 Lamb mode was selected as a test signal considering the characteristics of the transducer and previous studies. The experiment for mode identification using half-sectioned tube verified that the Hertzian-contact transducer effectively generated A0 Lamb mode. Tests performed using steam-generator tubes with EDM (electric discharge machined) axial notches showed that the deeper notches produced the higher reflection echo. A0 Lamb mode interacted with the notch having a depth larger than 1/40 of wave length, or corresponding to 30% of the wall thickness. This finding was in good agreement with previous studies and the prediction by MDM. The experiment using real crack specimens to estimate the deviation of reflection amplitude showed that the reflection cross-section of real crack was very similar with that of EDM notch. Therefore, specimens with EDM notches can be used as reference blocks for Lamb wave UT calibration

  8. Acoustic emission characteristics of stress corrosion cracks in a type 304 stainless steel tube

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Woong Gi; Bae, Seung Gi; Lee, Bo Young [School of Aerospace and Mechanical Engineering, Korea Aerospace University, Goyang (Korea, Republic of); Kim, Jae Seong [Center for Robot Technology and Manufacturing, Institute for Advanced Engineering, Yongin (Korea, Republic of); Kang, Sung Sik [Dept. of Nuclear Safety Research, Daejeon (Korea, Republic of); Kwag, Nog Won [Ultrasonic Division, RM910, Byucksan Digital Valley II, Seoul (Korea, Republic of)

    2015-06-15

    Acoustic emission (AE) is one of the promising methods for detecting the formation of stress corrosion cracks (SCCs) in laboratory tests. This method has the advantage of online inspection. Some studies have been conducted to investigate the characteristics of AE parameters during SCC propagation. However, it is difficult to classify the distinct features of SCC behavior. Because the previous studies were performed on slow strain rate test or compact tension specimens, it is difficult to make certain correlations between AE signals and actual SCC behavior in real tube-type specimens. In this study, the specimen was a AISI 304 stainless steel tube widely applied in the nuclear industry, and an accelerated test was conducted at high temperature and pressure with a corrosive environmental condition. The study result indicated that intense AE signals were mainly detected in the elastic deformation region, and a good correlation was observed between AE activity and crack growth. By contrast, the behavior of accumulated counts was divided into four regions. According to the waveform analysis, a specific waveform pattern was observed during SCC development. It is suggested that AE can be used to detect and monitor SCC initiation and propagation in actual tubes.

  9. Dealing with control rod guide tube support pin cracking in French PWRs

    International Nuclear Information System (INIS)

    Guicherd, L.

    1984-01-01

    Cracking and failure of control rod guide tube support pins has been encountered at a number of PWRs around the world. To deal with the problem, the French embarked on an extremely ambitious backfitting programme, involving the installation of replacement pins at all their operating 900MWe units. This highly successful programme, which will be completed in 1985, has been carried out with very low occupational doses and, in the last two years, has required no extensions to annual refuelling outage periods at the plants concerned. The French approach has involved a number of innovations, which should be of considerable interest to other PWR owners worldwide. (author)

  10. Crack initiation and propagation paths in small diameter FSW 6082-T6 aluminium tubes under fatigue loading

    Directory of Open Access Journals (Sweden)

    Roberto Tovo

    2016-03-01

    Full Text Available This paper reports results of fatigue tests of friction stir welded (FSW aluminium tubes. Relatively small 38 mm diameter tubes were used and hence an automated FSW process using a retracting tool was designed for this project, as the wall thickness of the aluminium tube was similar to the diameter of the FSW tool. This is a more complex joint geometry to weld than the more usual larger diameter tube reported in the literature. S-N fatigue testing was performed using load ratios of R = 0.1 and R = -1. Crack path analysis was performed using both low magnification stereo microscopy and scanning electron microscopy, in order to identify crack initiation sites and to determine the direction of crack propagation. Work is still in progress to follow the crack path through the various microstructural zones associated with the weld. A simple statistical analysis was used to characterize the most typical crack initiation site. This work forms part of a wider project directed at determining multiaxial fatigue design rules for small diameter 6082-T6 aluminium tubes that could be of use in the ground vehicle industry.

  11. Delayed hydride cracking behavior of Zr-2.5Nb alloy pressure tubes for PHWR700

    Energy Technology Data Exchange (ETDEWEB)

    Sunil, S.; Bind, A.K.; Khandelwal, H.K.; Singh, R.N., E-mail: rnsingh@barc.gov.in; Chakravartty, J.K.

    2015-11-15

    In order to attain improved in-reactor performance few prototypes pressure tubes of Zr-2.5Nb alloy were manufactured by employing forging to break the cast structure and to obtain more homogeneous microstructure. Both double forging and single forging were employed. The forged material was further processed by employing hot extrusion, cold pilgering and autoclaving. A detailed characterization in terms of mechanical properties and microstructure of the prototype tubes were carried for qualifying it for intended use as pressure tubes in PHWR700 reactors. In this work, Delayed Hydride Cracking (DHC) behavior of the forged Zr-2.5Nb pressure tube material characterized in terms of DHC velocity and threshold stress intensity factor associated with DHC (K{sub IH}) was compared with that of conventionally manufactured material in the temperature range of 200–283 °C. Activation energy associated with the DHC in this alloy was found to be ∼60 kJ/mol for the forged materials.

  12. Crack

    Science.gov (United States)

    ... spending time in a rehab facility or getting cognitive-behavioral therapy or other treatments. Right now, there are no medicines to treat a crack addiction. If you smoke crack, talking with a counselor ...

  13. 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)

  14. Fatigue crack growth behaviour of 21/4Cr1Mo steel tube at elevated temperature

    International Nuclear Information System (INIS)

    Bulloch, J.H.; Buchanan, L.W.

    1987-01-01

    The fatigue crack growth characteristics of 21/4Cr1Mo steel tube have been examined at 588 0 C over the frequency range 0.02-20 Hz and dwell time range 10-960 min. All tests were conducted under load control in laboratory air at an R-ratio of 0.5. The elevated temperature fatigue crack growth characteristics were adequately described in terms of the stress intensity range ΔKAPPA. The continuous cyclic test data exhibited a significant effect of frequency that agreed well with predicted effects using a simple mathematical model of the high temperature fatigue process. With the dwell time range of 10-100 min there was a significant dwell time effect on the critical ΔKAPPA level for creep-fatigue interactive growth. At dwell times > 100 min the dwell time effect saturates. When creep-fatigue interactive growth occurs, growth rates reside above the maximum for continuum-controlled fatigue crack growth, and exhibit a da/dN varies as ΔKAPPA 10 dependence; failure is then intergranular in nature. (author)

  15. Effect of heat treatment and composition on stress corrosion cracking of steam generation tubing materials

    International Nuclear Information System (INIS)

    Kim, H. P.; Hwang, S. S.; Kuk, I. H.; Kim, J. S.; Oh, C. Y.

    1998-01-01

    Effects of heat treatment and alloy composition on stress corrosion cracking (SCC) of steam generator tubing materials have been studied in 40% NaOH at 315.deg.C at potential of +200mV above corrosion potential using C-ring specimen and reverse U bend specimen. The tubing materials used were commercial Alloy 600, Alloy 690 and laboratory alloys, Ni-χCr-10Fe. Commercial Alloy 600, Alloy 690 were mill annealed or thermally treated.Laboratory alloy Ni-χCr-10Fe, and some of Alloy 600 and Alloy 690 were solution annealed. Polarization curves were measured to find out any relationship between SCC susceptibility and electrochemical behaviour. The variation in thermal treatment of Alloy 600 and Alloy 690 had no effect on polarization behaviour probably due to small area fraction of carbide and Cr depletion zone near grain boundary. In anodic polarization curves, the first and second anodic peaks at about 170mV and about at 260mV, respectively, above corrosion potential were independent of Cr content, whereas the third peak at 750mV above corrosion potential and passive current density in-creased with Cr content. SCC susceptibility decreased with Cr content and thermal treatment producing semicontinuous grain boundary decoration. Examination of cross sectional area of C-ring specimen showed deep SCC cracks for the alloys with less than 17%Cr and many shallow attacks for alloy 690. The role of Cr content in steam generator tubing materials and grain boundary carbide on SCC were discussed

  16. Evaluation of plugging criteria on steam generator tubes and coalescence model of collinear axial through-wall cracks

    International Nuclear Information System (INIS)

    Lee, Jin Ho; Park, Youn Won; Song, Myung Ho; Kim, Young Jin; Moon, Seong In

    2000-01-01

    In a nuclear power plant, steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus very conservative approaches have been taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess 40% should be plugged whatever causes are. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about twenty years ago when wear and pitting were dominant causes for steam generator tube degradation. And it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram

  17. Determination of delayed hydride cracking velocity of CANDU Zr-2.5Nb pressure tube

    International Nuclear Information System (INIS)

    Kim, Young Suk; Kim, Chan Jung; Rheem, Y. W.; Im, K. S.; Kwon, Sang Chul

    2000-07-01

    As agreed upon the contract with an IAEA Co-ordinated Research Project 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium Based Alloys', we conducted DHC tests at 3 different temperatures of 144, 182 and 250 deg C on the curved compact tension specimens made from a Zr-2.5Nb pressure tube. Additional tests were carried out at 200 and 230 deg C with an aim to determine the activation energy for delayed hydride cracking. This report summarizes the results of DHC tests obtained so far. All the DHC tests were conducted in accordance with the procedures suggested by the Host Lab. 7 DHCV values determined at the same temperature such as 250 deg C show very low standard deviation, whose average values are very comparable to those reported by the participants. Thus, one of the most important results we have got is that we establish qualified DHC testing procedure through the IAEA CRP. An activation energy for DHC of unirradiated Zr-2.5Nb pressure tube was 49 KJ/mol which is very similar to the activation energy of 43 KJ/mol for irradiated Zr-2.5Nb pressure tubes. DHCV increased linearly with the hydrogen content up to around 25 ppm and then became saturated at higher hydrogen concentration

  18. Fatigue crack initiation at complex flaws in hydrided Zr-2.5%Nb samples from CANDU pressure tubes

    International Nuclear Information System (INIS)

    Stoica, L.; Radu, V.

    2016-01-01

    The paper addresses the phenomena which occur at locations where the oxide layer of the inner surface of CANDU tube pressure is damaged by the contact with the fuel element or due to the action of hard particles at the interface between the tube pressure and bearing pad of fuel element. In such situations generate defects, which most often are defects known as ''bearing pad fretting flaws'' or ''debris fretting flaws''. In this paper the experiments are completed in a series of previous works on the mechanical fatigue phenomenon on samples prepared from the pressure tube Zr-2.5% Nb alloy. The phenomenon of variable mechanical stress (or fatigue) may lead to initiation of cracks at the tip of volumetric flaws, according to the accumulation of hydrides, which then fractures and can propagate through the tube wall pressure due to the mechanism of type DHC (Delayed Hydride Cracking). (authors)

  19. Comparative evaluation of preventive measures against primary side stress corrosion cracking of mill annealed Inconel 600 steam generator tubes

    International Nuclear Information System (INIS)

    Frederick, G.; Hernalsteen, P.

    1986-01-01

    Significant amounts of primary side cracking have been reported in the mechanically expanded area of the tubes of PWR steam generators in Europe, in Japan and to a lesser extent in the USA. The Belgian utilities are faced with the same problem. At Doel 2, where the tubes are rolled for only a part of the tubesheet, primary side cracking appeared in the roll transition. The Doel 3 and Tihange 2 steam generators, whose tubes are expanded for the full depth of the tube sheet, have experienced cracking after about 10 000 h of operation not only in the roll transition but also at roll overlaps. While some leaks and eddy current indications are associated with tubesheet or rolling anomalies, many of them are found on normal tubes. A programme was launched by the Belgian utilities and was further co-sponsored by the Electric Power Research Institute (EPRI) to develop preventive actions applicable not only to hot steam generators but also to cold steam generators already installed on site. These preventive measures include stress relaxation and metallurgical improvement of the material by an in situ heat treatment of the whole tube sheet (a steam generator model was used to evaluate the feasibility of this treatment), and the introduction of residual compressive stresses on ID by rotopeening or shotpeening without inducing unacceptable tensile stresses on OD. A comparative evaluation of these measures was established on the basis of tests performed on representative mock-ups and specimens. (author)

  20. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1992-03-01

    The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

  1. Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Weiguo; Gao Fengqin; Zhou Hongyi

    1993-01-01

    The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

  2. A Fundamental study of remedial technology development to prevent stress corrosion cracking of steam generator tubing

    Energy Technology Data Exchange (ETDEWEB)

    Park, In Gyu; Lee, Chang Soon [Sunmoon University, Asan (Korea)

    1998-04-01

    Most of the PWR Steam generators with tubes in Alloy 600 alloy are affected by Stress Corrosion Cracking, such as PWSCC(Primary Water Stress Corrosion Cracking) and ODSCC(Outside Diameter Stress Corrosion Cracking). This study was undertaken to establish the background for remedial technology development to prevent SCC. in the report are included the following topics: (1) General: (i) water chemistry related factors, (ii) Pourbaix(Potential-pH) Diagram, (iii) polarization plot, (iv) corrosion mode of Alloy 600, 690, and 800, (v) IGA/SCC growth rate, (vi) material suspetibility of IGA/SCC, (vii) carbon solubility of Alloy 600 (2) Microstructures of Alloy 600 MA, Alloy 600 TT, Alloy 600 SEN Alloy 690 TT(Optical, SEM, and TEM) (3) Influencing factors for PWSCC initiation rate of Alloy 600: (i) microstructure, (ii) water chemistry(B, Li), (iii) temperature, (iv) plastic deformation, (v) stress relief annealing (4) Influencing factors for PWSCC growth rate of Alloy 600: (i) water chemistry(B, Li), (ii) Scott Model, (iii) intergranular carbide, (iv) temperature, (v) hold time (5) Laboratory conditions for ODSCC initiation rate: 1% NaOH, 316 deg C; 1% NaOH, 343 deg C; 50% NaOH, 288 deg C; 10% NaOH, 302 deg C; 10% NaOH, 316 deg C; 50% NaOH, 343 deg C (6) Sludge effects for ODSCC initiation rate: CuO, Cr{sub 2}O{sub 3}, Fe{sub 3}O{sub 4} (7) Influencing factors for PWSCC growth rate of Alloy 600: (i) Caustic concentration effect, (ii) carbonate addition effect (8) Sulfate corrosion: (i) sulfate ratio and pH effect, (ii) wastage rate of Alloy 600 and Alloy 690 (9) Crevice corrosion: (i) experimental setup for crevice corrosion, (ii) organic effect, (iii) (Na{sub 2}SO{sub 4} + NaOH) effect (10) Remedial measures for SCC: (i) Inhibitors, (ii) ZnO effect. (author). 30 refs., 174 figs., 51 tabs.

  3. Stress corrosion cracking of steam generator tubing materials in lead containing solution

    International Nuclear Information System (INIS)

    Kim, H.P.; Hwang, S.S.; Kim, J.S.; Hong, J.H.

    2007-01-01

    Stress corrosion cracking (SCC) in lead (Pb) containing environments has been one of key issues in the nuclear power industry since Pb had been identified as a cause of the SCC of steam generator (SG) tubing materials in some power plants. To mitigate or prevent degradation of SG tubing materials, a mechanistic understanding of SCC in Pb containing environment is needed, along with an understanding of the source and transport behaviors of Pb species in the secondary circuit. In this work, SCC behaviors of Alloy 600 in Pb containing environments were studied. Influences of microstructures of Alloy 600 and the inhibitive additives were investigated using the C-ring and the slow strain rate tests in caustic solution and demineralized water at 315 o C. Microstructures of Alloy 600 were varied by heat treatment at different temperatures. The additives examined were nickel boride (NiB) and cerium boride (CeB 6 ). The surface films were analyzed using Auger Electron Spectroscopy (AES) and Energy Dispersive X-ray Spectroscopy (EDS). The SCC mode varied with microstructure. Effectiveness of the additives in Pb containing environments is discussed. (author)

  4. Tube structural integrity evaluation of Palo Verde Unit 1 steam generators for axial upper-bundle cracking

    International Nuclear Information System (INIS)

    Woodman, B.W.; Begley, J.A.; Brown, S.D.; Sweeney, K.; Radspinner, M.; Melton, M.

    1995-01-01

    The analysis of the issue of upper bundle axial ODSCC as it apples to steam generator tube structural integrity in Unit 1 at the Palo Verde Nuclear generating Station is presented in this study. Based on past inspection results for Units 2 and 3 at Palo Verde, the detection of secondary side stress corrosion cracks in the upper bundle region of Unit 1 may occur at some future date. The following discussion provides a description and analysis of the probability of axial ODSCC in Unit 1 leading to the exceedance of Regulatory Guide 1.121 structural limits. The probabilities of structural limit exceedance are estimated as function of run time using a conservative approach. The chosen approach models the historical development of cracks, crack growth, detection of cracks and subsequent removal from service and the initiation and growth of new cracks during a given cycle of operation. Past performance of all Palo Verde Units as well as the historical performance of other steam generators was considered in the development of cracking statistics for application to Unit 1. Data in the literature and Unit 2 pulled tube examination results were used to construct probability of detection curves for the detection of axial IGSCC/IGA using an MRPC (multi-frequency rotating panake coil) eddy current probe. Crack growth rates were estimated from Unit 2 eddy current inspection data combined with pulled tube examination results and data in the literature. A Monte-Carlo probabilistic model is developed to provide an overall assessment of the risk of Regulatory Guide exceedance during plant operation

  5. Stress corrosion cracking susceptibility of steam generator tube materials in AVT (all volatile treatment) chemistry contaminated with lead

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Castano, M.L.; Garcia, M.S.

    1996-01-01

    Alloy 600 steam generator tubing has shown a high susceptibility to stress corrosion degradation at the operation conditions of pressurized water reactors. Several contaminants, such as lead, have been postulated as being responsible for producing the secondary side stress corrosion cracking that has occurred mainly at the location where these contaminants can concentrate. An extensive experimental work has been carried out in order to better understand the effects of lead on the stress corrosion cracking susceptibility of steam generator tube materials, namely Alloys 600, 690 and 800. This paper presents the experimental work conducted with a view to determining the influence of lead oxide concentration in AVT (all volatile treatment) conditions on the stress corrosion resistance of nickel alloys used in the fabrication of steam generator tubing. (orig.)

  6. A method of multi-crack shape identification from eddy current testing signals of steam generator tubes including support plates as noise sources

    International Nuclear Information System (INIS)

    Nagaya, Yoshiaki; Endo, Hisashi; Takagi, Toshiyuki; Uchimoto, Tetsuya

    2005-01-01

    This paper deals with identifying multiple cracks from eddy current testing (ECT) signals obtained in a steam generator tube with a support plate and deposits. Assume two-dimensionally scanned ECT signals to be a picture image, then the signal processing by a multi-frequency technique eliminates noise caused by the support plate and deposits. A template matching with help of genetic algorithms detects number and positions of cracks from the image after the signal processing. Inverse analysis estimates the crack profile based on the predicted position of cracks. The number and positions of the cracks are sufficiently well predicted. Crack shape reconstructions are achieved with a satisfactory degree of accuracy. (author)

  7. Size determinations, by ultrasonic techniques, of cracks in hydride blisters formed in Zr-2.5 % Nb pressure tubes

    International Nuclear Information System (INIS)

    Trujillo Badillo, Giovanna; Desimone, Carlos; Domizzi, Gladys

    1999-01-01

    Non destructive techniques (NDT) are very useful in the detection of flaws produced in structural components in service. During the service of CANDU nuclear power reactors, it is possible that pressure tubes (PT) may contact calandria tubes (CT). After the PT/CT contact, zirconium hydride blisters may form at the point of contact depending on the concentration of hydrogen/deuterium. Zirconium hydride is brittle and is therefore prone to cracking under stress. Ultrasonic NDT is routinely use during PT in service inspection. In order to be able of detecting cracked blisters, it is of great importance the development of standards to calibrate the employed equipment. On this purpose, hydride blisters were grown, in laboratory, on sections of pressure tube. The cracks in the blisters were detected and measured by ultrasonic techniques. The obtained results were compared with measurements carried out in optic microscope, on successive sections of the samples. The crack tip diffraction technique was found to be the more effective for the mentioned ends. (author)

  8. Development of an ECT technique for discriminating between a through-wall/non through-wall crack in a steam generator tube

    International Nuclear Information System (INIS)

    Lee, D.H.; Choi, M.S.; Hur, D.H.; Han, J.H.; Lee, U.C.

    2004-01-01

    In a lot of pressurized water reactors, PWSCC (primary water stress corrosion cracking) has been observed in the expansion or u-bend transitions of the alloy 600 steam generator tubes. Particularly, the development of a through-wall crack may cause leakage of the primary coolant during operation and the resultant forced outage, and the in-situ pressure test has often been used to evaluate the integrity and the leakage of the cracked tube during an in-service inspection. However, this process requires additional equipment and hours, and the tested tubes are plugged due to the plastic deformation induced by the internal pressurization. This paper describes a new evaluation technique of the Eddy current test, by which it can be determined whether the crack is through-wall or non through-wall. The technique is based upon the analysis of the characteristics in the Eddy current signals from the crack and also includes the method of measuring exact through-wall length of the crack. The proof and applicability of the technique is discussed with the results of the destructive tests on cracked tubes extracted from a commercial power plant. Also, the effect of crack opening, which reflects the driving force for a crack propagation of non through-wall crack and the leak rate of a through-wall crack, upon the characteristics of the Eddy current signals from the coils of the motorized rotating probe is investigated and discussed using steam generator tube samples with manufactured axial cracks of a through-wall and non through-wall. (orig.)

  9. Predicting crack instability behavior of burst tests from small specimens for irradiated Zr-2.5Nb pressure tubes

    International Nuclear Information System (INIS)

    Davies, P.H.

    1997-01-01

    A scaling approach, based on the deformation J-integral at maximum load obtained from small specimens, is proposed for predicting the crack instability behavior of burst tests on irradiated Zr-2.5Nb pressure tubes. An assessment of this approach is carried out by comparison with other toughness criteria such as the modified J-integral and the plastic work dissipation rate approach. The largest discrepancy between the different parameters occurs for materials of intermediate toughness which exhibit the most stable crack growth and tunnelling up to maximum load. A study of one material of intermediate toughness suggests crack-front tunnelling has a significant influence on the results obtained from the 17-mm-wide specimens. It is shown that for a tube of intermediate toughness the different approaches can significantly underpredict the extent of stable crack growth before instability in a burst test even after correcting for tunnelling. The usefulness of a scaling approach in reducing the discrepancy between the small- and large-scale specimen results for this material is demonstrated

  10. Delayed Hydride Cracking in Zr-2.5Nb Tubes with the Direction of An Approach to Temperature

    International Nuclear Information System (INIS)

    Kim, Young Suk; Im, Kyung Soo; Kim, Kang Soo; Ahn, Sang Bok; Cheong, Yong Moo

    2006-01-01

    One of the unique features of delayed hydride cracking (DHC) of zirconium alloys is that the DHC velocity (DHCV) of zirconium alloys strongly depends on the path to the test temperature. Ambler reported that the DHCV of Zr-2.5Nb tubes at temperatures above 180 .deg. C depended upon the direction of an approach to the test temperatures, and reported on a presence of the DHC arrest temperature or TDAT above which the DHCV decreased upon an approach to the test temperature by a heating. Ambler proposed a hydrogen transfer from the bulk to the crack tip assuming that the hydrides formed at the crack tip and in the bulk region are fully constrained and partially constrained at the crack tip, respectively. In other words, the terminal solid solubility (TSS) of hydrogen would be governed by elastic strain energy induced by the precipitating hydrides, leading to a higher TSS in the bulk region than that at the crack tip. In a sense, his assumption that the hydrogen concentration is higher in the bulk region than that at the crack tip due to a higher TSS in the bulk region is, in a way, similar to Kim's DHC model. Even though Ambler assumed a different strain energy of the matrix hydrides with the direction of an approach to the test temperature, the peak temperature, hydrogen concentration and the hydride phase, a feasible rationale for this assumption is yet to be given. In this study, a path dependence of DHC velocity of Zr-2.5Nb tubes will be investigated using Kim's DHC model where a driving force for DHC is the supersaturated hydrogen concentration between the crack tip and the bulk region. To this ends, the furnace cooled and water-quenched Zr-2.5Nb specimens were subjected to DHC tests at different test temperatures that were approached by a heating or by a cooling. Kim's DHC model predicts that the water-quenched Zr- 2.5Nb will have DHC crack growth even at temperatures above 180 .deg. C where the furnace-cooled Zr-2.5Nb will not. This experiment will provide

  11. Delayed Hydride Cracking in Zr-2.5Nb Tubes with the Direction of An Approach to Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Im, Kyung Soo; Kim, Kang Soo; Ahn, Sang Bok; Cheong, Yong Moo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    One of the unique features of delayed hydride cracking (DHC) of zirconium alloys is that the DHC velocity (DHCV) of zirconium alloys strongly depends on the path to the test temperature. Ambler reported that the DHCV of Zr-2.5Nb tubes at temperatures above 180 .deg. C depended upon the direction of an approach to the test temperatures, and reported on a presence of the DHC arrest temperature or TDAT above which the DHCV decreased upon an approach to the test temperature by a heating. Ambler proposed a hydrogen transfer from the bulk to the crack tip assuming that the hydrides formed at the crack tip and in the bulk region are fully constrained and partially constrained at the crack tip, respectively. In other words, the terminal solid solubility (TSS) of hydrogen would be governed by elastic strain energy induced by the precipitating hydrides, leading to a higher TSS in the bulk region than that at the crack tip. In a sense, his assumption that the hydrogen concentration is higher in the bulk region than that at the crack tip due to a higher TSS in the bulk region is, in a way, similar to Kim's DHC model. Even though Ambler assumed a different strain energy of the matrix hydrides with the direction of an approach to the test temperature, the peak temperature, hydrogen concentration and the hydride phase, a feasible rationale for this assumption is yet to be given. In this study, a path dependence of DHC velocity of Zr-2.5Nb tubes will be investigated using Kim's DHC model where a driving force for DHC is the supersaturated hydrogen concentration between the crack tip and the bulk region. To this ends, the furnace cooled and water-quenched Zr-2.5Nb specimens were subjected to DHC tests at different test temperatures that were approached by a heating or by a cooling. Kim's DHC model predicts that the water-quenched Zr- 2.5Nb will have DHC crack growth even at temperatures above 180 .deg. C where the furnace-cooled Zr-2.5Nb will not. This experiment

  12. Evaluation of nondestructive evaluation size measurement for integrity assessment of axial outside diameter stress corrosion cracking in steam generator tubes

    International Nuclear Information System (INIS)

    Joo, Kyung Mun; Hong, Jun Hee

    2015-01-01

    Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy 600 HTMA tubes has been increasing. As a result, SGs with Alloy 600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and ability of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

  13. Evaluation of nondestructive evaluation size measurement for integrity assessment of axial outside diameter stress corrosion cracking in steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Kyung Mun [Korea Hydro and Nuclear Power Company Ltd., Central Research Institute, Daejeon (Korea, Republic of); Hong, Jun Hee [Dept. of mechanical Engineering, Chungnam National University, Daejeon (Korea, Republic of)

    2015-02-15

    Recently, the initiation of outside diameter stress corrosion cracking (ODSCC) at the tube support plate region of domestic steam generators (SG) with Alloy 600 HTMA tubes has been increasing. As a result, SGs with Alloy 600 HTMA tubes must be replaced early or are scheduled to be replaced prior to their designed lifetime. ODSCC is one of the biggest threats to the integrity of SG tubes. Therefore, the accurate evaluation of tube integrity to determine ODSCC is needed. Eddy current testing (ECT) is conducted periodically, and its results could be input as parameters for evaluating the integrity of SG tubes. The reliability of an ECT inspection system depends on the performance of the inspection technique and ability of the analyst. The detection probability and ECT sizing error of degradation are considered to be the performance indices of a nondestructive evaluation (NDE) system. This paper introduces an optimized evaluation method for ECT, as well as the sizing error, including the analyst performance. This study was based on the results of a round robin program in which 10 inspection analysts from 5 different companies participated. The analysis of ECT sizing results was performed using a linear regression model relating the true defect size data to the measured ECT size data.

  14. Elucidating the iodine stress corrosion cracking (SCC) process for zircaloy tubing

    International Nuclear Information System (INIS)

    Nagai, M.; Shimada, S.; Nishimura, S.; Amano, K.

    1984-01-01

    Several experimental investigations were made to enhance understanding of the iodine stress corrosion cracking (SCC) process for Zircaloy: (1) oxide penetration process, (2) crack initiation process, and (3) crack propagation process. Concerning the effect of the oxide layer produced by conventional steam-autoclaving, no significant difference was found between results for autoclaved and as-pickled samples. Tests with 15 species of metal iodides revealed that only those metal iodides which react thermodynamically with zirconium to produce zirconium tetraiodide (ZrI 4 ) caused SCC of Zircaloy. Detailed SEM examinations were made on the SCC fracture surface of irradiated specimens. The crack propagation rate was expressed with a da/dt=C Ksup(n) type equation by combining results of tests and calculations with a finite element method. (author)

  15. Searching for pelvic floor muscle exercises on YouTube: what individuals may find and where this might fit with health service programmes to promote continence.

    Science.gov (United States)

    Stephen, Kate; Cumming, Grant P

    2012-09-01

    This paper describes the investigation, categorization/characterization and viewing of pelvic floor muscle exercises (PFME) on YouTube from the perspective of the 'wisdom of the crowd'. The aim of the research was to increase awareness of the type of clips that individuals are likely to come across when searching YouTube and to describe trends and popularity. This awareness will be useful for the design of continence promotion services, especially for hard-to-reach individuals. Web-based videos relating to PFE were identified by searching YouTube using the snowball technique. Main outcome measures Number of views; the approach taken (health, fitness, sexual and pregnancy); product promotion; and the use of music, visual cues and elements designed to encourage exercise. The number of views of each video was recorded at three points over a seven-month period. Twenty-two videos were identified. Overall these videos had been viewed over 430,000 times during the study period. One video was viewed over 100,000 times and overall the median increase in views was 59.4%. YouTube is increasingly used to access information about pelvic floor exercises. Different approaches are used to communicate PFME information but there are no formal structures for quality control. Further research is required to identify which elements of the video clips are effective in communicating information and in motivating exercise and to establish appropriate protocols. Kitemarking is recommended in order that women obtain correct advice.

  16. Study on the PWSCC Crack Growth Rate for Steam Generator Tubing

    International Nuclear Information System (INIS)

    Kang, Shin Hoo; Hwang, Il Soon; Lim, Jun; Lee, Seung Gi; Ryu, Kyung Ha

    2008-03-01

    Using in-situ Raman spectroscopy and crack growth rate lest system in simulated PWR primary water environment, the relationship between the oxide film chemistry and the PWSCC growth rate has been studied. We used I/2T compact tension specimen and disk specimen made of Alloy 182 and Alloy 600 for crack growth rate test and in-situ Raman spectroscopy measurement. Test was made in a refreshed autoclave with 30 cc STP / kg of dissolved hydrogen concentration. Conductivity, pH, dissolved hydrogen and oxygen concentration were continuously monitored at the outlet. The crack growth rate was measured by using switching DCPD technique under cyclinc triangular loading and at the same time oxide phase was determined by using in-situ Raman spectra at the elevation of the temperature. Additionally Raman spectroscopy was achieved for oxide phase transition of Alloy 600 according to the temperature and dissolved hydrogen concentration, 2 and 30cc STP / kg

  17. Numerical Analysis of Hot Cracking in Laser-Hybrid Welded Tubes

    Directory of Open Access Journals (Sweden)

    Moritz Oliver Gebhardt

    2013-01-01

    Full Text Available In welding experiments conducted on heavy wall pipes, the penetration mode (full or partial penetration occurred to be a significant factor influencing appearance of solidification cracks. To explain the observed phenomena and support further optimization of manufacturing processes, a computational model was developed, which used a sophisticated strategy to model the material. High stresses emerged in the models in regions which showed cracking during experiments. In partial penetration welding, they were caused by the prevention of weld shrinkage due to the cold and strong material below the joint. Another identified factor having an influence on high stress localization is bulging of the weld.

  18. A simple mechanistic model for particle penetration and plugging in tubes and cracks

    Energy Technology Data Exchange (ETDEWEB)

    Mitrakos, D.; Chatzidakis, S. [' Demokritos' National Centre for Scientific Research, Institute of Nuclear Technology and Radiation Protection, 15310 Agia Paraskevi, Athens (Greece); National Technical University of Athens, Faculty of Mechanical Engineering, 15780 Athens (Greece); Hinis, E.P. [National Technical University of Athens, Faculty of Mechanical Engineering, 15780 Athens (Greece); Herranz, L.E. [CIEMAT, Nuclear Safety Unit, Avda. Complutense 22, Madrid 28040 (Spain); Parozzi, F. [Department of Generation Systems, CESI RICERCA, Milano (Italy); Housiadas, C. [' Demokritos' National Centre for Scientific Research, Institute of Nuclear Technology and Radiation Protection, 15310 Agia Paraskevi, Athens (Greece)], E-mail: christos@ipta.demokritos.gr

    2008-12-15

    In the course of a severe accident, some nuclear aerosols may be released to the environment through penetrating the containment concrete cracks, even if a catastrophic failure of the containment does not occur. There is experimental and theoretical evidence of strong retention of aerosol particles in the cracks that act as a filter. In this work a Eulerian model is developed based on the numerical solution of the one-dimensional aerosol transport equation. Plug formation is accommodated by allowing the crack diameter to change with time, based on the volume of the deposited mass. Brownian diffusion, gravitational settling and turbulence-driven deposition are considered as the removal mechanisms of the particles along the leak path. The model is verified against analytical solutions and validated by comparing with early as well as recent experimental data. It is concluded that a one-dimensional model of aerosol flow through a hydraulically equivalent leaking duct can simulate with enough accuracy aerosol transport in cracks, so that it may be an appropriate option for a large system code like ASTEC.

  19. Diagnosis of 3-dimensional geometry and stress corrosion cracking in steam generator tubes

    International Nuclear Information System (INIS)

    Lee, D.H.; Choi, M.S.; Hur, D.H.; Kim, K.M.; Han, J.H.; Song, M.H.

    2015-01-01

    Most of the corrosive degradations in steam generator tubes of nuclear power plants are closely related to the residual stress existing in the local region of a geometric change, that is, an expansion transition, u-bend, dent, bulge, etc. Therefore, accurate information on a geometric anomaly (precursor of degradation) in a tube is a prerequisite to the activity of pre- and in-service non destructive inspection for a precise and earlier detection of a defect in order to prevent a failure during an operation, and also for a root cause analysis of a failure. In this paper, a new diagnostic eddy current probe technology which has simultaneous dual function of a 3-dimensional geometry measurement and defect detection in steam generator tube is introduced. The D-Probe is a rotary type eddy current coil probe equipped with 3 different eddy current coil units (surface riding type plus-point and pancake coils for defect detection, and non-surface riding type shielded high frequency pancake coil for tube profile measurement). A specific data analysis software has been developed. By comparing the eddy current data from the defect with those from the geometric changes, the relationship between the degradation and geometric changes can be revealed. Also, it supplies information on tube location at which defect is most probable and thus, a more efficient detection of earlier degradation. The use of D-probe and analysis software has been demonstrated for steam generator tubes with various geometric anomalies in manufacturing and operating nuclear power plants

  20. Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

    International Nuclear Information System (INIS)

    Kim, Young Suk

    2007-01-01

    In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

  1. Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

    International Nuclear Information System (INIS)

    Gutkin, L.; Scarth, D.A.

    2014-01-01

    Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

  2. Probabilistic modeling of material resistance to crack initiation due to hydrided region overloads in CANDU Zr-2.5%Nb pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Gutkin, L.; Scarth, D.A. [Kinectrics Inc., Toronto, ON (Canada)

    2014-07-01

    Zr-2.5%Nb pressure tubes in CANDU nuclear reactors are susceptible to hydride-assisted cracking at the locations of stress concentration, such as in-service flaws. Probabilistic methodology is being developed to evaluate such flaws for crack initiation due to hydrided region overloads, which occur when the applied stress acting on a flaw with an existing hydrided region at its tip exceeds the stress at which the hydrided region is formed. As part of this development, probabilistic modeling of pressure tube material resistance to overload crack initiation has been performed on the basis of a set of test data specifically designed to study the effects of non-ratcheting hydride formation conditions and load reduction prior to hydride formation. In the modeling framework, the overload resistance is represented as a power-law function of the material resistance to initiation of delayed hydride cracking under constant loading, where both the overload crack initiation coefficient and the overload crack initiation exponent vary with the flaw geometry. In addition, the overload crack initiation coefficient varies with the extent of load reduction prior to hydride formation as well as the number of non-ratcheting hydride formation thermal cycles. (author)

  3. Liquid metal fast breeder reactor steam generator: behaviour of heat exchange tubes in face of a through crack resulting in a contact between sodium and water

    International Nuclear Information System (INIS)

    Quinet, J.L.; Lannou, L.

    1978-01-01

    The results of a survey made Electricite de France on the behaviour of cracked tubes under operating conditions of an industrial steam generator are submitted in this communication. A comparison is made of the tube material: INCOLOY 800, 2 1/4 Cr-1 Mo, 9 Cr-2 Mo land to the initial leak. Finally, a description is given of the self-development process of a water leak into sodium. (author)

  4. Estimating the number of latent cracks in pressure tube joints at Bruce unit 2

    International Nuclear Information System (INIS)

    Schwarz, C.J.

    1983-10-01

    A model was built to estimate the number of hydride cracks which might have arisen in the rolled joints of Bruce unit 2 prior to the stress relieving operation. The model estimated that about 100 such cracks might exist. Since this estimate is based on experiments that were thermally cycled and since cycling did not occur in Bruce, prior to stress relieving the actual number is expected to be substantially lower. A sensitivity analysis of the model showed that it is sensitive to the assumptions of stress levels, probability of initiation and distribution of initiation time. A better estimate could be made if more data were available on these parameters under realistic conditions. Therefore, the recommendation is made to collect more information about these factors under realistic conditions

  5. Analysis on the Acoustic Emission Signals in the Crack Evolution of Steam Generator Tube

    International Nuclear Information System (INIS)

    Han, Jung Ho; Hur, Do Haeng; Kim, Kyung Mo; Choi, Myung Sik; Lee, Deok Hyun

    2007-01-01

    The evolution of a defect in steam generator (SG) tube during plant operation can be classified into the stages of initiation and propagation. However, the detection and discrimination of these two stages are difficult, and the real time monitoring of the defect evolution in plant operation is impossible. Moreover, it was generally known that the commercial nondestructive examination techniques such as eddy current test(ECT) can detect the defect already grown up to the size of more than 40% in tube wall thickness. Therefore, the scope of the present study is to develop the fundamental technology for monitoring the degradation process from the initiation stage to the subsequent propagation stage by acoustic emission (AE) signal measurement

  6. Cracking and Corrosion of Composite Tubes in Black Liquor Recovery Boiler Primary Air Ports

    Energy Technology Data Exchange (ETDEWEB)

    Keiser, James R.; Singbeil, Douglas L.; Sarma, Gorti B.; Kish, Joseph R.; Yuan, Jerry; Frederick, Laurie A.; Choudhury, Kimberly A.; Gorog, J. Peter; Jetté, Francois R.; Hubbard, Camden R.; Swindeman, Robert W.; Singh, Prett M.; Maziasz, Phillip J.

    2006-10-01

    Black liquor recovery boilers are an essential part of kraft mills. Their design and operating procedures have changed over time with the goal of providing improved boiler performance. These performance improvements are frequently associated with an increase in heat flux and/or operating temperature with a subsequent increase in the demand on structural materials associated with operation at higher temperatures and/or in more corrosive environments. Improvements in structural materials have therefore been required. In most cases the alternate materials have provided acceptable solutions. However, in some cases the alternate materials have solved the original problem but introduced new issues. This report addresses the performance of materials in the tubes forming primary air port openings and, particularly, the problems associated with use of stainless steel clad carbon steel tubes and the solutions that have been identified.

  7. 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)

  8. 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)

  9. Actions undertaken in order to solve the tube guide pins cracking

    International Nuclear Information System (INIS)

    Benhamou, C.; Briois, J.P.; Tuncer, T.; Drean, H.

    1985-01-01

    Short-term actions have been developed after the discovery of defects on the guide tube pins (upper internal rector components) of 21 units in France. The impact analysis of a pin failure on the safety by Framatome show that short-term consequences are not very important. Improvements in regard with the design and the stress corrosion resistance of the material, which can be applied to the pins, have been defined. Framatome and EDF developed tools and the mastery of replacement operations what led to limit considerably the unavailability of units; 17 units have been treated for two years and half; the four last ones will be treated in 1985. Parallel studies allow to estimate the available marge of in-service behavior of the replacement pins; this marge is increased in units which are built now by means of new developments concerning the design and the mounting conditions. A method and tools utilizing ultrasonic waves have been developed for on-site inspection of the pins [fr

  10. Iodine stress corrosion cracking (SCC) of unirradiated Zircaloy-4 tubing by means of internal gas pressurization, (1)

    International Nuclear Information System (INIS)

    Onchi, Takeo; Inoue, Tadashi

    1982-01-01

    The internal gas pressurization tests were conducted at 360 0 C, to examine the influence of iodine concentration on the iodine stress corrosion cracking (SCC) susceptibility of Zircaloy-4 tubing of 17 x 17 type PWR design. The iodine contents studied were ranging of 0.06 to 6 mg/cm 2 , corresponding to 30 from 0.3 mg/cm 3 . Applied hoop stress vs. time-to-failure relationships were obtained in argon gas with iodine, as well as without iodine, from the tests of maximum holding times up to 72 hrs. The relationships obtained were insensitive to iodine contents. The applied stress lowering in iodine atmosphere approached a threshold stress below which SCC failure did not occur within the holding time, but not in argon gas alone. The threshold stresses were approximately 25.5 kg/mm 2 (250 Mpa), independent on iodine concentrations. Based on fracture mechanics approach and fractographic analysis, an interpretation was made of those applied stress and time-to-failure relationships. (author)

  11. Effect of error in crack length measurement on maximum load fracture toughness of Zr-2.5Nb pressure tube material

    International Nuclear Information System (INIS)

    Bind, A.K.; Sunil, Saurav; Singh, R.N.; Chakravartty, J.K.

    2016-03-01

    Recently it was found that maximum load toughness (J max ) for Zr-2.5Nb pressure tube material was practically unaffected by error in Δ a . To check the sensitivity of the J max to error in Δ a measurement, the J max was calculated assuming no crack growth up to the maximum load (P max ) for as received and hydrogen charged Zr-2.5Nb pressure tube material. For load up to the P max , the J values calculated assuming no crack growth (J NC ) were slightly higher than that calculated based on Δ a measured using DCPD technique (JDCPD). In general, error in the J calculation found to be increased exponentially with Δ a . The error in J max calculation was increased with an increase in Δ a and a decrease in J max . Based on deformation theory of J, an analytic criterion was developed to check the insensitivity of the J max to error in Δ a . There was very good linear relation was found between the J max calculated based on Δ a measured using DCPD technique and the J max calculated assuming no crack growth. This relation will be very useful to calculate J max without measuring the crack growth during fracture test especially for irradiated material. (author)

  12. Floor interaction

    DEFF Research Database (Denmark)

    Petersen, Marianne Graves; Krogh, Peter; Ludvigsen, Martin

    2005-01-01

    Within architecture, there is a long tradition of careful design of floors. The design has been concerned with both decorating floors and designing floors to carry information. Ubiquitous computing technology offers new opportunities for designing interactive floors. This paper presents three...... different interactive floor concepts. Through an urban perspective it draws upon the experiences of floors in architecture, and provides a set of design issues for designing interactive floors....

  13. The Cause of an Eddy Current Signal Noise from a Steam Generator Tube and its Effect on the Detectability of a Crack

    International Nuclear Information System (INIS)

    Lee, Deok Hyun; Choi, Myung Sik; Hur, Do Haeng; Kim, Kyung Mo; Han, Jung Ho

    2008-01-01

    An eddy current inspection has been applied for a pre-service and in-service examination of a steam generator in nuclear power plants. The experience from the inspection of steam generators showed that many plants had an excessive number of tubes with eddy current noise signals over several hundreds, which originated from manufacturing anomalies. The plants in U.S suffered significant downstream inspection costs, history reviews, and diagnostic testing because some signals resembled flaws and others masked a flaw. These lessens learned resulted in issuing the guidelines for steam generator tubing specifications and repair, in order to reduce the number of anomalous signals in the tubes and also to provide the requirement of a signal to noise ratio by applying a field type examination with bobbin coil eddy current probes at a manufacturing process. Besides the noise signals of a bobbin coil eddy current probe from manufacturing anomalies, the excessive background noise of the rotating coil eddy current probe signal is frequently observed from a tube and it negatively affects the detection and sizing estimate of a defect. Since the inspection intervals are being extended up to 60 months for the more recent steam generator of corrosion resistant alloy 690TT tubing, the detection of an earlier crack and an accurate sizing are becoming more important in the activity of a non-destructive examination. In this study, the cause of an eddy current signal noise of a rotating coil probe from a steam generator tube was examined and its influence on the detectability of a crack was analyzed

  14. Evaluation of the crack initiation of curved compact tension specimens of a Zr-2.5Nb pressure tube using the unloading compliance and direct current potential drop methods

    International Nuclear Information System (INIS)

    Jeong, Hyeon Cheol; Ahn, Sang Bok; Park, Joong Chul; Kim, Young Suk

    2005-01-01

    Zr-2.5Nb pressure tubes, carrying fuel bundles and heavy water coolant inside, degrade due to neutron irradiation and hydrogen embrittlement during their operation in heavy water reactors. The safety criterion for the Zr-2.5Nb tubes to meet is a leak-before-break (LBB) requirement. To evaluate a safety margin related to the LBB criterion, facture toughness of the pressure tubes are to be determined periodically with their operational time. For a reliable evaluation of the LBB safety criterion of the pressure tubes, it is required to precisely determine their fracture toughness. Since the fracture toughness or J of the pressure tubes is determined only by the extended crack length, it is important to reliably and precisely evaluate the advanced crack length. However, the problem lies with the detection of the crack opening point because prior plastic deformation before a start of the crack makes it difficult. The aim of this work is to evaluate which method can define the crack initiation point in the Zr- 2.5Nb compact tension specimens more precisely between the unloading compliance method with a crack opening displacement (COD) gauge and the direct current potential drop (DCPD) methods

  15. Stress corrosion crack initiation of Zircaloy-4 cladding tubes in an iodine vapor environment during creep, relaxation, and constant strain rate tests

    Science.gov (United States)

    Jezequel, T.; Auzoux, Q.; Le Boulch, D.; Bono, M.; Andrieu, E.; Blanc, C.; Chabretou, V.; Mozzani, N.; Rautenberg, M.

    2018-02-01

    During accidental power transient conditions with Pellet Cladding Interaction (PCI), the synergistic effect of the stress and strain imposed on the cladding by thermal expansion of the fuel, and corrosion by iodine released as a fission product, may lead to cladding failure by Stress Corrosion Cracking (SCC). In this study, internal pressure tests were conducted on unirradiated cold-worked stress-relieved Zircaloy-4 cladding tubes in an iodine vapor environment. The goal was to investigate the influence of loading type (constant pressure tests, constant circumferential strain rate tests, or constant circumferential strain tests) and test temperature (320, 350, or 380 °C) on iodine-induced stress corrosion cracking (I-SCC). The experimental results obtained with different loading types were consistent with each other. The apparent threshold hoop stress for I-SCC was found to be independent of the test temperature. SEM micrographs of the tested samples showed many pits distributed over the inner surface, which tended to coalesce into large pits in which a microcrack could initiate. A model for the time-to-failure of a cladding tube was developed using finite element simulations of the viscoplastic mechanical behavior of the material and a modified Kachanov's damage growth model. The times-to-failure predicted by this model are consistent with the experimental data.

  16. Determination of He and D permeability of neutron-irradiated SiC tubes to examine the potential for release due to micro-cracking

    Energy Technology Data Exchange (ETDEWEB)

    Katoh, Yutai [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hu, Xunxiang [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Koyanagi, Takaaki [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Singh, Gyanender P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-07-01

    Driven by the need to enlarge the safety margins of light water reactors in both design-basis and beyond-design-basis accident scenarios, the research and development of accident-tolerant fuel (ATF) has become an importance topic in the nuclear engineering and materials community. Continuous SiC fiber-reinforced SiC matrix ceramic composites are under consideration as a replacement for traditional zirconium alloy cladding owing to their high-temperature stability, chemical inertness, and exceptional irradiation resistance. Among the key technical feasibility issues, potential failure of the fission product containment due to probabilistic penetrating cracking has been identified as one of the two most critical feasibility issues, together with the radiolysisassisted hydrothermal corrosion of SiC. The experimental capability to evaluate the hermeticity of SiC-based claddings is an urgent need. In this report, we present the development of a comprehensive permeation testing station established in the Low Activation Materials Development and Analysis laboratory at Oak Ridge National Laboratory. Preliminary results for the hermeticity evaluation of un-irradiated monolithic SiC tubes, uncoated and coated SiC/SiC composite tubes, and neutron-irradiated monolithic SiC tubes at room temperature are exhibited. The results indicate that this new permeation testing station is capable of evaluating the hermeticity of SiC-based tubes by determining the helium and deuterium permeation flux as a function of gas pressure at a high resolution of 8.07 x 10-12 atm-cc/s for helium and 2.83 x 10-12 atm-cc/s for deuterium, respectively. The detection limit of this system is sufficient to evaluate the maximum allowable helium leakage rate of lab-scale tubular samples, which is linearly extrapolated from the evaluation standard used for a commercial as-manufactured light water reactor fuel rod at room temperature. The un-irradiated monolithic SiC tube is hermetic, as

  17. A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Kwang Sik

    1992-02-15

    Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values.

  18. A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Choi, Kwang Sik

    1992-02-01

    Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values

  19. The role of time-dependent deformation in intergranular crack initiation of alloy 600 steam generator tubing material

    International Nuclear Information System (INIS)

    Was, G.S.; Lian, K.

    1998-03-01

    Intergranular stress corrosion cracking (IGSCC) of two commercial alloy 600 conditions (600LT, 600HT) and controlled- purity Ni-18Cr-9Fe alloys (CDMA, CDTT) were investigated using constant extension rate tensile (CERT) tests in primary water (0.01M LiOH+0.01M H 3 BO 3 ) with 1 bar hydrogen overpressure at 360 degrees C and 320 degrees C. Heat treatments produced two types of microstructures in both commercial and controlled-purity alloys: one dominated by grain boundary carbides (600HT and CDTT) and one dominated by intragranular carbides (600LT and CDMA). CERT tests were conducted over a range of strain rates and at two temperatures with interruptions at specific strains to determine the crack depth distributions. Results show that in all samples, IGSCC was the dominant failure mode. For both the commercial alloy and the controlled-purity alloys, the microstructure with grain boundary carbides showed delayed crack initiation and shallower crack depths than did the intragranular carbide microstructure under all experimental conditions. This data indicates that a grain boundary carbide microstructure is more resistant to IGSCC than an intragranular carbide microstructure. Observations support both the film rupture/slip dissolution mechanism and enhanced localized plasticity. The advantage of these results over previous studies is that the different carbide distributions were obtained in the same commercial alloy using different heat treatments, and in the other case, in nearly identical controlled-purity alloys. Therefore, observations of the effects of carbide distribution on IGSCC can more confidently be attributed to the carbide distribution alone rather than other potentially significant differences in microstructure or composition

  20. 基于微观裂隙扩张的采场底板突水机理研究%Mechanism study of floor water inrush around mining field based on micro-crack extension

    Institute of Scientific and Technical Information of China (English)

    高玉兵; 刘世奇; 吕斌; 李昆奇

    2016-01-01

    基于矿山压力和岩层控制理论,分析了工作面附近支承压力及底板水压的空间分布特征,确定了两者对底板的力学作用机制。通过建立裂隙力学模型,从微观角度研究了支承压力和水压对裂隙的作用效果;以薄板理论为依据,从宏观角度研究了含水层水压对底板有效隔水层的作用机理,并给出底板突水极限水压值的计算公式。结果表明:支承压力的增大会导致裂隙渗透系数的减小;水压的增大会导致岩体强度的降低和裂隙劈裂长度的增加。采场底板突水的实质是采动引起的矿山压力和承压水水压共同作用下微观裂隙的扩张和底板有效隔水层的断裂。%Based on the mine ground pressure and strata control theory, spatial distribution characteris-tics of ground pressure and floor water pressure around mining field have been analyzed in this paper, and then the mechanical effects of both these pressures on the coal floor have been determined. By set-ting up mechanics model, the effects of ground pressure and water pressure on a set of tilted cracks have been studied from a microcosmic point of view. Based on the thin plate theory, the effects of water pressure on the effective water-resisting layer have also been studied from a macroscopic point of view. Finally, a theoretical formulae of limiting hydraulic pressure has been deduced. The results have shown that the increase of abutment pressure can lead to the decrease of permeability coefficient; the increase of water pressure can cause the attenuation of rock mass strength and the increase of fissure length. The essence of water inrush in mining field is the result of propagation of micro-cracks and fracture damage of the effective water-resisting layer caused by ground pressure and water pressure.

  1. Cracking and corrosion recovery boiler

    Energy Technology Data Exchange (ETDEWEB)

    Suik, H. [Tallinn Technical University, Horizon Pulp and Paper, Tallinn (Estonia)

    1998-12-31

    The corrosion of heat surfaces and the cracking the drums are the main problems of the recovery boiler. These phenomena have been appeared during long-term operation of boiler `Mitsubishi - 315` erected at 1964. Depth of the crack is depending on the number of shutdowns and on operation time. Corrosion intensity of different heat surfaces is varying depend on the metal temperature and the conditions at place of positioning of tube. The lowest intensity of corrosion is on the bank tubes and the greatest is on the tubes of the second stage superheater and on the tubes at the openings of air ports. (orig.) 5 refs.

  2. Cracking and corrosion recovery boiler

    Energy Technology Data Exchange (ETDEWEB)

    Suik, H [Tallinn Technical University, Horizon Pulp and Paper, Tallinn (Estonia)

    1999-12-31

    The corrosion of heat surfaces and the cracking the drums are the main problems of the recovery boiler. These phenomena have been appeared during long-term operation of boiler `Mitsubishi - 315` erected at 1964. Depth of the crack is depending on the number of shutdowns and on operation time. Corrosion intensity of different heat surfaces is varying depend on the metal temperature and the conditions at place of positioning of tube. The lowest intensity of corrosion is on the bank tubes and the greatest is on the tubes of the second stage superheater and on the tubes at the openings of air ports. (orig.) 5 refs.

  3. Floors: Selection and Maintenance.

    Science.gov (United States)

    Berkeley, Bernard

    Flooring for institutional, commercial, and industrial use is described with regard to its selection, care, and maintenance. The following flooring and subflooring material categories are discussed--(1) resilient floor coverings, (2) carpeting, (3) masonry floors, (4) wood floors, and (5) "formed-in-place floors". The properties, problems,…

  4. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1984-10-01

    A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization

  5. Delayed hydride cracking in zirconium alloys in pressure tube nuclear reactors. Final report of a coordinated research project 1998-2002

    International Nuclear Information System (INIS)

    2004-10-01

    This report describes all of the research work undertaken as part of the IAEA coordinated research project on hydrogen and hydride induced degradation of the mechanical and physical properties of zirconium based alloys, and includes a review of the state of the art in understanding crack propagation by Delayed Hydride Cracking (DHC), and details of the experimental procedures that have produced the most consistent set of DHC rates reported in an international round-robin exercise to this date. It was concluded that 1) the techniques for performing measurements of the rate of delayed hydride cracking in zirconium alloys have been transferred from the host laboratory to other countries; 2) by following a strict procedure, a very consistent set of values of crack velocity were obtained by both individual laboratories and between the different laboratories; 3) the results over a wide range of test temperatures from materials with various microstructures fitted into the current theoretical framework for delayed hydride cracking; 4) an inter-laboratory comparison of hydrogen analysis revealed the importance of calibration and led to improvements in measurement in the participating laboratories and 5) the success of the CRP in achieving its goals has led to the initiation of some national programmes

  6. Strain rate and temperature effects on the stress corrosion cracking of Inconel 600 steam generator tubing in the primary water conditions

    International Nuclear Information System (INIS)

    Kim, U.C.; van Rooyen, D.

    1985-01-01

    A single heat of Inconel Alloy 600 was examined in this work, using slow strain rate tests (SSRT) in simulated primary water at temperatures of 325 0 -345 0 -365 0 C. The best measure of stress corrosion cracking (SCC) was percent SCC present on the fracture surface. Strain rate did not seem to affect crack growth rate significantly, but there is some question about the accuracy of calculating these values in the absence of a direct indication of when a crack initiates. Demarcation was determined between domains of temperature/strain rate where SCC either did, or did not, occur. Slower extension rates were needed to produce SCC as the temperature was lowered. 10 figs

  7. Corrosion cracking

    International Nuclear Information System (INIS)

    Goel, V.S.

    1985-01-01

    This book presents the papers given at a conference on alloy corrosion cracking. Topics considered at the conference included the effect of niobium addition on intergranular stress corrosion cracking, corrosion-fatigue cracking in fossil-fueled-boilers, fracture toughness, fracture modes, hydrogen-induced thresholds, electrochemical and hydrogen permeation studies, the effect of seawater on fatigue crack propagation of wells for offshore structures, the corrosion fatigue of carbon steels in seawater, and stress corrosion cracking and the mechanical strength of alloy 600

  8. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1983-08-01

    A review of the performance of steam generator tubes in 110 water-cooled nuclear power reactors showed that tubes were plugged at 46 (42 percent) of the reactors. The number of tubes removed from service increased from 1900 (0.14 percent) in 1980 to 4692 (0.30 percent) in 1981. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that used all-volatile treatment since start-up. At one reactor a large number of degraded tubes were repaired by sleeving which is expected to become an important method of tube repair in the future

  9. As the crack in the Geiger counter came. Historical scientific analysis and didactic aspects of the Geiger-Mueller counting tube

    International Nuclear Information System (INIS)

    Korff, Sebastian

    2014-01-01

    This thesis studies the creation and establishment history of this instrument called first electron counting tube in the years 1928 and 1929. It deals thereby with the last two years of the common work of Hans Geiger and Walter Mueller, from which the measuring instrument later renamed to Geiger-Mueller counting tube. The results of this scientific case study are didactically worked out and made usable for the teaching of physics in the school.

  10. Creep-fatigue crack initiation assessment on thick circumferentially notched 316L tubes under cyclic thermal shocks and uniform tension with the σd approach

    International Nuclear Information System (INIS)

    Michel, B.; Poette, C.

    1997-01-01

    For crack initiation assessment under creep fatigue loading, in high temperature Fast Reactor's components, specific approaches based on fracture mechanics analysis had to be developed. In the present paper the crack initiation assessment method proposed in the A16 document is presented. The so called ''σ d method'' is also validated on experimental results for tubular specimens with internal axisymmetric surface cracks. Experimental data are extracted from the TERFIS program carried out on a sodium test device at the CEA Cadarache. Metallurgical examinations on TERFIS specimens confirm that the initiation assessment of the ''σ d '' approach is conservative even for a different geometry than the CT specimen on which the method was set up. However, the conservatism is reduced when the creep residual stress field is relaxed during the hold time. An investigation concerning this last point is needed in order to know if relaxing the stress, when using a lower bound of the mechanical properties, always keeps a safety margin. (author). 14 refs, 10 figs, 4 tabs

  11. Delayed hydride cracking: alternative pre-cracking method

    International Nuclear Information System (INIS)

    Mieza, Juan I.; Ponzoni, Lucio M.E.; Vigna, Gustavo L.; Domizzi, Gladys

    2009-01-01

    The internal components of nuclear reactors built-in Zr alloys are prone to a failure mechanism known as Delayed Hydride Cracking (DHC). This situation has triggered numerous scientific studies in order to measure the crack propagation velocity and the threshold stress intensity factor associated to DHC. Tests are carried out on fatigued pre-crack samples to ensure similar test conditions and comparable results. Due to difficulties in implementing the fatigue pre-crack method it would be desirable to replace it with a pre-crack produced by the same process of DHC, for which is necessary to demonstrate equivalence of this two methods. In this work tests on samples extracted from two Zr-2.5 Nb tubes were conducted. Some of the samples were heat treated to obtain a range in their metallurgical properties as well as different DHC velocities. A comparison between velocities measured in test samples pre-cracked by fatigue and RDIH is done, demonstrating that the pre-cracking method does not affect the measured velocity value. In addition, the incubation (t inc ), which is the time between the application of the load and the first signal of crack propagation, in samples pre-cracked by RDIH, was measured. It was found that these times are sufficiently short, even in the worst cases (lower speed) and similar to the ones of fatigued pre-cracked samples. (author)

  12. Deflection of resilient materials for reduction of floor impact sound.

    Science.gov (United States)

    Lee, Jung-Yoon; Kim, Jong-Mun

    2014-01-01

    Recently, many residents living in apartment buildings in Korea have been bothered by noise coming from the houses above. In order to reduce noise pollution, communities are increasingly imposing bylaws, including the limitation of floor impact sound, minimum thickness of floors, and floor soundproofing solutions. This research effort focused specifically on the deflection of resilient materials in the floor sound insulation systems of apartment houses. The experimental program involved conducting twenty-seven material tests and ten sound insulation floating concrete floor specimens. Two main parameters were considered in the experimental investigation: the seven types of resilient materials and the location of the loading point. The structural behavior of sound insulation floor floating was predicted using the Winkler method. The experimental and analytical results indicated that the cracking strength of the floating concrete floor significantly increased with increasing the tangent modulus of resilient material. The deflection of the floating concrete floor loaded at the side of the specimen was much greater than that of the floating concrete floor loaded at the center of the specimen. The Winkler model considering the effect of modulus of resilient materials was able to accurately predict the cracking strength of the floating concrete floor.

  13. Detection and mode identification of axial cracks in the steam generator tube of the nuclear power plant using ultrasonic guided wave

    International Nuclear Information System (INIS)

    Yoon, Byungsik; Yang, Seunghan; Lee, Heejong; Kim, Yongsik

    2010-01-01

    For those people who are involved in NDE, there is a growing concern regarding the significant traveling distance of a guided wave in a structure, which ensures the inspection of a large area of the structure from a single location. A significant number of studies on the guided wave have therefore been made to apply the foregoing to a nondestructive evaluation in many different industries and resulted in an increase in the efficiency of practical guided wave inspection. Unlike the previous studies based mainly on the detection of circumferential flaws, this study is focused on the axial flaw detection in the steam generator tubes of Korean standard nuclear power plants by generating the guided wave by changing frequency and selecting the applicable mode from the dispersion curve for the steam generator tube calculated in this study, where the dispersion-based short-time Fourier transform (D-STFT) algorithm is used to enhance mode identification. In conclusion, the L (0,1) mode at 2.25 MHz is found to be most sensitive in detecting axial flaws in a steam generator tube. (author)

  14. 27 CFR 46.195 - Floor stocks requirements.

    Science.gov (United States)

    2010-04-01

    ... Tubes Held for Sale on April 1, 2009 General § 46.195 Floor stocks requirements. (a) Take inventory. The dealer must establish the quantity of articles subject to the floor stocks tax held for sale on April 1, 2009. The dealer may take a physical inventory or may use a record (book) inventory, as specified in...

  15. Effects of dissolved calcium and magnesium ions on lead-induced stress corrosion cracking susceptibility of nuclear steam generator tubing alloy in high temperature crevice solutions

    International Nuclear Information System (INIS)

    Lu, B.T.; Tian, L.P.; Zhu, R.K.; Luo, J.L.; Lu, Y.C.

    2011-01-01

    The effects of Ca 2+ and Mg 2+ ions on the stress corrosion cracking (SCC) susceptibility of UNS N08800 are investigated using constant extension rate tensile (CERT) tests at 300 o C in simulated crevice chemistries. The presence of lead contamination in the crevice chemistries increases significantly the SCC susceptibility of the alloy. The lead-assisted SCC (PbSCC) susceptibility is reduced markedly by the addition of Ca 2+ and Mg 2+ ions into the solution and this mitigating effect is enhanced by increasing the total concentration of Ca 2+ + Mg 2+ . The CERT test results are consistent with the types of fracture surfaces shown by Scanning Electron Microscopy (SEM). There is a reasonable correlation between the SCC susceptibility and the donor densities in the anodic films in accord with the role of lead-induced passivity degradation in PbSCC.

  16. Stress corrosion cracking experience in steam generators at Bruce NGS

    International Nuclear Information System (INIS)

    King, P.J.; Gonzalez, F.; Brown, J.

    1993-01-01

    In late 1990 and through 1991, units 1 and 2 at the Bruce A Nuclear Generating Station (BNGS-A) experienced a number of steam generator tube leaks. Tube failures were identified by eddy current to be circumferential cracks at U-bend supports on the hot-leg side of the boilers. In late 1991, tubes were removed from these units for failure characterization. Two active failure modes were found: corrosion fatigue in both units 1 and 2 and stress corrosion cracking (SCC) in unit 2. In unit 2, lead was found in deposits, on tubes, and in cracks, and the cracking was mixed-mode: transgranular and intergranular. This convincingly indicated the involvement of lead in the stress corrosion cracking failures. A program of inspection and tube removals was carried out to investigate more fully the extent of the problem. This program found significant cracking only in lead-affected boilers in unit 2, and also revealed a limited extent of non-lead-related intergranular stress corrosion cracking in other boilers and units. Various aspects of the failures and tube examinations are presented in this paper. Included is discussion of the cracking morphology, measured crack size distributions, and chemical analysis of tube surfaces, crack faces, and deposits -- with particular emphasis on lead

  17. SUSTAINABLE TRAILER FLOORING

    OpenAIRE

    John Lu; Marc Chorney; Lowell Peterson

    2009-01-01

    Different trailer flooring materials, including wood-based, aluminum, steel, and synthetic plastic floors, were evaluated in accordance with their durability and sustainability to our natural environment. Wood-based trailer flooring is an eco-friendly product. It is the most sustainable trailer flooring material compared with fossil fuel-intensive steel, aluminum, and plastics. It is renewable and recyclable. Oak, hard maple, and apitong are strong and durable hardwood species that are curren...

  18. Decontamination of floor surfaces

    International Nuclear Information System (INIS)

    Smirous, F.

    1983-01-01

    Requirements are presented put on the surfaces of floors of radiochemical workplaces. The mechanism is described of retaining the contaminant in the surface of the flooring, ways of reducing the hazards of floor surface contamination, decontamination techniques and used decontamination agents. (J.P.)

  19. Numerical analysis of interacting cracks in biaxial stress field

    International Nuclear Information System (INIS)

    Kovac, M.; Cizelj, L.

    1999-01-01

    The stress corrosion cracks as seen for example in PWR steam generator tubing made of Inconel 600 usually produce highly irregular kinked and branched crack patterns. Crack initialization and propagation depends on stress state underlying the crack pattern. Numerical analysis (such as finite element method) of interacting kinked and branched cracks can provide accurate solutions. This paper discusses the use of general-purpose finite element code ABAQUS for evaluating stress fields at crack tips of interacting complex cracks. The results obtained showed reasonable agreement with the reference solutions and confirmed use of finite elements in such class of problems.(author)

  20. SUSTAINABLE TRAILER FLOORING

    Directory of Open Access Journals (Sweden)

    John Lu

    2009-05-01

    Full Text Available Different trailer flooring materials, including wood-based, aluminum, steel, and synthetic plastic floors, were evaluated in accordance with their durability and sustainability to our natural environment. Wood-based trailer flooring is an eco-friendly product. It is the most sustainable trailer flooring material compared with fossil fuel-intensive steel, aluminum, and plastics. It is renewable and recyclable. Oak, hard maple, and apitong are strong and durable hardwood species that are currently extensively used for trailer flooring. For manufacture, wood-based flooring is higher in energy efficiency and lower in carbon emission than steel, aluminum and plastics. Moreover, wood per se is a natural product that sequesters carbon. Accordingly, using more wood-based trailer flooring is effective to reduce global warming.

  1. Influence of precipitation behavior on mechanical properties and hydrogen induced cracking during tempering of hot-rolled API steel for tubing

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Joonoh, E-mail: mjo99@kims.re.kr [Ferrous Alloy Department, Advanced Metallic Materials Division, Korea Institute of Materials Science, 797 Changwondae-ro, Seongsan-gu, Changwon, Gyeongnam 642-831 (Korea, Republic of); Choi, Jongmin; Han, Seong-Kyung; Huh, Sungyul; Kim, Seong-Ju [Sheet Products Design Team, Technical Research Center, Hyundai Steel Company, 1480 Bukbusaneop-ro, Dangjin, Chungnam 343-823 (Korea, Republic of); Lee, Chang-Hoon; Lee, Tae-Ho [Ferrous Alloy Department, Advanced Metallic Materials Division, Korea Institute of Materials Science, 797 Changwondae-ro, Seongsan-gu, Changwon, Gyeongnam 642-831 (Korea, Republic of)

    2016-01-15

    Precipitation behavior and its effect on hydrogen embrittlement during tempering process of hot-rolled API steel designed with 0.4 wt% Cr and 0.25 wt% Mo were investigated. The base steel was normalized and then tempered at 650 °C for up to 60 min. The precipitation behavior of the examined steel was explored using transmission electron microscopy (TEM) analysis, and it was found that the precipitation sequence during tempering at 650 °C were as follows: MX+M{sub 3}C→MX→MX+M{sub 7}C{sub 3}+M{sub 23}C{sub 6}. The change of particle fraction was measured by electrolytic extraction technique. At the early stage of tempering, the particle fraction greatly decreased due to dissolution of M{sub 3}C particle, and increased after 10 min by the precipitation of M{sub 7}C{sub 3} and M{sub 23}C{sub 6} particles. The particle fraction showed a peak at 30 min tempering and decreased again due to the dissolution of M{sub 7}C{sub 3} particle. Vickers hardness tests of base steel and tempered samples were carried out, and then the hardness was changed by accompanying with the change of particle fraction. The sensitivity of hydrogen embrittlement was evaluated through hydrogen induced cracking (HIC) tests, and the results clearly proved that HIC resistance of tempered samples was better than that of base steel due to the formation of tempered martensite, and then the HIC resistance changed depending on the precipitation behavior during tempering, i.e., the precipitation of coarse M{sub 23}C{sub 6} and M{sub 7}C{sub 3} particles deteriorated the HIC resistance.

  2. Influence of precipitation behavior on mechanical properties and hydrogen induced cracking during tempering of hot-rolled API steel for tubing

    International Nuclear Information System (INIS)

    Moon, Joonoh; Choi, Jongmin; Han, Seong-Kyung; Huh, Sungyul; Kim, Seong-Ju; Lee, Chang-Hoon; Lee, Tae-Ho

    2016-01-01

    Precipitation behavior and its effect on hydrogen embrittlement during tempering process of hot-rolled API steel designed with 0.4 wt% Cr and 0.25 wt% Mo were investigated. The base steel was normalized and then tempered at 650 °C for up to 60 min. The precipitation behavior of the examined steel was explored using transmission electron microscopy (TEM) analysis, and it was found that the precipitation sequence during tempering at 650 °C were as follows: MX+M_3C→MX→MX+M_7C_3+M_2_3C_6. The change of particle fraction was measured by electrolytic extraction technique. At the early stage of tempering, the particle fraction greatly decreased due to dissolution of M_3C particle, and increased after 10 min by the precipitation of M_7C_3 and M_2_3C_6 particles. The particle fraction showed a peak at 30 min tempering and decreased again due to the dissolution of M_7C_3 particle. Vickers hardness tests of base steel and tempered samples were carried out, and then the hardness was changed by accompanying with the change of particle fraction. The sensitivity of hydrogen embrittlement was evaluated through hydrogen induced cracking (HIC) tests, and the results clearly proved that HIC resistance of tempered samples was better than that of base steel due to the formation of tempered martensite, and then the HIC resistance changed depending on the precipitation behavior during tempering, i.e., the precipitation of coarse M_2_3C_6 and M_7C_3 particles deteriorated the HIC resistance.

  3. Heating tubes of cross-linked polyethylene

    International Nuclear Information System (INIS)

    Knoeppler, H.; Hoffmann, M.

    1981-01-01

    Oxygen permeability of plastic tubes for floor heating systems was measured as a function of the reduced oxygen content of water in plastic tubes at a flow rate of 0.5 m/s and a temperature of 30 0 C and as a function of oxygen uptake of low-oxygen water in floor heating tubes. Pipes of VEP, periodically cross-linked polyethylene (Engels process), polypropylene copolymeride, and polybutene were compared. The permeability of periodically cross-linked polyethylene is twice as high as that of VEP. Measurements, results, and consequences for floor heating systems are discussed. (KH) [de

  4. Preliminary research on eddy current bobbin quantitative test for heat exchange tube in nuclear power plant

    Science.gov (United States)

    Qi, Pan; Shao, Wenbin; Liao, Shusheng

    2016-02-01

    For quantitative defects detection research on heat transfer tube in nuclear power plants (NPP), two parts of work are carried out based on the crack as the main research objects. (1) Production optimization of calibration tube. Firstly, ASME, RSEM and homemade crack calibration tubes are applied to quantitatively analyze the defects depth on other designed crack test tubes, and then the judgment with quantitative results under crack calibration tube with more accuracy is given. Base on that, weight analysis of influence factors for crack depth quantitative test such as crack orientation, length, volume and so on can be undertaken, which will optimize manufacture technology of calibration tubes. (2) Quantitative optimization of crack depth. Neural network model with multi-calibration curve adopted to optimize natural crack test depth generated in in-service tubes shows preliminary ability to improve quantitative accuracy.

  5. Supporting shop floor intelligence

    DEFF Research Database (Denmark)

    Carstensen, Peter; Schmidt, Kjeld; Wiil, Uffe Kock

    1999-01-01

    Many manufacturing enterprises are now trying to introduce various forms of flexible work organizations on the shop floor. However, existing computer-based production planning and control systems pose severe obstacles for autonomous working groups and other kinds of shop floor control to become r......-to-day production planning by supporting intelligent and responsible workers in their situated coordination activities on the shop floor....

  6. Introductory guide to floors and flooring

    CSIR Research Space (South Africa)

    Billingham, PA

    1977-01-01

    Full Text Available not make use of the warming and cooling effects of direct contact with the ground. Indeed the precautions that are necessary to protect such floors against damp and decay may actually reduce the comfort levels within a house. This is because there is a... with resultant discomfort and extra heating costs. Today, in South Africa, most modern homesareof singlestorey con- struction with aconcrete floor slab in direct contact with theground which once again makes its full contribution to the comfort and structural...

  7. A new method for controlling of floor heave of deep tunnels in soft rocks by mini-tube grouting piles of crushed stones%微型碎石管注桩治理深部软岩巷道底鼓新方法

    Institute of Scientific and Technical Information of China (English)

    谢芳; 王金安

    2013-01-01

    针对造成巷道底鼓的物理和力学两种重要机制,研究并提出微型碎石管注桩治理深部软岩巷道底鼓的新方法.在物理机制方面,微型碎石桩具有渗透性极好的特点,在巷道施工和使用期间可汲取岩体中的渗水并透过碎石桩中的插管排出,从而降低软岩层因遇水膨胀导致的变形;在力学机制方面,微型碎石桩一方面通过钻孔置换出少部分底板软岩,从体量上减少变形岩体,另一方面能利用碎石桩体的侧向可压缩性耗散岩体水平变形,底板岩层中的水平地应力得以释放,极大减缓了促使底鼓变形的力学作用;通过微型碎石桩中的插管注浆加固底板岩体,提高了底板复合地基整体承载力.通过数值分析,阐明该方法在治理软岩巷道底鼓机制上的有效性.%Focusing on two important mechanisms of physics and mechanics that give rise to the floor heave of tunnels , a new method of controlling the floor heave of deep tunnels in soft rocks was proposed by means of mini-tube grouting piles of crushed stones. In physical aspect, the method utilizes the characteristics of good permeability of crushed stone pile to absorb the water in rockmass and drainage out through the tubes in the pile during the construction and application of the tunnels, resulting in the decrease in expansion deformation of soft rock due to water saturation. In mechanical aspect, partial volume in floor stratum has been replaced, on one hand, by the installed mini piles which reduce the volume of the deforming rock body, and on the other hand, amount of horizontal deformation are dissipated through the lateral compressive deformation of the crushed stone pile, and the horizontal stress is released. As a result, the mechanical affect that induces the heave deformation in floor stratum is considerably reduced. By means of grouting the crushed stones through the tube inserted in the pile, the floor stratum is reinforced and the

  8. Maxillary Sinus Floor Augmentation

    DEFF Research Database (Denmark)

    Starch-Jensen, Thomas; Jensen, Janek Dalsgaard

    2017-01-01

    , radiological and histomorphometric outcome as well as complications are presented after maxillary sinus floor augmentation applying the lateral window technique with a graft material, maxillary sinus membrane elevation without a graft material and osteotome-mediated sinus floor elevation with or without...

  9. Floors: Care and Maintenance.

    Science.gov (United States)

    Post Office Dept., Washington, DC.

    Guidelines, methods and policies regarding the care and maintenance of post office building floors are overviewed in this handbook. Procedures outlined are concerned with maintaining a required level of appearance without wasting manpower. Flooring types and characteristics and the particular cleaning requirements of each type are given along with…

  10. School Flooring Factors

    Science.gov (United States)

    McGrath, John

    2012-01-01

    With all of the hype that green building is receiving throughout the school facility-management industry, it's easy to overlook some elements that may not be right in front of a building manager's nose. It is helpful to examine the role floor covering plays in a green building project. Flooring is one of the most significant and important systems…

  11. Solving decontaminable flooring problems

    International Nuclear Information System (INIS)

    Anon.

    1989-01-01

    Pennsylvania Power and Light wanted to cover deteriorating concrete in unit 2 of its Susquehanna BWR with a smooth, durable, decontaminable coating. Traditionally, floors in the plant had been coated with epoxy paint, but many of these floors suffered delamination, and failed in three to five years. Painting with epoxy would also interrupt operations for as much as three days while the floor dried, yet critical instruments in some areas had to be monitored at least once per shift. In addition, conventional floor surface preparation produced dust and vibration around sensitive equipment. The solution was a dustless scabbling system for surface preparation, followed by the installation of a high-strength acrylic industrial floor known as Silakal. The work was carried out by Pentek. Silikal bonds to the underlying concrete, so that delamination of the floor will not occur even under severe traffic conditions. Another advantage of this type of flooring is that it cures in one hour, so floor resurfacing has only minimal impact on plant operations. (author)

  12. Floor heating systems

    Energy Technology Data Exchange (ETDEWEB)

    Radtke, U

    1984-02-01

    The question of whether PPC- and VPE-floor heating pipes can endure damage when incompletely imbedded in the floor finish is investigated in an experimental setup. An expansion of the pipe, caused by a temperature increase from 20/sup 0/C to 50/sup 0/C was measured and considered too small to deduce the degree of danger from the damage.

  13. Perceived floor slipperiness and floor roughness in a gait experiment.

    Science.gov (United States)

    Yu, Ruifeng; Li, Kai Way

    2015-01-01

    Slips and falls contribute to occupational injuries and fatalities globally. Both floor slipperiness and floor roughness affect the occurrence of slipping and falling. Investigations on fall-related phenomena are important for the safety and health of workers. The purposes of this study were to: compare the perceived floor slipperiness before and after walking on the floor; compare the perceived floor slipperiness with and without shoes for males and females; discuss the perceived floor roughness based on barefoot walking; and establish regression models to describe the relationship between perceived floor slipperiness and actual friction of the floors. Male and female subjects walked on 3 m walkways with or without shoes. The perceived floor slipperiness ratings both before and after their walk were collected. The perceived floor slipperiness both before and after walking were significantly affected by both floor and surface conditions. Gender, floor, surface, and footwear conditions were all significant factors affecting the adjustment of perceived floor slipperiness. The subjects made more adjustment on perceived floor slipperiness rating when they had shoes on than when they were barefooted. Regression models were established to describe the relationship between perceived floor slipperiness and floor coefficient of friction. These models may be used to estimate perceived floor slipperiness, or in reverse, the coefficient of friction of the floor, so as to prevent slipping and falling in workplaces.

  14. Cracks in Utopia

    Science.gov (United States)

    1999-01-01

    Many of the craters found on the northern plains of Mars have been partly filled or buried by some material (possibly sediment). The Mars Global Surveyor (MGS) Mars Orbiter Camera (MOC) image presented here (MOC2-136b, above left) shows a high-resolution view of a tiny portion of the floor of one of these northern plains craters. The crater, located in Utopia Planitia at 44oN, 258oW, is shown on the right (MOC2-136a)with a small white box to indicate the location of the MOC image. The MOC image reveals that the material covering the floor of this crater is cracked and pitted. The origin and source of material that has been deposited in this crater is unknown.The MOC image was acquired in June 1999 and covers an area only 1.1 kilometers (0.7 miles) wide at a resolution of 1.8 meters (6 feet) per pixel. The context picture is a mosaic of Viking 2 orbiter images 010B53 and 010B55, taken in 1976. Both images are illuminated from the left. Malin Space Science Systems and the California Institute of Technology built the MOC using spare hardware from the Mars Observer mission. MSSS operates the camera from its facilities in San Diego, CA. The Jet Propulsion Laboratory's Mars Surveyor Operations Project operates the Mars Global Surveyor spacecraft with its industrial partner, Lockheed Martin Astronautics, from facilities in Pasadena, CA and Denver, CO.

  15. 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

  16. Delayed hydride cracking: theoretical model testing to predict cracking velocity

    International Nuclear Information System (INIS)

    Mieza, Juan I.; Vigna, Gustavo L.; Domizzi, Gladys

    2009-01-01

    Pressure tubes from Candu nuclear reactors as any other component manufactured with Zr alloys are prone to delayed hydride cracking. That is why it is important to be able to predict the cracking velocity during the component lifetime from parameters easy to be measured, such as: hydrogen concentration, mechanical and microstructural properties. Two of the theoretical models reported in literature to calculate the DHC velocity were chosen and combined, and using the appropriate variables allowed a comparison with experimental results of samples from Zr-2.5 Nb tubes with different mechanical and structural properties. In addition, velocities measured by other authors in irradiated materials could be reproduced using the model described above. (author)

  17. Steam generator tube performance

    International Nuclear Information System (INIS)

    Tatone, O.S.; Pathania, R.S.

    1982-04-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1980. Tube defects occurred at 38% of the 97 reactors surveyed. This is a marginal improvement over 1979 when defects occurred at 41% of the reactors. The number of failed tubes was also lower, 0.14% of the tubes in service in 1980 compared with 0.20% of those in service in 1979. Analysis of the causes of these failures indicates that stress corrosion cracking was the leading failure mechanism. Reactors that used all-volatile treatment of secondary water, with or without full-flow condensate demineralization since start-up showed the lowest incidence of corrosion-related defects

  18. Radiant Floor Cooling Systems

    DEFF Research Database (Denmark)

    Olesen, Bjarne W.

    2008-01-01

    In many countries, hydronic radiant floor systems are widely used for heating all types of buildings such as residential, churches, gymnasiums, hospitals, hangars, storage buildings, industrial buildings, and smaller offices. However, few systems are used for cooling.This article describes a floor...... cooling system that includes such considerations as thermal comfort of the occupants, which design parameters will influence the cooling capacity and how the system should be controlled. Examples of applications are presented....

  19. Evaluation of the crack initiation of curved compact tension specimens of a Zr-2.5Nb pressure tube using the unloading compliance and direct current potential drop methods

    International Nuclear Information System (INIS)

    Kim, Young Suk; Jeong, Hyeon Cheol; Ahn, Sang Bok

    2005-01-01

    The Direct Current Potential Drop(DCPD) method and the Unloading Compliance(UC) method with a crack opening displacement gauge were applied simultaneously to the Zr-2.5Nb Curved Compact Tension (CCT) specimens to determine which of the two methods can precisely determine the crack initiation point and hence the crack length for evaluation of their fracture toughness. The DCPD method detected the crack initiation at a smaller load-time displacement compared to the UC method. As a verification, a direct observation of the fracture surfaces on the curved compact tension specimens was made on the CCT specimens experiencing either 0.8 to 1.0 mm load line displacement or various loads from 50% to 80% of the maximum peak load, or P max . The DCPD method is concluded to be more precise in determining the crack initiation and fracture toughness, J in Zr-2.5Nb CCT specimens than the UC method

  20. Password cracking

    OpenAIRE

    Χριστοφάκης, Μιχαήλ Κ.

    2014-01-01

    Information security is the next big thing in computers society because of the rapidly growing security incidents and the outcomes of those. Hacking and cracking existed even from the start of the eighties decade when there was the first step of the interconnection through the internet between humans. From then and ever after there was a big explosion of such incidents mostly because of the worldwide web which was introduced in the early nineties. Following the huge steps forward of computers...

  1. Cracking resistance study of steel for gas pipelines; Badania odpornosci stali przeznaczonej na rurociagi gazowe

    Energy Technology Data Exchange (ETDEWEB)

    Wasiak, J.; Bilous, W.; Hajewska, E.; Szteke, W.; Wagner, T. [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1996-12-31

    The results of cracking resistance of steel tubes for gas pipelines have been performed. The temperature dependence of mechanical properties of X56 steel used as tube material have been shown. 2 refs, 6 figs, 4 tabs.

  2. Stress corrosion crack growth in unirradiated zircaloy

    International Nuclear Information System (INIS)

    Pettersson, K.

    1978-10-01

    Experimental techniques suitable for the determination of stress corrosion crack growth rates in irradiated Zircaloy tube have been developed. The techniques have been tested on unirradiated. Zircaloy and it was found that the results were in good agreement with the results of other investigations. Some of the results were obtained at very low stress intensities and the crack growth rates observed, gave no indication of the existance of a K sub(ISCC) for iodine induced stress corrosion cracking in Zircaloy. This is of importance both for fuel rod behavior after a power ramp and for long term storage of spent Zircaloy-clad fuel. (author)

  3. Square through tube

    International Nuclear Information System (INIS)

    Akita, Junji; Honma, Toei.

    1975-01-01

    Object: To provide a square through tube involving thermal movement in pipelines such as water supply pump driving turbine exhaust pipe (square-shaped), which is wide in freedom with respect to shape and dimension thereof for efficient installation at site. Structure: In a through tube to be airtightly retained for purpose of decontamination in an atomic power plant, comprising a seal rubber plate, a band and a bolt and a nut for securing said plate, the seal rubber plate being worked into the desired shape so that it may be placed in intimate contact with the concrete floor surface by utilization of elasticity of rubber, thereby providing airtightness at a corner portion of the square tube. (Kamimura, M.)

  4. Astrobiology Training in Lava Tubes (ATiLT): Characterizing coralloid speleothems in basaltic lava tubes as a Mars analogue

    Science.gov (United States)

    Ni, J.; Leveille, R. J.; Douglas, P.

    2017-12-01

    Coralloid speleothems or cave corals are small mineralised nodes that can take a variety of forms, and which develop through groundwater seepage and water-rock interaction in caves. They are found commonly on Earth in a plethora of caves, including lava tubes. Since lava tubes have been identified on the surface of Mars from remotely sensed images, there has been interest in studying Earth's lava tube systems as an analogue for understanding Martian lava environments. If cave minerals were found on Mars, they could indicate past or present water-rock interaction in the Martian subsurface. Martian lava tubes could also provide insights into habitable subsurface environments as well as conditions favourable for the synthesis and preservation of biosignatures. One of the aims of the Astrobiology Training in Lava Tubes (ATiLT) project is to analyze biosignatures and paleoenvironmental indicators in secondary cave minerals, which will be looked at in-situ and compared to collected field samples. In this study, secondary mineralization in lava cave systems from Lava Beds National Monument, CA is examined. In the field, coralloid speleothems have been observed growing on all surfaces of the caves, including cave ceilings, floors, walls and overhangs. They are also observed growing adjacent to biofilms, which sometimes fill in the cracks of the coralloid nodes. Preliminary results show the presence of opal, calcite, quartz and other minor minerals in the speleothems. This study seeks to understand the formation mechanism and source of these secondary minerals, as well as determine their possible relation to the biofilms. This will be done through the analysis of the water chemistry, isotope geochemistry and microscale mineralogy.

  5. A study on stress corrosion cracking of explosive plugged part

    International Nuclear Information System (INIS)

    Kaga, Seiichi; Fujii, Katsuhiro; Yamamoto, Yoshiaki; Sakuma, Koosuke; Hibi, Seiji; Morimoto, Hiroyoshi.

    1986-01-01

    Studies on the stress corrosion cracking of explosive plugged part are conducted. SUS 304 stainless steel is used as testing material. The distribution of residual stress in plug and tube plate after plugging is obtained. The effect of residual stress on the stress corrosion cracking is studied. Residual stress in tube plate near the plug is compressive and stress corrosion cracking dose not occur in the tube plate there, and it occurs on the inner surface of plug because of residual tensile stress in axial direction of the plug. Stress corrosion test in MgCl 2 solution under constant load is conducted. The susceptibility to stress corrosion cracking of the explosive bonded boundary is lower than that of base metal because of greater resistance to plastic deformation. Stress corrosion test in high temperature and high pressure pure water is also conducted by means of static type of autoclave but stress corrosion cracking does not occur under the testing condition used. (author)

  6. ABC of floor heating systems. Planning, installation, operation. A compendium for architects, building and heating system designers, trade, heating engineers and technicians, and floor finishers and tile setters. Das ABC der Warmwasser-Fussbodenheizung. Planung, Ausfuehrung, Betrieb. Ein Kompendium fuer Architekten, Bau- und Heizungsplaner, Fachhandel, Heizungsbauer, Installateure sowie das Estrich- und Fliesenleger-Handwerk

    Energy Technology Data Exchange (ETDEWEB)

    Radtke, U

    1984-01-01

    40 illustrated chapters give detailed technical information on subjects such as floor heating systems; plastic and metal pipes; the dynamic behaviour and working properties of plastic tubes; oxygen diffusion (restrictive measures); insulation materials; thermal insulation and noise pollution abatement; floor pavements; structure of heated floors; flooring materials; thermal physics of floor heating systems; determination of thermal outputs; control of floor heating systems; distributors; pipe joints/connections; tools; antifreezes; expertises and damage - practical experiences; floor heating systems from the medical point of view; dispatch of orders; legal requirements of floor heating systems; standards and regulations. (HWJ).

  7. Cracking hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    Forwood, G F; Lane, M; Taplay, J G

    1921-10-07

    In cracking and hydrogenating hydrocarbon oils by passing their vapors together with steam over heated carbon derived from shale, wood, peat or other vegetable or animal matter, the gases from the condenser are freed from sulfuretted hydrogen, and preferably also from carbon dioxide, and passed together with oil vapors and steam through the retort. Carbon dioxide may be removed by passage through slaked lime, and sulfuretted hydrogen by means of hydrated oxide of iron. Vapors from high-boiling oils and those from low-boiling oils are passed alternately through the retort, so that carbon deposited from the high-boiling oils is used up during treatment of low-boiling oils.

  8. Analysis of WWER 1000 collector cracking mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Matocha, K.; Wozniak, J. [Vitkovice J.S.C., Ostrava (Switzerland)

    1997-12-31

    The presentation reviews the large experimental program, started in 1993 in Vitkovice, where the main aim was: (1) a detailed study of strain and thermal ageing, dissolved oxygen content and temperature on subcritical crack growth in 10NiMo8.5 (10GN2MFA) steel, (2) a detailed study of the effect of high temperature water and tube expansion technology on fracture behaviour of ligaments between holes for heat exchange tubes, and (3) a detailed study of the effect of drilling, tube expansion technology and heat treatment on residual stresses on the surface of holes for heat exchange tubes. The aim of all these investigations was to find a dominant damage mechanism responsible for collector cracking to be able to judge the efficiency of implemented modifications and suggested countermeasures and to answer a very important question whether proper operation conditions (mainly water chemistry) make the operation of steam generators made in Vitcovice safe throughout the planned lifetime. 10 refs.

  9. Analysis of WWER 1000 collector cracking mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Matocha, K; Wozniak, J [Vitkovice J.S.C., Ostrava (Switzerland)

    1998-12-31

    The presentation reviews the large experimental program, started in 1993 in Vitkovice, where the main aim was: (1) a detailed study of strain and thermal ageing, dissolved oxygen content and temperature on subcritical crack growth in 10NiMo8.5 (10GN2MFA) steel, (2) a detailed study of the effect of high temperature water and tube expansion technology on fracture behaviour of ligaments between holes for heat exchange tubes, and (3) a detailed study of the effect of drilling, tube expansion technology and heat treatment on residual stresses on the surface of holes for heat exchange tubes. The aim of all these investigations was to find a dominant damage mechanism responsible for collector cracking to be able to judge the efficiency of implemented modifications and suggested countermeasures and to answer a very important question whether proper operation conditions (mainly water chemistry) make the operation of steam generators made in Vitcovice safe throughout the planned lifetime. 10 refs.

  10. Polygons on Crater Floor

    Science.gov (United States)

    2003-01-01

    MGS MOC Release No. MOC2-357, 11 May 2003This Mars Global Surveyor (MGS) Mars Orbiter Camera (MOC) picture shows a pattern of polygons on the floor of a northern plains impact crater. These landforms are common on crater floors at high latitudes on Mars. Similar polygons occur in the arctic and antarctic regions of Earth, where they indicate the presence and freeze-thaw cycling of ground ice. Whether the polygons on Mars also indicate water ice in the ground is uncertain. The image is located in a crater at 64.8oN, 292.7oW. Sunlight illuminates the scene from the lower left.

  11. Crack growth rates in vessel head penetration materials

    International Nuclear Information System (INIS)

    Gomez Briceno, D.; Lapena, J.; Blazquez, F.

    1994-01-01

    The cracks detected in reactor vessel head penetrations in certain European plants have been attributed to Primary Water Stress Corrosion Cracking (PWSCC). The penetrations in question are made from Inconel 600. The susceptibility of this alloy to PWSCC has been widely studied in relation to use of this material for steam generator tubes. When the first reactor vessel head penetration cracks were detected, most of the available data on crack propagation rates were from test specimens made from steam generator tubes and tested under conditions that questioned the validity of these data for assessment of the evolution of cracks in penetrations. For this reason, the scope of the Spanish Research Project on the Inspection and Repair of PWR reactor vessel head penetrations included the acquisition of data on crack propagation rates in Inconel 600, representative of the materials used for vessel head penetrations. (authors). 1 fig., 2 tabs., 6 refs

  12. CAISSON TYPE HOLLOW FLOOR SLABS OF MONOLITHIC MULTI-STOREYED BUILDINGS

    OpenAIRE

    Malakhova Anna Nikolaevna

    2016-01-01

    One of the disadvantages of building structures made of reinforced concrete is their considerable weight. One of the trends to decrease the weight of concrete structures, including floor slabs, is the arrangement of voids in the cross-sectional building structures. In Russian and foreign practice paper, cardboard and plastic tubes has been used for creation of voids in the construction of monolithic floor slabs. Lightweight concretes were also used for production of precast hollow core floor ...

  13. 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

  14. Mechanical model of water inrush from coal seam floor based on triaxial seepage experiments

    Institute of Scientific and Technical Information of China (English)

    Yihui Pang; Guofa Wang; Ziwei Ding

    2014-01-01

    In order to study the mechanism of confined water inrush from coal seam floor, the main influences on permeability in the process of triaxial seepage experiments were analyzed with methods such as laboratory experiments, theoretical analysis and mechanical model calculation. The crack extension rule and the ultimate destruction form of the rock specimens were obtained. The mechanism of water inrush was explained reasonably from mechanical point of view. The practical criterion of water inrush was put forward. The results show that the rock permeability ‘‘mutation’’ phe-nomenon reflects the differences of stress state and cracks extension rate when the rock internal crack begins to extend in large-scale. The rock ultimate destruction form is related to the rock lithology and the angle between crack and principal stress. The necessary condition of floor water inrush is that the mining pressure leads to the extension and transfixion of the crack. The sufficient condition of floor water inrush is that the confined water’s expansionary stress in normal direction and shear stress in tangential direction must be larger than the internal stress in the crack. With the two conditions satisfied at the same time, the floor water inrush accident will occur.

  15. Pelvic floor muscle function in women with pelvic floor dysfunction

    DEFF Research Database (Denmark)

    Tibaek, Sigrid; Dehlendorff, Christian

    2014-01-01

    The objectives of this study were to investigate the level of pelvic floor muscle (PFM) function in women with pelvic floor dysfunction (PFD) referred by gynaecologists and urologists for in-hospital pelvic floor muscle training (PFMT), and to identity associated factors for a low level of PFM...

  16. Ultrasonic inspection of tube to tube plate welds

    International Nuclear Information System (INIS)

    Telford, D.W.; Peat, T.S.

    1985-01-01

    To monitor the deterioration of a weld between a tube and tube plate which has been repaired by a repair sleeve inside the tube and brazed at one end to the tube, ultrasound from a crystal at the end of a rod is launched, in the form of Lamb-type waves, into the tube through the braze and allowed to travel along the tube to the weld and be reflected back along the tube. The technique may also be used for the type of heat exchanger in which, during construction, the tubes are welded to the tube plate via external sleeves in which case the ultrasound is used in a similar manner to inspect the sleeve/tube plate weld. an electromagnetic transducer may be used to generate the ultrasound. The ultrasonic head comprising the crystal and an acoustic baffle is mounted on a Perspex (RTM) rod which may be rotated by a stepping motor. Echo signals from the region of deterioration may be isolated by use of a time gate in the receiver. The device primarily detects circumferentially orientated cracks, and may be used in heat exchangers in nuclear power plants. (author)

  17. The residual stress evaluation for expansion process of steam generator tubes

    International Nuclear Information System (INIS)

    King, C.-S.; Lee, S.-C.; Shim, D.-N.

    2004-01-01

    The reliability of a nuclear power plant is affected by the reliability of steam generator tube and the reliability of steam generator tube is affected by stress corrosion cracking(SCC). Many steam generator tubes were experiencing stress corrosion cracking and stress corrosion cracking is affected material characteristics, corrosive environments and added stresses. The added stresses have the manufacturing stresses and operating stresses, the manufacturing stresses include the residual stresses generating in the tube manufacture and tube expanding procedure. We will investigate for influence which affected to residual stresses with tube plastic deformation method and measurement region. (author)

  18. Automated Diagnosis and Classification of Steam Generator Tube Defects

    International Nuclear Information System (INIS)

    Garcia, Gabe V.

    2004-01-01

    A major cause of failure in nuclear steam generators is tube degradation. Tube defects are divided into seven categories, one of which is intergranular attack/stress corrosion cracking (IGA/SCC). Defects of this type usually begin on the outer surface of the tubes and propagate both inward and laterally. In many cases these defects occur at or near the tube support plates. Several different methods exist for the nondestructive evaluation of nuclear steam generator tubes for defect characterization

  19. Impacts of weld residual stresses and fatigue crack growth threshold on crack arrest under high-cycle thermal fluctuations

    International Nuclear Information System (INIS)

    Taheri, Said; Julan, Emricka; Tran, Xuan-Van; Robert, Nicolas

    2017-01-01

    Highlights: • For crack growth analysis, weld residual stress field must be considered through its SIF in presence of a crack. • Presence of cracks of same depth proves their arrest, where equal depth is because mean stress acts only on crack opening. • Not considering amplitudes under a fatigue crack growth threshold (FCGT) does not compensate the lack of FGCT in Paris law. • Propagation rates are close for axisymmetric and circumferential semi-elliptical cracks. - Abstract: High cycle thermal crazing has been observed in some residual heat removal (RHR) systems made of 304 stainless steel in PWR nuclear plants. This paper deals with two types of analyses including logical argumentation and simulation. Crack arrest in networks is demonstrated due to the presence of two cracks of the same depth in the network. This identical depth may be proved assuming that mean stress acts only on crack opening and that cracks are fully open during the load cycle before arrest. Weld residual stresses (WRS) are obtained by an axisymmetric simulation of welding on a tube with a chamfer. Axisymmetric and 3D parametric studies of crack growth on: representative sequences for variable amplitude thermal loading, fatigue crack growth threshold (FCGT), permanent mean stress, cyclic counting methods and WRS, are performed with Code-Aster software using XFEM methodology. The following results are obtained on crack depth versus time: the effect of WRS on crack growth cannot be determined by the initial WRS field in absence of crack, but by the associated stress intensity factor. Moreover the relation between crack arrest depth and WRS is analyzed. In the absence of FCGT Paris’s law may give a significant over-estimation of crack depth even if amplitudes of loading smaller than FCGT have not been considered. Appropriate depth versus time may be obtained using different values of FCGT, but axisymmetric simulations do not really show a possibility of arrest for shallow cracks in

  20. Impacts of weld residual stresses and fatigue crack growth threshold on crack arrest under high-cycle thermal fluctuations

    Energy Technology Data Exchange (ETDEWEB)

    Taheri, Said, E-mail: Said.taheri@edf.fr [EDF-LAB, IMSIA, 7 Boulevard Gaspard Monge, 91120 Palaiseau Cedex (France); Julan, Emricka [EDF-LAB, AMA, 7 Boulevard Gaspard Monge, 91120 Palaiseau Cedex (France); Tran, Xuan-Van [EDF Energy R& D UK Centre/School of Mechanical, Aerospace and Civil Engineering, The University of Manchester, Manchester M13 9PL (United Kingdom); Robert, Nicolas [EDF-DPN, UNIE, Strategic Center, Saint Denis (France)

    2017-01-15

    Highlights: • For crack growth analysis, weld residual stress field must be considered through its SIF in presence of a crack. • Presence of cracks of same depth proves their arrest, where equal depth is because mean stress acts only on crack opening. • Not considering amplitudes under a fatigue crack growth threshold (FCGT) does not compensate the lack of FGCT in Paris law. • Propagation rates are close for axisymmetric and circumferential semi-elliptical cracks. - Abstract: High cycle thermal crazing has been observed in some residual heat removal (RHR) systems made of 304 stainless steel in PWR nuclear plants. This paper deals with two types of analyses including logical argumentation and simulation. Crack arrest in networks is demonstrated due to the presence of two cracks of the same depth in the network. This identical depth may be proved assuming that mean stress acts only on crack opening and that cracks are fully open during the load cycle before arrest. Weld residual stresses (WRS) are obtained by an axisymmetric simulation of welding on a tube with a chamfer. Axisymmetric and 3D parametric studies of crack growth on: representative sequences for variable amplitude thermal loading, fatigue crack growth threshold (FCGT), permanent mean stress, cyclic counting methods and WRS, are performed with Code-Aster software using XFEM methodology. The following results are obtained on crack depth versus time: the effect of WRS on crack growth cannot be determined by the initial WRS field in absence of crack, but by the associated stress intensity factor. Moreover the relation between crack arrest depth and WRS is analyzed. In the absence of FCGT Paris’s law may give a significant over-estimation of crack depth even if amplitudes of loading smaller than FCGT have not been considered. Appropriate depth versus time may be obtained using different values of FCGT, but axisymmetric simulations do not really show a possibility of arrest for shallow cracks in

  1. 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

  2. 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.

  3. 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.

  4. Chronic pelvic floor dysfunction.

    Science.gov (United States)

    Hartmann, Dee; Sarton, Julie

    2014-10-01

    The successful treatment of women with vestibulodynia and its associated chronic pelvic floor dysfunctions requires interventions that address a broad field of possible pain contributors. Pelvic floor muscle hypertonicity was implicated in the mid-1990s as a trigger of major chronic vulvar pain. Painful bladder syndrome, irritable bowel syndrome, fibromyalgia, and temporomandibular jaw disorder are known common comorbidities that can cause a host of associated muscular, visceral, bony, and fascial dysfunctions. It appears that normalizing all of those disorders plays a pivotal role in reducing complaints of chronic vulvar pain and sexual dysfunction. Though the studies have yet to prove a specific protocol, physical therapists trained in pelvic dysfunction are reporting success with restoring tissue normalcy and reducing vulvar and sexual pain. A review of pelvic anatomy and common findings are presented along with suggested physical therapy management. Copyright © 2014 Elsevier Ltd. All rights reserved.

  5. A Study on the Profile Change Measurement of Steam Generator Tubes with Tube Expansion Methods

    International Nuclear Information System (INIS)

    Kim, Young Kyu; Song Myung Ho; Choi, Myung Sik

    2011-01-01

    Steam generator tubes for nuclear power plants contain the local shape transitions on their inner or outer surface such as dent, bulge, over-expansion, eccentricity, deflection, and so on by the application of physical force during the tube manufacturing and steam generator assembling and by the sludge (that is, corrosion products) produced during the plant operation. The structural integrity of tubes will be degraded by generating the corrosive crack at that location. The profilometry using the traditional bobbin probes which are currently applied for measuring the profile change of tubes gives us basic information such as axial locations and average magnitudes of deformations. However, the three-dimensional quantitative evaluation on circumferential locations, distributional angle, and size of deformations will have to be conducted to understand the effects of residual stresses increased by local deformations on corrosive cracking of tubes. Steam generator tubes of Korean standard nuclear power plants expanded within their tube-sheets by the explosive expansion method and suffered from corrosive cracks in the early stage of power operation. Thus, local deformations of steam generator tubes at the top of tube-sheet were measured with an advanced rotating probe and a laser profiling system for the two cases where the tubes expanded by the explosive expansion method and hydraulic expansion. Also, the trends of eccentricity, deflection, and over-expansion of tubes were evaluated. The advanced eddy current profilometry was confirmed to provide accurate information of local deformations compared with laser profilometry

  6. 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 ...

  7. Unloading Effect on Delayed Hydride Cracking in Zirconium Alloys

    International Nuclear Information System (INIS)

    Kim, Young Suk; Kim, Sung Soo

    2010-01-01

    It is well-known that a tensile overload retards not only the crack growth rate (CGR) in zirconium alloys during the delayed hydride cracking (DHC) tests but also the fatigue crack growth rate in metals, the cause of which is unclear to date. A considerable decrease in the fatigue crack growth rate due to overload is suggested to occur due either to the crack closure or to compressive stresses or strains arising from unloading of the overload. However, the role of the crack closure or the compressive stress in the crack growth rate remains yet to be understood because of incomplete understanding of crack growth kinetics. The aim of this study is to resolve the effect of unloading on the CGR of zirconium alloys, which comes in last among the unresolved issues as listed above. To this end, the CGRs of the Zr-2.5Nb tubes were determined at a constant temperature under the cyclic load with the load ratio, R changing from 0.13 to 0.66 where the extent of unloading became higher at the lower R. More direct evidence for the effect of unloading after an overload is provided using Simpson's experiment investigating the effect on the CGR of a Zr-2.5Nb tube of the stress states of the prefatigue crack tip by unloading or annealing after the formation of a pre-fatigue crack

  8. Modular Flooring System

    Science.gov (United States)

    Thate, Robert

    2012-01-01

    The modular flooring system (MFS) was developed to provide a portable, modular, durable carpeting solution for NASA fs Robotics Alliance Project fs (RAP) outreach efforts. It was also designed to improve and replace a modular flooring system that was too heavy for safe use and transportation. The MFS was developed for use as the flooring for various robotics competitions that RAP utilizes to meet its mission goals. One of these competitions, the FIRST Robotics Competition (FRC), currently uses two massive rolls of broadloom carpet for the foundation of the arena in which the robots are contained during the competition. The area of the arena is approximately 30 by 72 ft (approximately 9 by 22 m). This carpet is very cumbersome and requires large-capacity vehicles, and handling equipment and personnel to transport and deploy. The broadloom carpet sustains severe abuse from the robots during a regular three-day competition, and as a result, the carpet is not used again for competition. Similarly, broadloom carpets used for trade shows at convention centers around the world are typically discarded after only one use. This innovation provides a green solution to this wasteful practice. Each of the flooring modules in the previous system weighed 44 lb (.20 kg). The improvements in the overall design of the system reduce the weight of each module by approximately 22 lb (.10 kg) (50 %), and utilize an improved "module-to-module" connection method that is superior to the previous system. The MFS comprises 4-by-4-ft (.1.2-by- 1.2-m) carpet module assemblies that utilize commercially available carpet tiles that are bonded to a lightweight substrate. The substrate surface opposite from the carpeted surface has a module-to-module connecting interface that allows for the modules to be connected, one to the other, as the modules are constructed. This connection is hidden underneath the modules, creating a smooth, co-planar flooring surface. The modules are stacked and strapped

  9. As the crack in the Geiger counter came. Historical scientific analysis and didactic aspects of the Geiger-Mueller counting tube; Wie das Knacken in den Geigerzaehler kam. Wissenschaftshistorische Analyse und fachdidaktische Aspekte des Geiger-Mueller Zaehlrohrs

    Energy Technology Data Exchange (ETDEWEB)

    Korff, Sebastian

    2014-11-10

    This thesis studies the creation and establishment history of this instrument called first electron counting tube in the years 1928 and 1929. It deals thereby with the last two years of the common work of Hans Geiger and Walter Mueller, from which the measuring instrument later renamed to Geiger-Mueller counting tube. The results of this scientific case study are didactically worked out and made usable for the teaching of physics in the school.

  10. Modified Dugdale cracks and Fictitious cracks

    DEFF Research Database (Denmark)

    Nielsen, Lauge Fuglsang

    1998-01-01

    A number of theories are presented in the literature on crack mechanics by which the strength of damaged materials can be predicted. Among these are theories based on the well-known Dugdale model of a crack prevented from spreading by self-created constant cohesive flow stressed acting in local...... areas, so-called fictitious cracks, in front of the crack.The Modified Dugdale theory presented in this paper is also based on the concept of Dugdale cracks. Any cohesive stress distribution, however, can be considered in front of the crack. Formally the strength of a material weakened by a modified...... Dugdale crack is the same as if it has been weakened by the well-known Griffith crack, namely sigma_CR = (EG_CR/phi)^1/2 where E and 1 are Young's modulus and crack half-length respectively, and G_CR is the so-called critical energy release rate. The physical significance of G_CR, however, is different...

  11. Evaluation of steam generator tube integrity during earthquakes

    Energy Technology Data Exchange (ETDEWEB)

    Kusakabe, Takaya; Kodama, Toshio [Mitsubishi Heavy Industries Ltd., Kobe (Japan). Kobe Shipyard and Machinery Works; Takamatsu, Hiroshi; Matsunaga, Tomoya

    1999-07-01

    This report shows an experimental study on the strength of PWR steam generator (SG) tubes with various defects under cyclic loads which simulate earthquakes. The tests were done using same SG tubing as actual plants with axial and circumferential defects with various length and depth. In the tests, straight tubes were loaded with cyclic bending moments to simulate earthquake waves and number of load cycles at which tube leak started or tube burst was counted. The test results showed that even tubes with very long crack made by EDM more than 80% depth could stand the maximum earthquake, and tubes with corrosion crack were far stronger than those. Thus the integrity of SG tubes with minute potential defects was demonstrated. (author)

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

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

    1996-01-01

    The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

  13. Sea floor magnetic observatory

    Science.gov (United States)

    Korepanov, V.; Prystai, A.; Vallianatos, F.; Makris, J.

    2003-04-01

    The electromagnetic precursors of seismic hazards are widely accepted as strong evidence of the approaching earthquake or volcano eruption. The monitoring of these precursors are of main interest in densely populated areas, what creates serious problems to extract them at the strong industrial noise background. An interesting possibility to improve signal-to-noise ratio gives the installation of the observation points in the shelf zones near the possible earthquake places, what is fairly possible in most seismically active areas in Europe, e. g. in Greece and Italy. The serious restriction for this is the cost of the underwater instrumentation. To realize such experiments it requires the unification of efforts of several countries (e. g., GEOSTAR) or of the funds of some great companies (e. g., SIO magnetotelluric instrument). The progress in electronic components development as well as the appearance of inexpensive watertight glass spheres made it possible to decrease drastically the price of recently developed sea floor magnetic stations. The autonomous vector magnetometer LEMI-301 for sea bed application is described in the report. It is produced on the base of three-component flux-gate sensor. Non-magnetic housing and minimal magnetism of electronic components enable the instrument to be implemented as a monoblock construction where the electronic unit is placed close to the sensor. Automatic circuit provides convenient compensation of the initial field offset and readings of full value (6 digits) of the measured field. Timing by internal clock provides high accuracy synchronization of data. The internal flash memory assures long-term autonomous data storage. The system also has two-axes tilt measurement system. The methodological questions of magnetometer operation at sea bed were studied in order to avoid two types of errors appearing at such experimental cases. First is sea waving influence and second one magnetometer orientation at its random positioning on

  14. Cocaine (Coke, Crack) Facts

    Science.gov (United States)

    ... That People Abuse » Cocaine (Coke, Crack) Facts Cocaine (Coke, Crack) Facts Listen Cocaine is a white ... 69 KB) "My life was built around getting cocaine and getting high." ©istock.com/ Marjot Stacey is ...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.

    1997-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  18. Floor heating maximizes residents` comfort

    Energy Technology Data Exchange (ETDEWEB)

    Tirkkanen, P.; Wikstroem, T.

    1996-11-01

    Storing heat in floors by using economical night-time electricity does not increase the specific consumption of heating. According to studies done by IVO, the optimum housing comfort is achieved if the room is heated mainly by means of floor heating that is evened out by window or ceiling heating, or by a combination of all three forms of heating. (orig.)

  19. Equipment for inspection and carrying out repairs, if required, for tube bundles of steam raising units

    International Nuclear Information System (INIS)

    Gugel, G.

    1976-01-01

    The equipment solves the problem of being able to inspect and possibly to repair U-tubes of a vertical steam raising unit standing on a tube floor, without draining the primary medium and bringing the test equipment and tools into the inside of the boiler first. This is achieved by leaving a considerable part of the equipment permanently in the hemispherical space under the tube floor and operating it from the outside, on the other side of the concrete shielding. An inspection tube is threaded in turn horizontally through a concrete shield, a tube duct in the heat insulation of the steam raising unit, and through a hole in the hemispherical space under the tube floor into this space. The end of an angle tube can be moved axially from outside the concrete shield and can be rotated in a semicircle above the tube axis. By interposing a, for example, 12 part distributor with 12 short, differently bent tubes 12 adjacent tubes opening into the tube floor can be controlled and tested, by axial movement of the angle tube together with the distributor, e.g. 4 x 12 other U tubes. A turbulent flow sensor, for example, can be introduced through the angle tube and distributor. In the non-operational condition the equipment is moved into a recess via a supporting angle and stopped there. (ORU) [de

  20. Working session 2: Tubing inspection

    International Nuclear Information System (INIS)

    Guerra, J.; Tapping, R.L.

    1997-01-01

    This session was attended by delegates from 10 countries, and four papers were presented. A wide range of issues was tabled for discussion. Realizing that there was limited time available for more detailed discussion, three topics were chosen for the more detailed discussion: circumferential cracking, performance demonstration (to focus on POD and sizing), and limits of methods. Two other subsessions were organized: one dealt with some challenges related to the robustness of current inspection methods, especially with respect to leaving cracked tubes in service, and the other with developing a chart of current NDE technology with recommendations for future development. These three areas are summarized in turn, along with conclusions and/or recommendations. During the discussions there were four presentations. There were two (Canada, Japan) on eddy current probe developments, both of which addressed multiarray probes that would detect a range of flaws, one (Spain) on circumferential crack detection, and one (JRC, Petten) on the recent PISC III results

  1. Formation of thermal fatigue cracks in periodic rapid quenching of metal

    Energy Technology Data Exchange (ETDEWEB)

    Ots, A. [Tallinn Technical University, Thermal Engineering Department, Tallinn (Estonia)

    1998-12-31

    Water lancing is an effective technique for cleaning boiler heating surfaces from ash deposits by burning low-grade fuels with complicated composition of mineral matter. In water cleaning cycles of boiler`s heat transfer surfaces due to rapid quenching destruction of corrosion protective oxide film and formation of thermal fatigue cracks on the outer surface of the tube`s metal occur. The criterion of the thermal fatigue cracks` formation and their growth intensity depend on the character of temperature field in the tube`s metal outer layer. The solution of non-stationary heat conductivity equation for metal rapid quenching conditions is given. The convective heat transfer coefficients from hot metal surface to water jet were established experimentally. Thermal fatigue crack growth intensity was investigated in real boilers` heat transfer surfaces` tubes as well as in laboratory conditions. The formula for predicting thermal fatigue cracks` depth depending on the number of cleaning cycles. (orig.) 5 refs.

  2. Formation of thermal fatigue cracks in periodic rapid quenching of metal

    Energy Technology Data Exchange (ETDEWEB)

    Ots, A [Tallinn Technical University, Thermal Engineering Department, Tallinn (Estonia)

    1999-12-31

    Water lancing is an effective technique for cleaning boiler heating surfaces from ash deposits by burning low-grade fuels with complicated composition of mineral matter. In water cleaning cycles of boiler`s heat transfer surfaces due to rapid quenching destruction of corrosion protective oxide film and formation of thermal fatigue cracks on the outer surface of the tube`s metal occur. The criterion of the thermal fatigue cracks` formation and their growth intensity depend on the character of temperature field in the tube`s metal outer layer. The solution of non-stationary heat conductivity equation for metal rapid quenching conditions is given. The convective heat transfer coefficients from hot metal surface to water jet were established experimentally. Thermal fatigue crack growth intensity was investigated in real boilers` heat transfer surfaces` tubes as well as in laboratory conditions. The formula for predicting thermal fatigue cracks` depth depending on the number of cleaning cycles. (orig.) 5 refs.

  3. Numerical forensic model for the diagnosis of a full-scale RC floor

    Directory of Open Access Journals (Sweden)

    Ahmed B. Shuraim

    Full Text Available The paper presents the results of an investigation on the diagnosis and assessment of a full-scale reinforced concrete floor utilizing a 3-D forensic model developed in the framework of plasticity-damage approach. Despite the advancement in nonlinear finite element formulations and models, there is a need to verify models on nontrivial challenging structures. Various standards on strengthening existing structures consider numerical diagnosis as a major stage involving safety and economical aspects. Accordingly, model validity is a major issue that should preferably be examined against realistic large-scale tests. This was done in this study by investigating a one-story joist floor with wide shallow beams supported on columns. The surveyed cracking patterns on the entire top side of the floor were reproduced by the forensic model to a reasonable degree in terms of orientation and general location. Concrete principal plastic tensile strain was shown to be a good indirect indicator of cracking patterns. However, identifying the underlying reasons of major cracks in the floor required correlating with other key field parameters including deflections, and internal moments. Therefore, the ability of the forensic model to reproduce the surveyed damage state of the floor provided a positive indication on the material models, spatial representation, and parameter selection. Such models can be used as forensic tools for assessing the existing conditions as required by various standards and codes.

  4. Lifetime forecasting of a WWER NPP steam generator tube bundle from stress corrosion conditions

    International Nuclear Information System (INIS)

    Sereda, E.V.; Gorbatykh, V.P.

    1984-01-01

    An approach is outlined to the description of corrosion cracking of austenitic stainless steels in hot chloride solutions to predict the failure of WWER NPP steam generator heat exchange tubes. The dependence of the corrosion cracking development rate on the chloride concentration and characteristic electrochemical potentials is suggsted. The approach permits also to determine the quantity of damaged tubes versus the operation parameters

  5. Oxygen pitting failure of a bagasse boiler tube

    CSIR Research Space (South Africa)

    Heyes, AM

    2001-04-01

    Full Text Available Examination of a failed roof tube from a bagasse boiler showed transverse through-cracks and extensive pitting. The pitting was typically oxygen induced pitting and numerous fatigue cracks had started within these pits. It is highly probable...

  6. Ultrasonic inspection of inpile tubes

    International Nuclear Information System (INIS)

    Boyd, D.M.; Bossi, H.

    1985-01-01

    The in-service inspection (ISI) of inpile tubes can be performed accurately and safely with a semiautomatic ultrasonic inspection system. The ultrasonic technique uses a set of multiple transducers to detect and size cracks, voids, and laminations radially and circumferentially. Welds are also inspected for defects. The system is designed to inspect stainless steel and Inconel tubes ranging from 53.8 mm (2.12 in.) to 101.6 mm (4 in.) inner diameter with wall thickness on the order of 5 mm. The inspection head contains seven transducers mounted in a surface-following device. Six angle-beam transducers generate shear waves in the tubes. Two of the six are oriented to detect circumferential cracks, and two detect axial cracks. Although each of these four transducers is used in the pulse-echo mode, they are oriented in aligned sets so pitch-catch operation is possible if desired. The remaining angle-beam transducers are angulated to detect flaws that are off axial or circumferential orientation. The seventh transducer is used for longitudinal inspection and detects and sizes laminar-type defects

  7. Internal ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

    Furlan, J.; Soleille, G.; Chalaye, H.

    1983-01-01

    The ''in situ'' inspection of steam generator tubes uses generally Foucault currents before starting and along its life. This inspection aims at searching cracks and corrosion defects. The Foucault current method is quite badly adapted to ''closed crack'' detection, for it doesn't introduce neither resistivity or magnetic permeability variation, or lack of matter. More, it is sensible to the magnetic properties of the tube itself and to its environment (tubular or support plates). It is why, this first systematic inspection has to be completed by an ultrasonic one allowing to bring new elements in the uncertain cases. A device with an internal probe has been developed. It ''lights'' the tube wall with the aid of a transducer of which beam reflects on a mirror. Operating conditions are the same as for Foucault current testing, that is to say the probe moves inside the tube without rotation of the device (bent parts are excluded) [fr

  8. Crustal Ages of the Ocean Floor - Poster

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — The Crustal Ages of the Ocean Floor Poster was created at NGDC using the Crustal Ages of the Ocean Floor database draped digitally over a relief of the ocean floor...

  9. Steam generator tube performance. Experience with water-cooled nuclear power reactors during 1985

    International Nuclear Information System (INIS)

    Tatone, O.S.; Tapping, R.L.

    1988-12-01

    The performance of steam generator tubes at water-cooled reactors during 1985 has been reviewed. Seventy-three of 168 reactors in the survey experienced tube degradation sufficient for the tubes to be plugged. The number of tubes plugged was 6837 or 0.28% of those in service. The leading cause of tube failure was stress corrosion cracking from the primary side. Stress corrosion cracking or intergranular attack from the secondary side and pitting were also major causes of tube failure. Unlike most previous years, fretting was a substantial problem at some reactors. Overall, corrosion continued to account for more than 80% of the defects. 20 refs

  10. Thermal fatigue cracking of austenitic stainless steels

    International Nuclear Information System (INIS)

    Fissolo, A.

    2001-01-01

    This report deals with the thermal fatigue cracking of austenitic stainless steels as AISI 316 LN and 304 L. Such damage has been clearly observed for some components used in Fast Breeder reactors (FBR) and Pressure Water Reactor (PWR). In order to investigate thermal fatigue, quasi-structural specimen have been used. In this frame, facilities enforcing temperature variations similar to those found under the operation conditions have been progressively developed. As for components, loading results from impeded dilatation. In the SPLASH facility, the purpose was to establish accurate crack initiation conditions in order to check the relevance of the usual component design methodology. The tested specimen is continuously heated by the passage of an electrical DC current, and submitted to cyclic thermal down shock (up to 1000 deg C/s) by means of periodical spraying of water on two opposite specimen faces. The number of cycles to crack initiation N i is deduced from periodic examinations of the quenched surfaces, by means of optical microscopy. It is considered that initiation occurs when at least one 50μm to 150□m long crack is observed. Additional SPLASH tests were performed for N >> N i , with a view to investigate the evolution of a surface multiple cracking network with the number of cycles N. The CYTHIA test was mainly developed for the purpose of assessing crack growth dynamics of one isolated crack in thermal fatigue conditions. Specimens consist of thick walled tubes with a 1 mm circular groove is spark-machined at the specimen centre. During the test, the external wall of the tube is periodically heated by using a HF induction coil (1 MHz), while its internal wall is permanently cooled by flowing water. Total crack growth is derived from post-mortem examinations, whereby the thermal fatigue final rupture surface is oxidized at the end of the test. The specimen is broken afterwards under mechanical fatigue at room temperature. All the tests confirm that

  11. Flooring choices for newborn ICUs.

    Science.gov (United States)

    White, R D

    2007-12-01

    Floors are a major element of newborn intensive care unit (NICU) construction. They provide visual cues, sound control, and with certain materials, some degree of physical comfort for workers. Flooring materials may entail a significant cost for installation and upkeep and can have substantial ecological impact, both in the choice of the flooring itself, as well as the substances used to clean it. In this article the important aspects to consider for each factor are explored and recommendations are offered for appropriate choices in various NICU areas.

  12. Floor cooler for floor trough of a nuclear reactor

    International Nuclear Information System (INIS)

    Friedrich, H.J.

    1985-01-01

    Cooling pipes are situated below the floor trough of a BWR, which are connected to the annular distribution or collection pipes. The distribution and collection pipes are connected by parallel hairpin pipes with involute shape to the centre of the floor trough. These hairpin pipes are situated in a lower plane than the annular distribution pipe to the centre and in a higher plane from the centre to the outer annular collector pipe. (orig./HP) [de

  13. Pelvic floor physical therapy in urogynecologic disorders.

    Science.gov (United States)

    Kotarinos, Rhonda K

    2003-08-01

    Physical therapists are uniquely qualified to treat pelvic floor dysfunction with conservative management techniques. Techniques associated with incontinence and support functions of the pelvic floor include bladder training and pelvic floor rehabilitation: pelvic floor exercises, biofeedback therapy, and pelvic floor electrical stimulation. Pain associated with mechanical pelvic floor dysfunction can be treated by physical therapists utilizing various manual techniques and modalities. Research documents that conservative management is effective in treating many conditions associated with pelvic floor dysfunction. Research should be conducted to determine if addressing diastasis recti and contracture of the pelvic floor musculature should be a component of the standard physical therapy protocol.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  15. Investigation of Helicopter Longeron Cracks

    Science.gov (United States)

    Newman, John A.; Baughman, James; Wallace, Terryl A.

    2009-01-01

    Four cracked longerons, containing a total of eight cracks, were provided for study. Cracked regions were cut from the longerons. Load was applied to open the cracks, enabling crack surface examination. Examination revealed that crack propagation was driven by fatigue loading in all eight cases. Fatigue crack initiation appears to have occurred on the top edge of the longerons near geometric changes that affect component bending stiffness. Additionally, metallurgical analysis has revealed a local depletion in alloying elements in the crack initiation regions that may be a contributing factor. Fatigue crack propagation appeared to be initially driven by opening-mode loading, but at a crack length of approximately 0.5 inches (12.7 mm), there is evidence of mixed-mode crack loading. For the longest cracks studied, shear-mode displacements destroyed crack-surface features of interest over significant portions of the crack surfaces.

  16. Crack detecting method

    International Nuclear Information System (INIS)

    Narita, Michiko; Aida, Shigekazu

    1998-01-01

    A penetration liquid or a slow drying penetration liquid prepared by mixing a penetration liquid and a slow drying liquid is filled to the inside of an artificial crack formed to a member to be detected such as of boiler power generation facilities and nuclear power facilities. A developing liquid is applied to the periphery of the artificial crack on the surface of a member to be detected. As the slow-drying liquid, an oil having a viscosity of 56 is preferably used. Loads are applied repeatedly to the member to be detected, and when a crack is caused to the artificial crack, the permeation liquid penetrates into the crack. The penetration liquid penetrated into the crack is developed by the developing liquid previously coated to the periphery of the artificial crack of the surface of the member to be detected. When a crack is caused, since the crack is developed clearly even if it is a small opening, the crack can be recognized visually reliably. (I.N.)

  17. WWER steam generator tube structural and leakage integrity

    International Nuclear Information System (INIS)

    Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

    1998-01-01

    The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

  18. Feedwater heater tube-to-tubesheet connections

    International Nuclear Information System (INIS)

    Yokell, S.

    1993-01-01

    This paper discusses some practical aspects of expanded, welded, and welded-and-expanded feedwater heater tube-to-tubesheet joints. It outlines elastic-plastic tube expanding theory. It examines uniform-pressure-expanded tube joint strength and correlating roller-expanded joint strength with wall reduction and rolling torque. For materials subject to stress-corrosion cracking (SCC), it recommends heat treating tube ends before expanding. For materials subject to fatigue and tube-end cracking, it advocates two-stage expanding: (1) expanding enough to create firm tube-hole contact over the full tubesheet thickness; and (2) re-expanding at full pressure or torque. The paper emphasizes the desirability of segregating heats of tubing, mapping the tube-heat locations and making the heat map a permanent part of the heater maintenance file. It recommends when to provide TEMA/HEI Power Plant Standard annular grooves for roller-expanding and provides an equation for determining optimum groove width for uniform-pressure expanding. The paper also reviews welding requirements for welds of tubes to tubesheets. The review covers front-face welding before and after expanding and the reasons for welding first. It outlines current thinking about definitions of strength- and seal-welds of front-face welded joint in terms of their functions and load-carrying abilities. It presents a proposal for determining the required size of strength welds for use in Section VIII of the ASME Boiler and Pressure Vessel Code (the Code). It shows why welded-and-expanded feedwater heater tube-to-tubesheet joints should be full-strength and full-depth expanded. It makes recommendations for pressure- and leak-testing. This work also proposes the industry consider butt welding the tubes to the steam-side face of the tubesheet as a regular method of tube joining. The results of a survey of manufacturers practices are appended. 30 refs., 14 figs

  19. Price floors for emissions trading

    International Nuclear Information System (INIS)

    Wood, Peter John; Jotzo, Frank

    2011-01-01

    Price floors in greenhouse gas emissions trading schemes can guarantee minimum abatement efforts if prices are lower than expected, and they can help manage cost uncertainty, possibly as complements to price ceilings. Provisions for price floors are found in several recent legislative proposals for emissions trading. Implementation however has potential pitfalls. Possible mechanisms are government commitments to buy back permits, a reserve price at auction, or an extra fee or tax on acquittal of emissions permits. Our analysis of these alternatives shows that the fee approach has budgetary advantages and is more compatible with international permit trading than the alternatives. It can also be used to implement more general hybrid approaches to emissions pricing. - Research highlights: → Price floors for emissions trading schemes guarantee a minimum carbon price. → Price floors mean that emissions can be less than specified by the ETS cap. → We examine how price floors can relate to different policy objectives. → We compare different mechanisms for implementing a price floor. → We find that a mechanism where there is an extra tax or fee has advantages.

  20. Potential drop crack measurement systems for CANDU components

    Energy Technology Data Exchange (ETDEWEB)

    Sahney, R [Carleton Univ., Ottawa, ON (Canada)

    1994-12-31

    A project to develop an automated crack measurement system for CANDU pressure tube burst testing is currently underway. The system will utilize either Direct Current Potential Drop (DCPD) or Alternating Current Potential Drop (ACPD) techniques for crack measurement. The preliminary stage of the project involves testing and comparison of both ACPD and DCPD methods on a Zr - 2.5% Nb alloy plate with saw cuts (used to simulate cracks). Preliminary results show that both ACPD and DCPD techniques are capable of detecting cracks; further testing is in progress to determine the ability of each of the two systems to make accurate crack depth measurements. This paper will describe the two potential drop techniques and will present test results from the experimental program. (author). 10 refs., 7 figs.

  1. Curvilinear crack layer propagation

    Science.gov (United States)

    Chudnovsky, Alexander; Chaoui, Kamel; Moet, Abdelsamie

    1987-01-01

    An account is given of an experiment designed to allow observation of the effect of damage orientation on the direction of crack growth in the case of crack layer propagation, using polystyrene as the model material. The direction of crack advance under a given loading condition is noted to be determined by a competition between the tendency of the crack to maintain its current direction and the tendency to follow the orientation of the crazes at its tip. The orientation of the crazes is, on the other hand, determined by the stress field due to the interaction of the crack, the crazes, and the hole. The changes in craze rotation relative to the crack define the active zone rotation.

  2. Analytical TEM of service-induced SCC in alloy 600TT steam generator tubing

    International Nuclear Information System (INIS)

    Wolfe, R.; Legras, L.; Boccanfuso; Martin, A.

    2015-01-01

    In 2008, Vogtle Electric Generating Plant Unit 1 performed tube pulls to confirm outside diameter stress corrosion cracking (ODSCC) in a steam generator with thermally treated Alloy 600TT tubing. Subsequent metallographic and other laboratory work attributed the cracking to the non-optimal microstructure of the tubing and the elevated residual stresses at the expansion transition. In the current work, analytical transmission electron microscopy was performed to gain a better understanding of this in-service cracking through a detailed characterization of the oxides and crack tips. These examinations, which are the first of this kind for U.S. Alloy 600TT tubing service cracks, detected lead (Pb) in the region of the top-of-tube sheet crevice, in oxides at the crack tips, and at degraded grain boundaries. In addition, sulfur was observed in oxides on the outside surface of the tube in the free span area. The presence of Pb at the crack tip and the lack of plasticity on the observed failure surfaces suggest that the environment played a predominant role in the cracking of this tubing with a non-optimal microstructure. The significance of the degradation will be discussed in the context of overall corrosion indications in Alloy 600TT steam generators in the United States. (authors)

  3. Investigation on aerosol transport in containment cracks

    International Nuclear Information System (INIS)

    Parozzi, F.; Chatzidakis, S.; Housiadas, C.; Gelain, T.; Nahas, G.; Plumecocq, W.; Vendel, J.; Herranz, L.E.; Hinis, E.; Journeau, C.; Piluso, P.; Malgarida, E.

    2005-01-01

    Under severe accident conditions, the containment leak-tightness could be threatened by energetic phenomena that could yield a release to the environment of nuclear aerosols through penetrating concrete cracks. As few data are presently available to quantify this aerosol leakage, a specific action was launched in the framework of the Santar Project of the European 6 th Framework Programme. In this context, both theoretical and experimental investigations have been managed to develop a model that can readily be applied within a code like Aster (Accident Source Term Evaluation Code). Particle diffusion, settling, turbulent deposition, diffusiophoresis and thermophoresis have been considered as deposition mechanisms inside the crack path. They have been encapsulated in numerical models set up to reproduce experiments with small tubes and capillaries and simulate the plug formation. Then, an original lagrangian approach has been used to evaluate the crack retention under typical PWR accident conditions, comparing its predictions with those given by the eulerian approach implemented in the ECART code. On the experimental side, the paper illustrates an aerosol production and measurement system developed to validate aerosol deposition models into cracks and the results that can be obtained: a series of tests were performed with monodispersed fluorescein aerosols injected into a cracked concrete sample. A key result that should be further explored refers to the high enhancement of aerosol retention that could be due to steam condensation. Recommendations concerning future experimentation are also given in the paper. (author)

  4. Feeding Tubes

    Science.gov (United States)

    ... feeding therapies have been exhausted. Please review product brand and method of placement carefully with your physician ... Total Parenteral Nutrition. Resources: Oley Foundation Feeding Tube Awareness Foundation Children’s Medical Nutrition Alliance APFED’s Educational Webinar ...

  5. Automation of inspection methods for eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Meurgey, P.; Baumaire, A.

    1990-01-01

    Inspection of all the tubes of a steam generator when the reactor is stopped is required for some of these exchangers affected by stress corrosion cracking. Characterization of each crack, in each tube is made possible by the development of software for processing the signals from an eddy current probe. The ESTELLE software allows a rapid increase of tested tubes, more than 80,000 in 1989 [fr

  6. Probabilistic aspects of fatigue crack propagation data for zirconium-2.5 % niobium

    International Nuclear Information System (INIS)

    Wilkins, B.J.S.; Reich, A.R.

    1976-11-01

    Fatigue crack propagation data for Zr-2.5 % Nb pressure tube material at 20 and 400 deg C are presented. The practical application of these data in terms of error analysis and extrapolation errors is discussed. (author)

  7. Steam generator tube extraction

    International Nuclear Information System (INIS)

    Delorme, H.

    1985-05-01

    To enable tube examination on steam generators in service, Framatome has now developed a process for removing sections of steam generator tubes. Tube sections can be removed without being damaged for treating the tube section expanded in the tube sheet

  8. Functional anatomy of pelvic floor

    Directory of Open Access Journals (Sweden)

    Salvatore Rocca Rossetti

    2016-03-01

    Full Text Available Generally, descriptions of the pelvic floor are discordant, since its complex structures and the complexity of pathological disorders of such structures; commonly the descriptions are sectorial, concerning muscles, fascial developments, ligaments and so on. On the contrary to understand completely nature and function of the pelvic floor it is necessary to study it in the most unitary view and in the most global aspect, considering embriology, philogenesy, anthropologic development and its multiple activities others than urological, gynaecological and intestinal ones. Recent acquirements succeeded in clarifying many aspects of pelvic floor activity, whose musculature has been investigated through electromyography, sonography, magnetic resonance, histology, histochemistry, molecular research. Utilizing recent research concerning not only urinary and gynecologic aspects but also those regarding statics and dynamics of pelvis and its floor, it is now possible to study this important body part as a unit; that means to consider it in the whole body economy to which maintaining upright position, walking and behavior or physical conduct do not share less than urinary, genital, and intestinal functions. It is today possible to consider the pelvic floor as a musclefascial unit with synergic and antagonistic activity of muscular bundles, among them more or less interlaced, with multiple functions and not only the function of pelvic cup closure.

  9. 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

    /CANTIA code features and its use in the assessment of steam generator tube integrity assessment. Various parametric results on tube crack initiation, crack growth rate, distribution through tube cracks and leak rate are presented. Recommendation are made on improvement of the ANL/CANTIA code with new model for the crack growth rate using experimental data based initial crack distribution and improved model for leak rates in the cracks to account for different flaw type. (author)

  10. Crack layer theory

    Science.gov (United States)

    Chudnovsky, A.

    1987-01-01

    A damage parameter is introduced in addition to conventional parameters of continuum mechanics and consider a crack surrounded by an array of microdefects within the continuum mechanics framework. A system consisting of the main crack and surrounding damage is called crack layer (CL). Crack layer propagation is an irreversible process. The general framework of the thermodynamics of irreversible processes are employed to identify the driving forces (causes) and to derive the constitutive equation of CL propagation, that is, the relationship between the rates of the crack growth and damage dissemination from one side and the conjugated thermodynamic forces from another. The proposed law of CL propagation is in good agreement with the experimental data on fatigue CL propagation in various materials. The theory also elaborates material toughness characterization.

  11. Atomistics of crack propagation

    International Nuclear Information System (INIS)

    Sieradzki, K.; Dienes, G.J.; Paskin, A.; Massoumzadeh, B.

    1988-01-01

    The molecular dynamic technique is used to investigate static and dynamic aspects of crack extension. The material chosen for this study was the 2D triangular solid with atoms interacting via the Johnson potential. The 2D Johnson solid was chosen for this study since a sharp crack in this material remains stable against dislocation emission up to the critical Griffith load. This behavior allows for a meaningful comparison between the simulation results and continuum energy theorems for crack extension by appropriately defining an effective modulus which accounts for sample size effects and the non-linear elastic behavior of the Johnson solid. Simulation results are presented for the stress fields of moving cracks and these dynamic results are discussed in terms of the dynamic crack propagation theories, of Mott, Eshelby, and Freund

  12. Effect of rubber flooring on group-housed sows' gait and claw and skin lesions.

    Science.gov (United States)

    Bos, E-J; van Riet, M M J; Maes, D; Millet, S; Ampe, B; Janssens, G P J; Tuyttens, F A M

    2016-05-01

    This study evaluated the influence of floor type on sow welfare in terms of lameness, claw lesions, and skin lesions. In a 2 × 3 factorial design, we have investigated the effect of rubber coverings on concrete floors and the effect of 3 levels of dietary zinc supplementation on locomotion and claw and skin lesions in group-housed sows. Six groups of 21 ± 4 hybrid sows were monitored during 3 successive reproductive cycles. The sows were group housed from d 28 after insemination (d 0) until 1 wk before expected farrowing date (d 108) in pens with either exposed concrete floors or concrete floors covered with rubber in part of the lying area and the fully slatted area. During each reproductive cycle, locomotion and skin lesions were assessed 4 times (d 28, 50, 108, and 140) and claw lesions were assessed twice (d 50 and 140). Results are given as least squares means ± SE. Locomotion and claw scores were given in millimeters, on analog scales of 150 and 160 mm, respectively. Here, we report on the effect of floor type, which did not interact with dietary zinc concentration ( > 0.10 for all variables). At move to group (d 28) and mid gestation (d 50), no differences between floor treatments were seen in locomotion ( > 0.10). At the end of gestation (d 108), sows housed on rubber flooring scored 9.9 ± 4.1 mm better on gait ( flooring at mid gestation (d 50). However, sows on rubber flooring scored worse for "vertical cracks in the wall horn" (difference of 3.4 ± 1.7 mm; = 0.04). At the end of lactation (d 140), both "white line" (difference of 2.9 ± 1 mm; = 0.02) and "claw length" (difference of 4.7 ± 1.4 mm; flooring. No differences for skin lesions were observed between floor treatments. The improved scores for gait toward the end of gestation and some types of claw disorders at mid gestation suggest that rubber flooring in group housing has a beneficial effect on the overall leg health of sows. The documented increase in vertical cracks in the wall horn at d

  13. 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

  14. Evaluation of plastic collapse behavior for multiple cracked structures

    International Nuclear Information System (INIS)

    Moon, Seong In; Chang, Yoon Suk; Kim, Young Jin; Lee, Jin Ho; Song, Myung Ho; Choi, Young Hwan; Hwang, Seong Sik

    2004-01-01

    Until now, the 40% of wall thickness criterion, which is generally used for the plugging of steam generator tubes, has been applied only to a single cracked geometry. In the previous study by the authors, a total number of 9 local failure prediction models were introduced to estimate the coalescence load of two collinear through-wall cracks and, then, the reaction force model and plastic zone contact model were selected as the optimum ones. The objective of this study is to estimate the coalescence load of two collinear through-wall cracks in steam generator tube by using the optimum local failure prediction models. In order to investigate the applicability of the optimum local failure prediction models, a series of plastic collapse tests and corresponding finite element analyses for two collinear through-wall cracks in steam generator tube were carried out. Thereby, the applicability of the optimum local failure prediction models was verified and, finally, a coalescence evaluation diagram which can be used to determine whether the adjacent cracks detected by NDE coalesce or not has been developed

  15. 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

  16. Rejection index for pressure tubes

    International Nuclear Information System (INIS)

    Mitchell, A.B.; Meneley, D.

    1989-10-01

    The objective of the present study was to establish a set of criteria (or Rejection Index) which could be used to decide whether a zirconium-2 1/2 w/o niobium pressure tube in a CANDU reactor should be removed from service due to in-service degradation. A critique of key issues associated with establishing a realistic rejection index was prepared. Areas of uncertainty in available information were identified and recommendations for further analysis and laboratory testing made. A Rejection Index based on the following limits has been recommended: 1) Limits related to design intent and normal operation: any garter spring must remain within the tolerance band specified for its design location; the annulus gas system must normally be operated in a circulating mode with a procedure in place for purging to prevent accumulation of deuterium. It must remain sensitive to leaks into any part of the systems; and pressure tube dimensions and distortions must be limited to maintain the fuel channels within the original design intent; 2) Limits related to defect tolerance: adequate time margins between occurrence of a leaking crack and unstable failure must be demonstrated for all fuel channels; long lap-type flaws are unacceptable; crack-like defects of any size are unacceptable; and score marks, frat marks and other defects with contoured profiles must fall below certain depth, length and stress intensity limits; and 3) Limits related to property degradation: at operating temperature each pressure tube must be demonstrated to have a critical length in excess of a stipulated value; the maximum equivalent hydrogen level in any pressure tube should not exceed a limit which should be defined taking into account the known history of that tube; the maximum equivalent hydrogen level in any rolled joint should not exceed a limit which is presently recommended as 200 ppm equivalent hydrogen; and the maximum diametral creep strain should be limited to less than 5%

  17. Crack growth in an austenitic stainless steel at high temperature

    International Nuclear Information System (INIS)

    Polvora, J.P.

    1998-01-01

    This study deals with crack propagation at 650 deg C on an austenitic stainless steel referenced by Z2 CND 17-12 (316L(NN)). It is based on an experimental work concerning two different cracked specimens: CT specimens tested at 650 deg C in fatigue, creep and creep-fatigue with load controlled conditions (27 tests), tube specimens containing an internal circumferential crack tested in four points bending with displacement controlled conditions (10 tests). Using the fracture mechanics tools (K, J and C* parameters), the purpose here is to construct a methodology of calculation in order to predict the evolution of a crack with time for each loading condition using a fracture mechanics global approach. For both specimen types, crack growth is monitored by using a specific potential drop technique. In continuous fatigue, a material Paris law at 650 deg C is used to correlate crack growth rate with the stress intensity factor range corrected with a factor U(R) in order to take into account the effects of crack closure and loading ratio R. In pure creep on CT specimens, crack growth rate is correlated to the evolution of the C* parameter (evaluated experimentally) which can be estimated numerically with FEM calculations and analytically by using a simplified method based on a reference stress approach. A modeling of creep fatigue growth rate is obtained from a simple summation of the fatigue contribution and the creep contribution to the total crack growth. Good results are obtained when C* parameter is evaluated from the simplified expression C* s . Concerning the tube specimens tested in 4 point bending conditions, a simulation based on the actual A 16 French guide procedure proposed at CEA. (authors)

  18. Thermal protection system for the concrete core support floor at Fort St. Vrain

    International Nuclear Information System (INIS)

    Jones, H.; Hedgecock, P.D.

    1976-01-01

    A unique feature of the Fort St. Vrain HTGR is its steel jacketed concrete core support floor. The construction of this floor generally resembles that of the prestressed concrete reactor vessel, but its location immediately below the core hot gas outlets generates some particularly severe thermal protection requirements. A thermal barrier is used over the entire outer surface of the floor and in the twelve hot gas ducts which convey the primary coolant through the floor to the steam generators. A cooling water system of square tubes welded to the inside of the steel jacket is used to remove that heat which does pass through the thermal barrier and to maintain the concrete at acceptable temperatures. The design approach to the floor itself and to the thermal barriers and cooling system will be described, but the main emphasis of the paper will be on the total experience gained during construction and pre-operational testing. A particular problem experienced during construction was leakage from some cooling tubes, after their embedment in concrete. The solution to that problem was to develop a method for injecting catalyzed epoxy into the leaking tube. This method, which has general usefulness for in-service repairs, will be described. (author)

  19. Cracked gas generator

    Energy Technology Data Exchange (ETDEWEB)

    Abthoff, J; Schuster, H D; Gabler, R

    1976-11-17

    A small cracked-gas generator in a vehicle driven, in particular, by an air combustion engine has been proposed for the economic production of the gases necessary for low toxicity combustion from diesel fuel. This proceeds via catalytic crack-gasification and exploitation of residual heat from exhaust gases. This patent application foresees the insertion of one of the catalysts supporting the cracked-gas reaction in a container through which the reacting mixture for cracked-gas production flows in longitudinal direction. Further, air ducts are embedded in the catalyst through which exhaust gases and fresh air flow in counter direction to the cracked gas flow in the catalyst. The air vents are connected through heat conduction to the catalyst. A cracked gas constituting H/sub 2//CO/CO/sub 2//CH/sub 4/ and H/sub 2/O can be produced from the air-fuel mixture using appropriate catalysts. By the addition of 5 to 25% of cracked gas to the volume of air drawn in by the combustion engine, a more favourable combustion can be achieved compared to that obtained under normal combustion conditions.

  20. Methodology for failure assessment of SMART SG tube with once-through helical-coiled type

    International Nuclear Information System (INIS)

    Kim, Young Jin; Choi, Shin Beom; Cho, Doo Ho; Chang, Yoon Suk

    2010-09-01

    In this research project, existing integrity evaluation method for SMART steam generator tube with crack-like flaw was reviewed to determine subject analysis model and investigate possibility of failure under crack closure behavior. Furthermore, failure pressure estimation was proposed for SMART steam generator tubes containing wear-type defects. For each subject, the following issues are addressed: 1. Determination of subject analysis model for SMART SG tube contaning crack-like flaw 2. Applicability review on existing integrity evaluation method and investigation of failure possibility for SMART SG tube containing crack-like flaw 3. Development of failure pressure estimation model for SMART SG tube with wear type defect It is anticipated that if the technologies developed in this study are applied, structural integrity can be estimated accurately

  1. Ear Tubes

    Science.gov (United States)

    ... of the ear drum or eustachian tube, Down Syndrome, cleft palate, and barotrauma (injury to the middle ear caused by a reduction of air pressure, ... specialist) may be warranted if you or your child has experienced repeated ... fluid in the middle ear, barotrauma, or have an anatomic abnormality that ...

  2. Comparison of delayed hydride cracking behavior of two zirconium alloys

    Energy Technology Data Exchange (ETDEWEB)

    Ponzoni, L.M.E. [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Mieza, J.I. [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Instituto Sabato, UNSAM–CNEA, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); De Las Heras, E. [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Domizzi, G., E-mail: domizzi@cnea.gov.ar [CNEA – Centro Atómico Constituyentes, Hidrógeno en Materiales, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina); Instituto Sabato, UNSAM–CNEA, Av. Gral. Paz 1499, San Martín (B1650KNA), Bs. As. (Argentina)

    2013-08-15

    Delayed hydride cracking (DHC) is an important failure mechanism that may occur in Zr alloys during service in water-cooled reactors. Two conditions must be attained to initiate DHC from a crack: the stress intensity factor must be higher than a threshold value called K{sub IH} and, hydrogen concentration must exceed a critical value. Currently the pressure tubes for CANDU reactor are fabricated from Zr–2.5Nb. In this paper the critical hydrogen concentration for DHC and the crack velocity of a developmental pressure tube, Excel, was evaluated and compared with that of Zr–2.5Nb. The DHC velocity values measured in Excel were higher than usually reported in Zr–2.5Nb. Due to the higher hydrogen solubility limits in Excel, its critical hydrogen concentration for DHC initiation is 10–50 wppm over that of Zr–2.5Nb in the range of 150–300 °C.

  3. Constructive solutions for beamless capitalless floors with prestressed reinforcement

    Directory of Open Access Journals (Sweden)

    Bardysheva Yuliya Anatol'evna

    2014-07-01

    Full Text Available In the article the authors present advanced constructions of prestressed reinforced concrete flat ceiling, where high-strength ropes in elastic shell are used as stressed reinforcement. The novelty of the solution lays in diagonal arrangement of hard valves and use of high-strength ropes in a flexible shell of "Monostrand" type. This type of prestress, in our opinion, is the most acceptable from technical point of view for selective reinforcement of separate tense rods or cables. The use of pre-stressed reinforcement in the form of individual rods or cables increases the rigidity and crack resistance of concrete beamless slabs. The use of high-strength ropes in the monostrand-type shell makes it possible to prestress in frames of single cell plate or floor in general and to reduce labour input for stressing armature. The paper presents original solution with diagonal position of the valve. The authors suggest the use of prestressed diagonal valves as in all cells of the floor with the cells of the same or only slightly different size and in separate cells of the floor (for roofs with different cells. The diagonal location of stressed reinforcement proposed in the work is an efficient solution for extending the range of dimensions and loads size.

  4. Signal process and profile reconstruction of stress corrosion crack by eddy current test

    International Nuclear Information System (INIS)

    Zhang Siquan; Chen Tiequn; Liu Guixiong

    2008-01-01

    The reconstruction of crack profiles is very important in the NDE (nondestructive evaluation) of critical structures, such as pressure vessel and tubes in heat exchangers. First a wavelet transform signal processing technique is used to reduce noise and other non-defect signals from the signals of crack, and then based on an artificial neural network method, the crack profiles are reconstructed. Although the results reveal that this method is with many advantages such as a short CPU time and precision for reconstruction,it does have some drawbacks, for example, the database generation and network training is a much time consuming work. Moreover, this approach does not expressly reconstruct the distribution of conductivity inside a crack, so the reliability of a reconstructed crack shape is unknown. But in practical application, if we do not consider the multiple cracks, this method can be used to reconstruct the natural crack. (authors)

  5. Ploughing the deep sea floor.

    Science.gov (United States)

    Puig, Pere; Canals, Miquel; Company, Joan B; Martín, Jacobo; Amblas, David; Lastras, Galderic; Palanques, Albert

    2012-09-13

    Bottom trawling is a non-selective commercial fishing technique whereby heavy nets and gear are pulled along the sea floor. The direct impact of this technique on fish populations and benthic communities has received much attention, but trawling can also modify the physical properties of seafloor sediments, water–sediment chemical exchanges and sediment fluxes. Most of the studies addressing the physical disturbances of trawl gear on the seabed have been undertaken in coastal and shelf environments, however, where the capacity of trawling to modify the seafloor morphology coexists with high-energy natural processes driving sediment erosion, transport and deposition. Here we show that on upper continental slopes, the reworking of the deep sea floor by trawling gradually modifies the shape of the submarine landscape over large spatial scales. We found that trawling-induced sediment displacement and removal from fishing grounds causes the morphology of the deep sea floor to become smoother over time, reducing its original complexity as shown by high-resolution seafloor relief maps. Our results suggest that in recent decades, following the industrialization of fishing fleets, bottom trawling has become an important driver of deep seascape evolution. Given the global dimension of this type of fishery, we anticipate that the morphology of the upper continental slope in many parts of the world’s oceans could be altered by intensive bottom trawling, producing comparable effects on the deep sea floor to those generated by agricultural ploughing on land.

  6. Flooring for Schools: Unsightly Walkways

    Science.gov (United States)

    Baxter, Mark

    2011-01-01

    Many mattress manufacturers recommend that consumers rotate their mattresses at least twice a year to help prevent soft spots from developing and increase the product's life span. It's unfortunate that the same kind of treatment can't be applied to flooring for schools, such as carpeting, especially in hallways. Being able to flip or turn a carpet…

  7. Container floor at high temperatures

    International Nuclear Information System (INIS)

    Reutler, H.; Klapperich, H.J.; Mueller-Frank, U.

    1978-01-01

    The invention describes a floor for container which is stressed at high, changing temperatures and is intended for use in gas-cooled nuclear reactors. Due to the downward cooling gas flow in these types of reactor, the reactor floor is subjected to considerable dimensional changes during switching on and off. In the heating stage, the whole graphite structure of the reactor core and floor expands. In order to avoid arising constraining forces, sufficiently large expansion spaces must be allowed for furthermore restoring forces must be present to close the gaps again in the cooling phase. These restoring forces must be permanently present to prevent loosening of the core cuits amongst one another and thus uncontrollable relative movement. Spring elements are not suitable due to fast fatigue as a result of high temperatures and radiation exposure. It is suggested to have the floor elements supported on rollers whose rolling planes are downwards inclined to a fixed point for support. The construction is described in detail by means of drawings. (GL) [de

  8. Timber floors strengthened with concrete

    NARCIS (Netherlands)

    Blass, H.J.; Linden, M.L.R. van der; Schlager, M.

    1998-01-01

    Timber-concrete composite (tcc) beams may be used for the renovation of old timber floors. Although these systems are not new (Pokulka, 1997) and form a simple and practical solution, they are not widely adopted. One of the reasons for this is the Jack of uniform design rules. In this research

  9. Inspecting cracks in foam insulation

    Science.gov (United States)

    Cambell, L. W.; Jung, G. K.

    1979-01-01

    Dye solution indicates extent of cracking by penetrating crack and showing original crack depth clearly. Solution comprised of methylene blue in denatured ethyl alcohol penetrates cracks completely and evaporates quickly and is suitable technique for usage in environmental or structural tests.

  10. Modelling of Corrosion Cracks

    DEFF Research Database (Denmark)

    Thoft-Christensen, Palle

    Modelling of corrosion cracking of reinforced concrete structures is complicated as a great number of uncertain factors are involved. To get a reliable modelling a physical and mechanical understanding of the process behind corrosion in needed.......Modelling of corrosion cracking of reinforced concrete structures is complicated as a great number of uncertain factors are involved. To get a reliable modelling a physical and mechanical understanding of the process behind corrosion in needed....

  11. Cracking the Gender Codes

    DEFF Research Database (Denmark)

    Rennison, Betina Wolfgang

    2016-01-01

    extensive work to raise the proportion of women. This has helped slightly, but women remain underrepresented at the corporate top. Why is this so? What can be done to solve it? This article presents five different types of answers relating to five discursive codes: nature, talent, business, exclusion...... in leadership management, we must become more aware and take advantage of this complexity. We must crack the codes in order to crack the curve....

  12. Study of regularities in propagation of thermal fatigue cracks

    International Nuclear Information System (INIS)

    Tachkova, N.G.; Sobolev, N.D.; Egorov, V.I.; Rostovtsev, Yu.V.; Ivanov, Yu.S.; Sirotin, V.L.

    1978-01-01

    Regularities in the propagation of thermal fatigue cracks in the Cr-Ni steels of the austenite class depending upon deformation conditions in the crack zone, have been considered. Thin-walled tube samples of the Kh16N40, Kh18N20 and Kh16N15 steels have been tested in the 10O reversible 400 deg C and 100 reversible 500 deg C regimes. The samples have possessed a slot-shaped stress concentrator. Stress intensity pseudocoefficient has been calculated for the correlation of experimental data. The formula for determining crack propagation rate has been obtained. The experiments permit to conclude that propagation rate of thermal fatigue cracks in the above steels depends upon the scope of plastic deformation during a cycle and stress intensity pseudocoefficient, and is determined by plastic deformation resistance during thermal cyclic loading

  13. Vibration characteristics of tubes in a heat exchanger

    International Nuclear Information System (INIS)

    Simonis; Steininger, D.

    1985-01-01

    Circumferential tube cracking has occurred in the once-through steam generators used in nuclear power plants. Analyses of failed tubes indicate that a fatigue process induced by tube vibration could cause the leaks. To investigate the vibration amplitude of tube spans during reactor operation, twenty-three tube spans were instrumented with accelerometers and strain gages at Three Mile Island Unit 2. To aid in the interpretation of the operational vibration measurements, tests were performed, in air, to determine the predominant resonant frequencies and mode shapes of selected tubes. By adapting modal analysis techniques, the two predominant response frequencies were determined for 100 randomly selected tube spans and the 23 instrumented tube spans; plus, the predominant mode shape was determined for five tube spans bounded by the tube sheet and the fifteenth support plate and one tube span bounded by the ninth and tenth support plate. The average value for the first and second predominant response frequency was 65 Hz and 170 Hz, respectively. The predominant frequencies for the individual tube spans are distributed randomly with no spatial orientation. The first predominant mode shape for the six tube spans tested corresponded to a classical beam with elastic supports. The equivalent stiffness of the elastic supports depend upon the tube span tested

  14. SSRI Facilitated Crack Dancing

    Directory of Open Access Journals (Sweden)

    Ravi Doobay

    2017-01-01

    Full Text Available Choreoathetoid movement secondary to cocaine use is a well-documented phenomenon better known as “crack dancing.” It consists of uncontrolled writhing movements secondary to excess dopamine from cocaine use. We present a 32-year-old male who had been using cocaine for many years and was recently started on paroxetine, a selective serotonin reuptake inhibitor (SSRI for worsening depression four weeks before presentation. He had been doing cocaine every 2 weeks for the last three years and had never “crack danced” before this episode. The authors have conducted a thorough literature review and cited studies that suggest “crack dancing” is associated with excess dopamine. There has never been a documented case report of an SSRI being linked with “crack dancing.” The authors propose that the excess dopaminergic effect of the SSRI lowered the dopamine threshold for “crack dancing.” There is a communication with the Raphe Nucleus and the Substantia Nigra, which explains how the SSRI increases dopamine levels. This is the first documented case of an SSRI facilitating the “crack dance.”

  15. Natural zeolite bitumen cracking

    Energy Technology Data Exchange (ETDEWEB)

    Kuznicki, S.M.; McCaffrey, W.C.; Bian, J.; Wangen, E.; Koenig, A. [Alberta Univ., Edmonton, AB (Canada). Dept. of Chemical and Materials Engineering

    2006-07-01

    A study was conducted to demonstrate how low cost heavy oil upgrading in the field could reduce the need for diluents while lowering the cost for pipelining. Low cost field upgrading could also contribute to lowering contaminant levels. The performance of visbreaking processes could be improved by using disposable cracking agents. In turn, the economics of field upgrading of in-situ derived bitumen would be improved. However, in order to be viable, such agents would have to be far less expensive than current commercial cracking catalysts. A platy natural zeolite was selected for modification and testing due to its unique chemical and morphological properties. A catalyst-bearing oil sand was then heat-treated for 1 hour at 400 degrees C in a sealed microreactor. Under these mild cracking conditions, the catalyst-bearing oil sand produced extractable products of much lower viscosity. The products also contained considerably more gas oil and middle distillates than raw oil sand processed under the same conditions as thermal cracking alone. According to model cracking studies using hexadecane, these modified mineral zeolites may be more active cracking agents than undiluted premium commercial FCC catalyst. These materials hold promise for partial upgrading schemes to reduce solvent requirements in the field. tabs., figs.

  16. Electron tube

    Science.gov (United States)

    Suyama, Motohiro [Hamamatsu, JP; Fukasawa, Atsuhito [Hamamatsu, JP; Arisaka, Katsushi [Los Angeles, CA; Wang, Hanguo [North Hills, CA

    2011-12-20

    An electron tube of the present invention includes: a vacuum vessel including a face plate portion made of synthetic silica and having a surface on which a photoelectric surface is provided, a stem portion arranged facing the photoelectric surface and made of synthetic silica, and a side tube portion having one end connected to the face plate portion and the other end connected to the stem portion and made of synthetic silica; a projection portion arranged in the vacuum vessel, extending from the stem portion toward the photoelectric surface, and made of synthetic silica; and an electron detector arranged on the projection portion, for detecting electrons from the photoelectric surface, and made of silicon.

  17. Crack initiation and growth in welded structures

    International Nuclear Information System (INIS)

    Assire, A.

    2000-01-01

    This work concerns the remaining life assessment of a structure containing initial defects of manufacturing. High temperature crack initiation and growth are studied for austenitic stainless steels, and defect assessment methods are improved in order to take into account welded structures. For these one, the probability to have a defect is significant. Two kinds of approaches are commonly used for defect assessment analysis. Fracture mechanics global approach with an energetic criterion, and local approach with a model taking into account the physical damage mechanism. For both approaches mechanical fields (stress and strain) have to be computed everywhere within the structure. Then, Finite Element computation is needed. The first part of the thesis concerns the identification of non linear kinematic and isotropic constitutive models. A pseudo-analytical method is proposed for a 'Two Inelastic Strain' model. This method provides a strategy of identification with a mechanical meaning, and this enables to associate each parameter to a physical phenomenon. Existing identifications are improved for cyclic plasticity and creep on a large range of stress levels. The second part concerns high temperature crack initiation and growth in welded structures. Finite Element analysis on plate and tube experimental configuration enable to understand the phenomenons of interaction between base metal and weld metal under mechanical and thermal loading. Concerning global approach, criteria based on C* parameter (Rice integral for visco-plasticity) are used. Finite Element computations underline the fact that for a defect located in the weld metal, C* values strongly depend on the base metal creep strain rate, because widespread visco-plasticity is located in both metals. A simplified method, based on the reference stress approach, is proposed and validated with Finite Element results. Creep crack growth simplified assessment is a quite good validation of the experimental results

  18. Repair boundary for parent tube indications within the upper joint zone of hybrid expansion joint (HEJ) sleeved tubes

    International Nuclear Information System (INIS)

    Cullen, W.K.; Keating, R.F.

    1997-01-01

    In the Spring and Fall of 1994, and the Spring of 1995, crack-like indications were found in the upper hybrid expansion joint (HEJ) region of Steam Generator (S/G) tubes which had been sleeved using Westinghouse HEJ sleeves. As a result of these findings, analytic and test evaluations were performed to assess the effect of the degradation on the structural, and leakage, integrity of the sleeve/tube joint relative to the requirements of the United States Nuclear Regulatory Commission's (NRC) draft Regulatory Guide (RG) 1.121. The results of these evaluations demonstrated that tubes with implied or known crack-like circumferential parent tube indications (PTIs) located 1.1 inches or farther below the bottom of the hardroll upper transition, have sufficient, and significant, integrity relative to the requirements of RG 1.121. Thus, the purpose of this report is to provide background information related to the justification of the modified tube repair boundary

  19. Ultrasonic sizing of fatigue cracks

    International Nuclear Information System (INIS)

    Burns, D.J.

    1983-12-01

    Surface and buried fatigue cracks in steel plates have been sized using immersion probes as transmitters-receivers, angled to produce shear waves in the steel. Sizes have been estimated by identifying the ultrasonic waves diffracted from the crack tip and by measuring the time taken for a signal to travel to and from the crack tip. The effects of compression normal to a fatigue crack and of crack front curvature are discussed. Another diffraction technique, developed by UKAEA, Harwell, is reviewed

  20. Chest tube insertion

    Science.gov (United States)

    Chest drainage tube insertion; Insertion of tube into chest; Tube thoracostomy; Pericardial drain ... Be careful there are no kinks in your tube. The drainage system should always sit upright and be placed ...

  1. CAISSON TYPE HOLLOW FLOOR SLABS OF MONOLITHIC MULTI-STOREYED BUILDINGS

    Directory of Open Access Journals (Sweden)

    Malakhova Anna Nikolaevna

    2016-06-01

    Full Text Available One of the disadvantages of building structures made of reinforced concrete is their considerable weight. One of the trends to decrease the weight of concrete structures, including floor slabs, is the arrangement of voids in the cross-sectional building structures. In Russian and foreign practice paper, cardboard and plastic tubes has been used for creation of voids in the construction of monolithic floor slabs. Lightweight concretes were also used for production of precast hollow core floor slabs. The article provides constructive solutions of precast hollow core floor slabs and solid monolithic slabs that were used in the construction of buildings before wide use of large precast hollow core floor slabs. The article considers the application of caisson hollow core floor slabs for modern monolithic multi-storeyed buildings. The design solutions of such floor slabs, experimental investigations and computer modeling of their operation under load were described in this article. The comparative analysis of the calculation results of computer models of a hollow slabs formed of rod or plastic elements showed the similarity of calculation results.

  2. A phenomenological model for iodine stress corrosion cracking of zircaloy

    International Nuclear Information System (INIS)

    Miller, A.K.; Tasooji, A.

    1981-01-01

    To predict the response of Zircaloy tubing in iodine environments under conditions where either crack initiation or crack propagation predominates, a unified model of the SCC process has been developed based on the local conditions (the local stress, local strain, and local iodine concentration) within a small volume of material at the cladding inner surface or the crack tip. The methodology used permits computation of these values from simple equations. A nonuniform distribution of local stress and strain results once a crack has initiated. The local stress can be increased due to plastic constraint and triaxiality at the crack tip. Iodine penetration is assumed to be a surface diffusion-controlled process. Experimental data are used to derive criteria for intergranular failure, transgranular failure, and ductile rupture in terms of the local conditions. The same failure criteria are used for both crack initiation and crack propagation. Irradiation effects are included in the model by changing the value of constants in the equation governing iodine penetration and by changing the values used to represent the mechanical properties of the Zircaloy. (orig./HP)

  3. Alloy SCR-3 resistant to stress corrosion cracking

    International Nuclear Information System (INIS)

    Kowaka, Masamichi; Fujikawa, Hisao; Kobayashi, Taiki

    1977-01-01

    Austenitic stainless steel is used widely because the corrosion resistance, workability and weldability are excellent, but the main fault is the occurrence of stress corrosion cracking in the environment containing chlorides. Inconel 600, most resistant to stress corrosion cracking, is not necessarily safe under some severe condition. In the heat-affected zone of SUS 304 tubes for BWRs, the cases of stress corrosion cracking have occurred. The conventional testing method of stress corrosion cracking using boiling magnesium chloride solution has been problematical because it is widely different from actual environment. The effects of alloying elements on stress corrosion cracking are remarkably different according to the environment. These effects were investigated systematically in high temperature, high pressure water, and as the result, Alloy SCR-3 with excellent stress corrosion cracking resistance was found. The physical constants and the mechanical properties of the SCR-3 are shown. The states of stress corrosion cracking in high temperature, high pressure water containing chlorides and pure water, polythionic acid, sodium phosphate solution and caustic soda of the SCR-3, SUS 304, Inconel 600 and Incoloy 800 are compared and reported. (Kako, I.)

  4. A consistent partly cracked XFEM element for cohesive crack growth

    DEFF Research Database (Denmark)

    Asferg, Jesper L.; Poulsen, Peter Noe; Nielsen, Leif Otto

    2007-01-01

    Present extended finite element method (XFEM) elements for cohesive crack growth may often not be able to model equal stresses on both sides of the discontinuity when acting as a crack-tip element. The authors have developed a new partly cracked XFEM element for cohesive crack growth with extra...... enrichments to the cracked elements. The extra enrichments are element side local and were developed by superposition of the standard nodal shape functions for the element and standard nodal shape functions for a sub-triangle of the cracked element. With the extra enrichments, the crack-tip element becomes...... capable of modelling variations in the discontinuous displacement field on both sides of the crack and hence also capable of modelling the case where equal stresses are present on each side of the crack. The enrichment was implemented for the 3-node constant strain triangle (CST) and a standard algorithm...

  5. Factors affecting stress assisted corrosion cracking of carbon steel under industrial boiler conditions

    Science.gov (United States)

    Yang, Dong

    Failure of carbon steel boiler tubes from waterside has been reported in the utility boilers and industrial boilers for a long time. In industrial boilers, most waterside tube cracks are found near heavy attachment welds on the outer surface and are typically blunt, with multiple bulbous features indicating a discontinuous growth. These types of tube failures are typically referred to as stress assisted corrosion (SAC). For recovery boilers in the pulp and paper industry, these failures are particularly important as any water leak inside the furnace can potentially lead to smelt-water explosion. Metal properties, environmental variables, and stress conditions are the major factors influencing SAC crack initation and propagation in carbon steel boiler tubes. Slow strain rate tests (SSRT) were conducted under boiler water conditions to study the effect of temperature, oxygen level, and stress conditions on crack initation and propagation on SA-210 carbon steel samples machined out of boiler tubes. Heat treatments were also performed to develop various grain size and carbon content on carbon steel samples, and SSRTs were conducted on these samples to examine the effect of microstructure features on SAC cracking. Mechanisms of SAC crack initation and propagation were proposed and validated based on interrupted slow strain tests (ISSRT). Water chemistry guidelines are provided to prevent SAC and fracture mechanics model is developed to predict SAC failure on industrial boiler tubes.

  6. A crack growth evaluation method for interacting multiple cracks

    International Nuclear Information System (INIS)

    Kamaya, Masayuki

    2003-01-01

    When stress corrosion cracking or corrosion fatigue occurs, multiple cracks are frequently initiated in the same area. According to section XI of the ASME Boiler and Pressure Vessel Code, multiple cracks are considered as a single combined crack in crack growth analysis, if the specified conditions are satisfied. In crack growth processes, however, no prescription for the interference between multiple cracks is given in this code. The JSME Post-Construction Code, issued in May 2000, prescribes the conditions of crack coalescence in the crack growth process. This study aimed to extend this prescription to more general cases. A simulation model was applied, to simulate the crack growth process, taking into account the interference between two cracks. This model made it possible to analyze multiple crack growth behaviors for many cases (e.g. different relative position and length) that could not be studied by experiment only. Based on these analyses, a new crack growth analysis method was suggested for taking into account the interference between multiple cracks. (author)

  7. Control rod guide tube of nuclear reactor

    International Nuclear Information System (INIS)

    Suda, Yoshitaka; Ito, Kenji; Matsumoto, Kunio.

    1994-01-01

    Zr having a residual tensile stress of 3 to 10kg/mm 2 in a circumferential direction is used for the main ingredient of a control guide tube of a nuclear reactor. For this purpose, an appropriate correction method such as a roll-correction, tension-correction and press-correction method is applied to an existent Zr-base alloy tube with no substantial residual stress. If the residual tensile stress in the circumferential direction is smaller than 3kg/mm 2 , an effect sufficient to suppress irradiation growth is not obtainable, if it exceeds 10kg/mm 2 , dimensional changes, cracks or the like occurs locally since the wall thickness of the control rod guide tube is small and, accordingly, this often results in failed products as the control guide tube. (N.H.)

  8. 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

  9. French steam generator tubes: an overview of degradations

    International Nuclear Information System (INIS)

    Buisine, D.; Bouvier, O. de; Rupa, N.; Thebault, Y.; Barbe, V.; Pitner, P.

    2011-01-01

    The various damages (corrosion, fatigue cracks, wear, ...) observed on steam generator (SG) tubes are presented here as well as the techniques used to characterize these damages. The SG are equipped with tubes of 3 materials: 600 MA, 600 TT and 690 TT. Concerning PWSCC of 600 MA and 600 TT tubes, beyond the damages usually observed (corrosion in expansion transition zone and in 600 MA tubes small radius U-bend zone), a new event is to be noted: the phenomenon of denting (presumably induced by the deposit of sludge on the tubesheet) has induced circumferential cracking of the tube expansion transition zone. Concerning ODSCC of 600 MA tubes, beyond the classically observed damages (IGA and IGSCC in expansion transition zone and in TSP crevice), a new event is to be noted: the occurrence of circumferential cracks in tube- TSP crevice. Concerning fatigue cracking, two events have to be noted at upper TSP level in Cruas 1 and Cruas 4 units and in Fessenheim 2 unit. The first (Cruas) was due to the blockage in the broached hole tube support plate which can create critical velocity ratios for some tubes and the second (Fessenheim) to high-cycle fatigue. Concerning wear damage, beyond what is usually observed in the U-bend zone facing the anti-vibration bars (AVB), a new event is to be noted: a wear at TSP level is observed on SG equipped with an economizer, the wear indications being located at TSP 7 and 8 level, on outer tubes close to the central lane. The number of tubes plugged for ODSCC has declined due to the progressive replacement of SG with Alloy 600 MA tubing. Starting in 2004, the increasing plugging of 690 tubing is mainly due to AVB wear. Since 2006, extensive preventive plugging campaigns for tubes at risk of high-cycle fatigue at the upper support plate are performed. Risk of high-cycle fatigue has consequently become the dominant mechanism inducing plugging. PWSCC is the second dominant mechanism which affects 600 MA and 600 TT tube bundles: extensive

  10. Dynamic Response and Fracture of Composite Gun Tubes

    Directory of Open Access Journals (Sweden)

    Jerome T. Tzeng

    2001-01-01

    Full Text Available The fracture behavior due to dynamic response in a composite gun tube subjected to a moving pressure has been investigated. The resonance of stress waves result in very high amplitude and frequency strains in the tube at the instant and location of pressure front passage as the velocity of the projectile approaches a critical value. The cyclic stresses can accelerate crack propagation in the gun tube with an existing imperfection and significantly shorten the fatigue life of gun tubes. The fracture mechanism induced by dynamic amplification effects is particularly critical for composite overwrap barrels because of a multi-material construction, anisotropic material properties, and the potential of thermal degradation.

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

    International Nuclear Information System (INIS)

    Dvorsek, T.; Cizelj, L.

    1995-01-01

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

  12. 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.)

  13. Eddy Current Signature Classification of Steam Generator Tube Defects Using A Learning Vector Quantization Neural Network

    International Nuclear Information System (INIS)

    Garcia, Gabe V.

    2005-01-01

    A major cause of failure in nuclear steam generators is degradation of their tubes. Although seven primary defect categories exist, one of the principal causes of tube failure is intergranular attack/stress corrosion cracking (IGA/SCC). This type of defect usually begins on the secondary side surface of the tubes and propagates both inwards and laterally. In many cases this defect is found at or near the tube support plates

  14. Linear Alkylbenzenesulfonates in indoor Floor Dust

    DEFF Research Database (Denmark)

    Madsen, Jørgen Øgaard; Wolkoff, Peder; Madsen, Jørgen Øgaard

    1999-01-01

    The amount of Linear Alkylbenzenesulfonates (LAS) in the particle fraction of floor dust sampled from 7 selected public buildings varied between 34 and 1500 microgram per gram dust, while the contents of the fibre fractions generally were higher with up to 3500 microgram LAS/g dust. The use...... of a cleaning agent with LAS resulted in an increase of the amount of LAS in the floor dust after floor wash relative to just before floor wash. However, the most important source of LAS in the indoor floor dust appears to be residues of detergent in clothing. Thus, a newly washed shirt contained 2960 microgram...

  15. Integrated modeling and characterization of local crack chemistry

    International Nuclear Information System (INIS)

    Savchik, J.A.; Burke, M.S.

    1995-01-01

    The MULTEQ computer program has become an industry wide tool which can be used to calculate the chemical composition in a flow occluded region as the solution within concentrates due to a local boiling process. These results can be used to assess corrosion concerns in plant equipment such as steam generators. Corrosion modeling attempts to quantify corrosion assessments by accounting for the mass transport processes involved in the corrosion mechanism. MULTEQ has played an ever increasing role in defining the local chemistry for such corrosion models. This paper will outline how the integration of corrosion modeling with the analysis of corrosion films and deposits can lead to the development of a useful modeling tool, wherein MULTEQ is interactively linked to a diffusion and migration transport process. This would provide a capability to make detailed inferences of the local crack chemistry based on the analyses of the local corrosion films and deposits inside a crack and thus provide guidance for chemical fixes to avoid cracking. This methodology is demonstrated for a simple example of a cracked tube. This application points out the utility of coupling MULTEQ with a mass transport process and the feasibility of an option in a future version of MULTEQ that would permit relating film and deposit analyses to the local chemical environment. This would increase the amount of information obtained from removed tube analyses and laboratory testing that can contribute to an overall program for mitigating tubing and crevice corrosion

  16. Integrated modeling and characterization of local crack chemistry

    International Nuclear Information System (INIS)

    Savchik, J.A.; Burke, M.S.

    1996-01-01

    The MULTEQ computer program has become an industry wide tool which can be used to calculate the chemical composition in a flow occluded region as the solution within concentrates due to a local boiling process. These results can be used to assess corrosion concerns in plant equipment such as steam generators. Corrosion modeling attempts to quantify corrosion assessments by accounting for the mass transport processes involved in the corrosion mechanism. MULTEQ has played an ever increasing role in defining the local chemistry for such corrosion models. This paper will outline how the integration of corrosion modeling with the analysis of corrosion films and deposits can lead to the development of a useful modeling tool, wherein MULTEQ is interactively linked to a diffusion and migration transport process. This would provide a capability to make detailed inferences of the local crack chemistry based on the analyses of the local corrosion films and deposits inside a crack and thus provide guidance for chemical fixes to avoid cracking. This methodology is demonstrated for a simple example of a cracked tube. This application points out the utility of coupling MULTEQ with a mass transport process and the feasibility of an option in a future version of MULTEQ that would permit relating film and deposit analyses to the local chemical environment. This would increase the amount of information obtained from removed tube analyses and laboratory testing that can contribute to an overall program for mitigating tubing and crevice corrosion

  17. Nonlinear crack mechanics

    International Nuclear Information System (INIS)

    Khoroshun, L.P.

    1995-01-01

    The characteristic features of the deformation and failure of actual materials in the vicinity of a crack tip are due to their physical nonlinearity in the stress-concentration zone, which is a result of plasticity, microfailure, or a nonlinear dependence of the interatomic forces on the distance. Therefore, adequate models of the failure mechanics must be nonlinear, in principle, although linear failure mechanics is applicable if the zone of nonlinear deformation is small in comparison with the crack length. Models of crack mechanics are based on analytical solutions of the problem of the stress-strain state in the vicinity of the crack. On account of the complexity of the problem, nonlinear models are bason on approximate schematic solutions. In the Leonov-Panasyuk-Dugdale nonlinear model, one of the best known, the actual two-dimensional plastic zone (the nonlinearity zone) is replaced by a narrow one-dimensional zone, which is then modeled by extending the crack with a specified normal load equal to the yield point. The condition of finite stress is applied here, and hence the length of the plastic zone is determined. As a result of this approximation, the displacement in the plastic zone at the abscissa is nonzero

  18. photomultiplier tubes

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  19. photomultiplier tube

    CERN Multimedia

    photomultiplier tubes. A device to convert light into an electric signal (the name is often abbreviated to PM). Photomultipliers are used in all detectors based on scintillating material (i.e. based on large numbers of fibres which produce scintillation light at the passage of a charged particle). A photomultiplier consists of 3 main parts: firstly, a photocathode where photons are converted into electrons by the photoelectric effect; secondly, a multiplier chain consisting of a serie of dynodes which multiply the number of electron; finally, an anode, which collects the resulting current.

  20. Integrity evaluation of Alloy 600 RV head penetration tubes in Korean PWR plants

    International Nuclear Information System (INIS)

    Kang, Young Hwan; Park, Sung Ho; Hong, Sung Yull; Choi, Kwang Hee

    1995-01-01

    The structural integrity assessment of Alloy 600 RV head penetration tubes has been an important issue for the economical and reliable operation of power plants. In this paper, an overview of the integrity evaluation program for the RV head penetration tubes in Korean nuclear power plants is presented. Since the crack growth mechanism of the penetration tube is due to the primary water stress corrosion cracking (PWSCC) which is mainly related to the stress at the tube, the present paper consists of three primary activities: the stress evaluation, the flaw evaluation, and data generation through material and mechanical tests. (author). 5 refs, 2 figs, 1 tab

  1. Assesment of integrity of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Brozova, A.; Zdarek, J.

    1992-01-01

    Full text: The leak rates measurement project was held to give experimental data enabling the Czechoslovak Atomic Agency Inspection to decree the change in the Technical Specification allowable limit of steam generator activity release on secondary side. The WWER types of nuclear power plants in Czechoslovakia have horizontal steam generators. The tubes studying in frame of the project belong to steam generator WWER- 440 type, the diameter of tube is 16 mm, the wall thickness 1.4 mm. The subject of the project was the measurement of service leak rates of typical in service cracks. Secondary side stress corrosion cracks were determined as the typical crack created in service condition. These cracks were prepared in tubes artificially by exposition in chloride environment accompanied by an internal stress. The experimental device consisted of a pressure vessel connected with pressure water loop, a cooling vessel for leakage medium and a measuring vessel. The leak rates were determined as a slope of plots the leakage volume - time. Inside the pressure vessel the steam generator operation environment was simulated. It means: primary side of tube 12.5 MPa, Z90 deg. C, secondary side -4.6MPa, 250 deg. C, water service quality. We observed reduce of leak rate in course of time in each experiment. We suppose the tubes were stopped up by deposits formed in manufacturing of crack and in experiment. Our opinion has been proved by fractography. Project results in recommendation for in service leak rate limit based on safety factors with respect to critical crack lengths and for determination of tube plugging criteria. (author)

  2. Statistical crack mechanics

    International Nuclear Information System (INIS)

    Dienes, J.K.

    1993-01-01

    Although it is possible to simulate the ground blast from a single explosive shot with a simple computer algorithm and appropriate constants, the most commonly used modelling methods do not account for major changes in geology or shot energy because mechanical features such as tectonic stresses, fault structure, microcracking, brittle-ductile transition, and water content are not represented in significant detail. An alternative approach for modelling called Statistical Crack Mechanics is presented in this paper. This method, developed in the seventies as a part of the oil shale program, accounts for crack opening, shear, growth, and coalescence. Numerous photographs and micrographs show that shocked materials tend to involve arrays of planar cracks. The approach described here provides a way to account for microstructure and give a representation of the physical behavior of a material at the microscopic level that can account for phenomena such as permeability, fragmentation, shear banding, and hot-spot formation in explosives

  3. SCC testing of steam generator tubes repaired by welded sleeves

    International Nuclear Information System (INIS)

    Pierson, E.; Stubbe, J.

    1993-01-01

    One way to repair steam generator tubing is to introduce a sleeve inside the tube so that it spans the corroded area and to seal it at both ends. This technique has been studied at Laborelec with a particular attention paid to the occurrence of new SCC cracks at the upper joint. Tube segments coming from the same lot of mill annealed alloy 600 were sent to six manufacturers to be sleeved by their own procedure (including TIG, laser or kinetic welding, followed or not by a stress relief heat treatment), and then tested at Laborelecin 10% NaOH at 350 degrees C. The tests were performed with and without differential pressure i.e. in capsules (Δ = 9 and 19 MPa) and in autoclave (Δp = 0). Nearly all the not stress relieved mock-ups developed through cracks in several hundred hours in auto-clave. The cracks were circumferential and situated near the weld. At 9 and 19 MPa, the time to failure decreased and longitudinal cracks appeared near the weld and at the transition zone of expanded areas. Cracks were never observed in the alloy 690 sleeve, except in the weld bead. Reference capsules (roll expaned tubes) made of the same lot of alloy 600 were tested in the same environment

  4. Cracking the Cipher Challenge

    CERN Document Server

    CERN. Geneva. Audiovisual Unit; Singh, Simon

    2002-01-01

    In the back of 'The Code Book', a history of cryptography, Simon Singh included a series of 10 encoded messages, each from a different period of history. The first person to crack all 10 messages would win a prize of £10,000. Now that the prize has been won, Simon can reveal the story behind the Cipher Challenge. Along the way he will show how mathematics can be used to crack codes, the role it played in World War Two and how it helps to guarantee security in the Information Age.

  5. 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

  6. 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.

  7. Application of the leak-before-break concept to steam generator tubes

    International Nuclear Information System (INIS)

    Keim, E.; Kastner, W.

    1994-01-01

    The Leak-Before-Break (LBB) behaviour of a piping component means that the length of a crack resulting in a leak is smaller than the critical crack length and that the leak is safety detectable by a suitable monitoring system. The LBB-concept of Siemens/KWU is based on computer codes for the evaluation of critical crack lengths, crack openings, leakage areas and leakage rates, developed by Siemens/KWU. The fracture mechanics analysis supplies the input for the thermal-hydraulic analysis. The resulting leakage rate related to the crack length of a longitudinal or circumferential crack and the minimum detectable values of leakage rate and crack length lead to two criteria, which allow for the LBB-behaviour of the pipe: - the critical crack length must be larger than the crack length being safety detected by leakage monitoring systems (LMS) - the critical crack length must be larger than the crack length being safety detected by non-destructive examination (NDE). This LBB-concept is applied to steam generator (SG) tubes. Two examples, which will be presented, show that this concept is a very useful and effective tool which allows the prediction of LBB-behaviour of SG tubes. (Author)

  8. Resolution of lava tubes with ground penetrating radar: preliminary results from the TubeX project

    Science.gov (United States)

    Esmaeili, S.; Kruse, S.; Garry, W. B.; Whelley, P.; Young, K.; Jazayeri, S.; Bell, E.; Paylor, R.

    2017-12-01

    As early as the mid 1970's it was postulated that planetary tubes or caves on other planetary bodies (i.e., the Moon or Mars) could provide safe havens for human crews, protect life and shield equipment from harmful radiation, rapidly fluctuating surface temperatures, and even meteorite impacts. What is not clear, however, are the exploration methods necessary to evaluate a potential tube-rich environment to locate suitable tubes suitable for human habitation. We seek to address this knowledge gap using a suite of instruments to detect and document tubes in a terrestrial analog study at Lava Beds National Monument, California, USA. Here we describe the results of ground penetrating radar (GPR) profiles and light detection and ranging (LiDAR) scans. Surveys were conducted from the surface and within four lava tubes (Hercules Leg, Skull, Valentine and, Indian Well Caves) with varying flow composition, shape, and complexity. Results are shown across segments of these tubes where the tubes are 10 m in height and the ceilings are 1 - 10 m below the surface. The GPR profiles over the tubes are, as expected, complex, due to scattering from fractures in roof material and three-dimensional heterogeneities. Point clouds derived from the LiDAR scans of both the interior and exterior of the lava tubes provide precise positioning of the tube geometry and depth of the ceiling and floor with respect to the surface topography. GPR profiles over LiDAR-mapped tube cross-sections are presented and compared against synthetic models of radar response to the measured geometry. This comparison will help to better understand the origins of characteristic features in the radar profiles. We seek to identify the optimal data processing and migration approaches to aid lava tube exploration of planetary surfaces.

  9. Microstructural modelling of creep crack growth from a blunted crack

    NARCIS (Netherlands)

    Onck, P.R.; Giessen, E. van der

    1998-01-01

    The effect of crack tip blunting on the initial stages of creep crack growth is investigated by means of a planar microstructural model in which grains are represented discretely. The actual linking-up process of discrete microcracks with the macroscopic crack is simulated, with full account of the

  10. The effect of texture on delayed hydride cracking in Zr-2.5Nb alloy

    Energy Technology Data Exchange (ETDEWEB)

    Resta Levi, R.; Sagat, S

    1999-09-01

    Pressure tubes for CANDU reactors are made of Zr-2.5Nb alloy. They are produced by hot extrusion followed by cold work, which results in a material with a pronounced crystallographic texture with basal plane normals of its hexagonal structure around the circumferential direction. Under certain conditions, this material is susceptible to a cracking mechanism called delayed hydride cracking (DHC). Our work investigated the susceptibility of Zr-2.5Nb alloy pressure tube to DHC in this pressure tube material, in terms of crystallographic texture and grain shape. The results are presented in terms of crack velocity obtained on different planes and directions of the pressure tube. The results show that it is more difficult for a crack to propagate at right angles to crystallographic basal planes (which are close to the precipitation habit plane of hydrides) than for it to propagate parallel to the basal plane. However, if the cracking plane is oriented parallel to preexisting hydrides (hydrides formed as a result of the manufacturing process), the crack propagates along these hydrides easily, even if the hydride habit planes are not oriented favourably. (author)

  11. [Functional aspects of pelvic floor surgery].

    Science.gov (United States)

    Wagenlehner, F M E; Gunnemann, A; Liedl, B; Weidner, W

    2009-11-01

    Pelvic floor dysfunctions are frequently seen in females. The human pelvic floor is a complex structure and heavily stressed throughout female life. Recent findings in the functional anatomy of the pelvic floor have led to a much better understand-ing, on the basis of which enormous improvements in the therapeutic options have arisen. The pelvic floor activity is regulated by three main muscular forces that are responsible for vaginal tension and suspension of the pelvic floor -organs, bladder and rectum. For different reasons laxity in the vagina or its supporting ligaments as a result of altered connective tissue can distort this functional anatomy. A variety of symptoms can derive from these pelvic floor dysfunctions, such as urinary urge and stress incontinence, abnormal bladder emptying, faecal incontinence, obstructive bowel disease syndrome and pelvic pain. Pelvic floor reconstruction is nowadays driven by the concept that in the case of pelvic floor symptoms restoration of the anatomy will translate into restoration of the physiology and ultimately improve the patients' symptoms. The exact surgical reconstruction of the anatomy is there-fore almost exclusively focused on the restoration of the lax pelvic floor ligaments. An exact identification of the anatomic lesions preoperatively is eminently necessary, to allow for an exact anatomic reconstruction with respect to the muscular forces of the pelvic floor. Georg Thieme Verlag Stuttgart * New York.

  12. J-resistance curves for Inconel 690 and Incoloy 800 nuclear steam generators tubes at room temperature and at 300 °C

    Energy Technology Data Exchange (ETDEWEB)

    Bergant, Marcos A., E-mail: marcos.bergant@cab.cnea.gov.ar [Gerencia CAREM, Centro Atómico Bariloche (CNEA), Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Yawny, Alejandro A., E-mail: yawny@cab.cnea.gov.ar [División Física de Metales, Centro Atómico Bariloche (CNEA) / CONICET, Av. Bustillo 9500, San Carlos de Bariloche 8400 (Argentina); Perez Ipiña, Juan E., E-mail: juan.perezipina@fain.uncoma.edu.ar [Grupo Mecánica de Fractura, Universidad Nacional del Comahue / CONICET, Buenos Aires 1400, Neuquén 8300 (Argentina)

    2017-04-01

    The structural integrity of steam generator tubes is a relevant issue concerning nuclear plant safety. In the present work, J-resistance curves of Inconel 690 and Incoloy 800 nuclear steam generator tubes with circumferential and longitudinal through wall cracks were obtained at room temperature and 300 °C using recently developed non-standard specimens' geometries. It was found that Incoloy 800 tubes exhibited higher J-resistance curves than Inconel 690 for both crack orientations. For both materials, circumferential cracks resulted into higher fracture resistance than longitudinal cracks, indicating a certain degree of texture anisotropy introduced by the tube fabrication process. From a practical point of view, temperature effects have found to be negligible in all cases. The results obtained in the present work provide a general framework for further application to structural integrity assessments of cracked tubes in a variety of nuclear steam generator designs. - Highlights: •Non-standard fracture specimens were obtained from nuclear steam generator tubes. •Specimens with circumferential and longitudinal through-wall cracks were used. •Inconel 690 and Incoloy 800 steam generator tubes were tested at 24 and 300 °C. •Fracture toughness for circumferential cracks was higher than for longitudinal cracks. •Incoloy 800 showed higher fracture toughness than Inconel 690 steam generator tubes.

  13. Linear Cracking in Bridge Decks

    Science.gov (United States)

    2018-03-01

    Concrete cracking in bridge decks remains an important issue relative to deck durability. Cracks can allow increased penetration of chlorides, which can result in premature corrosion of the reinforcing steel and subsequent spalling of the concrete de...

  14. Improvement of Eddy Current testing methods of steam generator tubings due to field experience

    International Nuclear Information System (INIS)

    Comby, R.; Meurgey, P.; David, B.

    1985-01-01

    This paper presents the main stages of the long rotating probe developed by EDF, this probe detects stress corrosion cracks. The method has been validated by the examination of numerous cracked tubes that the probe detected before. Methods to better characterize the signals with regard to the defects are being improved to avoid a complementary examination of the rolling zone more particularly [fr

  15. Evaluation of stress intensity factor for craks in surface of tubes with internal pressure

    International Nuclear Information System (INIS)

    Cesari, F.; Hellen, T.K.

    1977-01-01

    In this report the authors have examined the different methods for calculation of the stress intensity factor in tubes subject at internal pressure with surface cracks. The analysis includes cracks in 2-D axialsymmetric and 3-D. Moreover the authors have clarified the difference between the ASME Sec.11 and the procedure more rigorous

  16. Development of a crack growth analysis is program for reactor head penetration

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Yull; Choi, Kwang Hee; Park, Jeong Il [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Kang, Young Hwan; Park, Sung Ho; Kim, Il; Kim, Young Jong; Yoo, Young Joon; Yoo, Wan; Maeng, Wan Young; Choi, Suk Nam; Kim, Kee Suk; Yoon, Sung Won; Kim, Jee Ho; Park, Myung Kyu [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    Crack growth analysis program for Reactor Head Penetration is being developed for applying to plants such as, Kori 1, Kori 2, Kori 3,4 YoungKwang 1,2 and Uljin 1,2 (1) Stress Evaluation - The stress analysis is required to evaluate the structure integrity for the RVH penetration tubes. The RVH penetration tubes are geometrically non-symmetry except center one. Thus, 3D finite element analysis should be employed for the stress analysis. The magnitude and distribution of residual stress resulted from welding can be determined analytically by simulation welding procedure. (2) Flaw Evaluation - There are two objectives of the penetration tube flaw evaluation to predict the time required for a crack to propagate to the acceptance criteria. The first objective is to perform the parametric evaluation for a postulated crack. The second objective is to develop the flaw evaluation program for the crack detected during the inspection. (3) Characterization of Material Properties of Alloy 600 - These study is to provide data which similarly represent the properties of PWR power plants in Korea. The data is used for analyzing of the stress distribution around penetration tubes. And the PWSCC data will be used for the crack growth rate of the penetration tubes. (author). 92 refs., 121 figs.

  17. Crack detection '86

    International Nuclear Information System (INIS)

    1986-01-01

    The participants of the conference heard 36 papers of which 13 were incorporated in INIS. The incorporated papers deal with the quality control of the equipment of nuclear power plants, with technical specifications and possibilities of diverse crack detection devices, as well as with personnel training for nondestructive materials testing. (E.S.)

  18. Examination of steam generator alloy 800 NG tube from the Almaraz unit 2 NPP

    International Nuclear Information System (INIS)

    Diego, G. de; Gomez Briceno, D.; Maffiotte, C.; Baladia, M.; Arias, C.J.

    2015-01-01

    The steam generators of Almaraz Unit 2 were replaced in 1997 by the model 61W/D3 (Siemens) with Alloy 800NG steam generator tubes. Denting indications were firstly detected in 2006 in the SG-3. Crack indications were identified in 2009. At the end of 2011, three tubes were recovered from this steam generator to carry out destructive examination in order to identify the root cause of the tubes degradation. Analysis of deposits point out the existence of multiples elements in the removed OD (Outer Diameter) deposits as well as in the deposits at the free tube under sludge and at the transition zone. Deposits are more abundant at the transition zone than at free tube. About 10% Na concentration has been detected, whereas S and Cl appear in small concentrations. Si appears regularly and Cr, Ni concentrations in the deposits are similar. Multiple intergranular cracks have been detected at 3 mm above the last contact point between the tube and the TS (tube support), in a band of around 5 mm, practically in the whole perimeter of the tube. Fracture surface of crack-B was partially covered by a Si rich layer, whereas fracture surface of crack-A seems to be cleaner. However, no significant differences in composition, except higher amount of S in crack-B, were found in the deposits of both cracks. EDX mapping and Auger profiles point out Ni enrichment with slight Cr enrichment or depletion and Fe depletion. The comparison of Auger profiles with available results for Alloy 800 tested in caustic and acid sulfate environments seems to indicate that the environment inside the cracks detected in the tube R67C48 is neutral or moderately caustic

  19. PWR steam generators tube integrity: plugging criteria for PWSCC in roll transition zone

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Cruz, Julio R.B.

    1999-01-01

    One of the most important causes for tube plugging in PWR (Pressurized Water Reactor) steam generators is the degradation mechanism called Primary Water Stress Corrosion Cracking (PWSCC) in roll transition zone (RTZ) near the tubesheet, mainly for Alloy 600 tubes. To avoid an excessive tube plugging, alternative criteria have been developed based on an approach that consists in withdrawing from service any tube containing a defect for which there is a high probability of a critical size under accident conditions to be reached during next operation cycle. Predictions of the number of tubes to be plugged can be done aiming at preventive maintenance and tube repair, and even a steam generator replacement, without a large and non-planned plant outage. This work presents important aspects related to tube plugging criteria for PWSCC in RTZ based on the risk of break after a leak detection. Calculations of allowable crack length and allowable leak rate for a particular situation are also shown. (author)

  20. Automated numerical simulation of cracked plates, pipes and elbows

    International Nuclear Information System (INIS)

    Reddy, Babu; Sreehari Kumar, B.; Bhate, S.R.; Kushwaha, H.S.

    2008-01-01

    In the nuclear industry, piping components are one of the key elements participating in its operation. Integrity of structural tubes and pipes plays a major role in nuclear power plants. The ideal procedure to ensure this aspect would be to conduct experimental studies on pilot/test specimens. However, it may not always be feasible to carry out the experimental investigation, as it requires pre-requisite infrastructure which may not be economically viable. This makes it imperative to conduct numerical simulations of the same particularly in the study of presence of cracks in the critical components. While performing the effect of cracks, the quality of the finite element mesh nearer to the crack tip plays a critical role while estimating J-integral value. The designer is often familiar with design methodology only and he obviously requires a convenient and reliable numerical tool to model and perform the analysis. In this context, an effort has been made in NISA, the general purpose finite element software, to automate the generation of FE meshes for a set of pre-defined components with different crack configurations. To simplify the procedure of FE mesh generation, analysis, and post processing, a graphical user interface (GUI) has been developed accordingly. This paper discusses the automated numerical simulation of plates and pipes with different crack configurations. This simulation software is also designed to help parametric study of cracked pipes. (author)

  1. Crack closure, a literature study

    Science.gov (United States)

    Holmgren, M.

    1993-08-01

    In this report crack closure is treated. The state of the art is reviewed. Different empirical formulas for determining the crack closure are compared with each other, and their benefits are discussed. Experimental techniques for determining the crack closure stress are discussed, and some results from fatigue tests are also reported. Experimental data from the literature are reported.

  2. Mode of delivery and Pelvic floor disorder

    International Nuclear Information System (INIS)

    Noor, R.; Neelam, H.; Bashir, M.S.

    2017-01-01

    Objective: To compare pelvic floor dysfunction in non pregnant women who had delivered vaginally versus those with cesarean delivery. Methodology: The prevalence of pelvic floor disorders among non pregnant women was assesses by using a standardized tool pelvic floor distress inventory short form (PFDI-20). Data was collected from Jinnah Hospital Lahore, Pakistan. Results: Total numbers of participants were 278. 47.12% subjects had moderate, 36.69% miner and 16.19% had severe pelvic floor dysfunction. The symptoms of pelvic organ prolapse were more prevalent (mean value is 59.1876) than Urinary Distress (mean value is 40.5426), while the Colorectal-Anal Distress (mean value is 35.9150) were least prevalent. Conclusion: Pelvic floor disorders are very common among females and are strongly associated with mode of delivery. Although spontaneous vaginal birth was extensively associated with pelvic floor disorders the instrumental delivery affects most. (author)

  3. Eddy current detection of corrosion damage in heat exchanger tubes

    International Nuclear Information System (INIS)

    Van Drunen, G.; Cecco, V.S.; Carter, J.R.

    1980-05-01

    Eddy current is often the most effective nondestructive test method available for in-service inspection of small bore tubing in heat exchangers. The basic principles, advantages and shortcomings of the technique are outlined. Typical eddy current indications from corrosion-related defects such as stress corrosion cracks, pitting and tube denting under support plates are presented. Eddy current signals from features such as magnetite deposits and ferromagnetic inclusions which might be mistaken for defects are also discussed. (auth)

  4. Crack resistance curve determination of zircaloy-4 cladding

    International Nuclear Information System (INIS)

    Bertsch, J.; Alam, A.; Zubler, R.

    2009-03-01

    Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 o C and 350 o C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could be

  5. Crack resistance curve determination of zircaloy-4 cladding

    Energy Technology Data Exchange (ETDEWEB)

    Bertsch, J.; Alam, A.; Zubler, R

    2009-03-15

    Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 {sup o}C and 350 {sup o}C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could

  6. Modelling of the bending behaviour of double floor systems for different contact surfaces

    Directory of Open Access Journals (Sweden)

    Attila PUSKAS

    2014-07-01

    Full Text Available In the practice of prefabricated concrete structures considerable surfaces of intermediate floors are constructed using double floor systems with prefabricated bottom layer and upper layer. This second layer is cast on site. The quality of the prefabricated concrete is often of superior class with respect to the monolithic layer. In the service state of the double floor system, important compressive stresses appear in the upper concrete layer. On the other hand, the bond quality between the concrete layers cast in successive stages raises questions especially in the case of hollow core floor units with no connecting reinforcement in-between. The paper presents results of the numerical models prepared for double floor elements having different thicknesses for the top and bottom layers, subjected to bending. Three situations have been studied: stepped top surface of the prefabricated slab with no connecting reinforcement, broom swept tracks on the prefabricated slab with no connecting reinforcement and broom swept tracks on the prefabricated slab with stirrups connecting the concrete layers. For each situation two different ratios of the thicknesses of the layers have been considered. The results are emphasizing the critical regions of the elements, the differences in crack development and in the behaviour resulting from surface preparation and use of connecting reinforcements.

  7. Environmentally assisted cracking in Light Water Reactors

    International Nuclear Information System (INIS)

    Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

    1993-09-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

  8. Magnetic resonance imaging of pelvic floor dysfunction.

    Science.gov (United States)

    Lalwani, Neeraj; Moshiri, Mariam; Lee, Jean H; Bhargava, Puneet; Dighe, Manjiri K

    2013-11-01

    Pelvic floor dysfunction is largely a complex problem of multiparous and postmenopausal women and is associated with pelvic floor or organ descent. Physical examination can underestimate the extent of the dysfunction and misdiagnose the disorders. Functional magnetic resonance (MR) imaging is emerging as a promising tool to evaluate the dynamics of the pelvic floor and use for surgical triage and operative planning. This article reviews the anatomy and pathology of pelvic floor dysfunction, typical imaging findings, and the current role of functional MR imaging. Copyright © 2013 Elsevier Inc. All rights reserved.

  9. Eustachian tube patency

    Science.gov (United States)

    Eustachian tube patency refers to how much the eustachian tube is open. The eustachian tube runs between the middle ear and the throat. It controls the pressure behind the eardrum and middle ear space. This helps keep ...

  10. Feeding tube - infants

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/007235.htm Feeding tube - infants To use the sharing features on this page, please enable JavaScript. A feeding tube is a small, soft, plastic tube placed ...

  11. Valuation of fissured steam generator tubes at the level of the roll transition area, repaired by nickel plating

    International Nuclear Information System (INIS)

    Laire, C.; Stubbe, J.; Slama, G.; Michaut, M.; Anxionnaz-Steltzlen, F.; Leblois

    1990-01-01

    At DOEL 2, SG-tubes cracked at the roll transition area were repaired by nickel plating in 1985 and in 1986 by Laborelec and Framatome using different process parameters. The characteristics of these different deposits and their service behaviour were investigated on tubes pulled out after 1 or 2 cycles. It is confirmed that this repair technique can be used for through wall cracked tubes, when: - the cracks are not too broad; - the deposit is of good quality, free of irregularities due to deposition. After this expertise the improvement of the plating procedure was focused on ductile nickel without initial deposit defects [fr

  12. Simulation of inter- and transgranular crack propagation in polycrystalline aggregates due to stress corrosion cracking

    International Nuclear Information System (INIS)

    Musienko, Andrey; Cailletaud, Georges

    2009-01-01

    The motivation of the study is the development of a coupled approach able to account for the interaction between environment and plasticity in a polycrystalline material. The paper recalls first the constitutive equations used to describe the behavior of the grain core and of the grain boundary (GB). The procedure that is applied to generate synthetic polycrystalline aggregates with an explicit representation of the grain boundary area by 2D or 3D finite elements is then described. The approach is applied to the modeling of iodine-assisted stress corrosion cracking (IASCC) in Zircaloy tubes used in nuclear power plants.

  13. Minimize corrosion degradation of steam generator tube materials

    International Nuclear Information System (INIS)

    Lu, Y.

    2006-01-01

    As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Experimental data suggest that all steam generator tube materials are susceptible to corrosion degradation under some specific off-specification conditions. The tolerance to the chemistry upset for each steam generator tube alloy is different. Electrochemical corrosion behaviors of major steam generator tube alloys were studied under the plausible aggressive crevice chemistry conditions. The potential hazardous conditions leading to steam generator tube degradation and the conditions, which can minimize steam generator tube degradation have been determined. Recommended electrochemical corrosion potential/pH zones were defined for all major steam generator tube materials, including Alloys 600, 800, 690 and 400, under CANDU steam generator operating and startup conditions. Stress corrosion cracking tests and accelerated corrosion tests were carried out to verify and revise the recommended electrochemical corrosion potential/pH zones. Based on this information, utilities can prevent steam generator material degradation surprises by appropriate steam generator water chemistry management and increase the reliability of nuclear power generating stations. (author)

  14. Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR)

    International Nuclear Information System (INIS)

    Fuentes C, P.

    2003-01-01

    This work presents the results of the assays carried out in the Laboratory of Hot Cells of the National Institute of Nuclear Research (ININ) to a type test tube Compact Tension (CT), built in steel austenitic stainless type 304L, simulating those conditions those that it operates a Boiling Water Reactor (BWR), at temperature 288 C and pressure of 8 MPa, to determine the speed to which the cracks spread in this material that is of the one that different components of a reactor are made, among those that it highlights the reactor core vessel. The application of the Hydrogen Chemistry of the Water is presented (HWC) that is one alternative to diminish the corrosion effect low stress in the component, this is gets controlling the quantity of oxygen and of hydrogen as well as the conductivity of the water. The rehearsal is made following the principles of the Mechanics of Elastic Lineal Fracture (LEFM) that considers a crack of defined size with little plastic deformation in the tip of this; the measurement of crack advance is continued with the technique of potential drop of direct current of alternating signal, this is contained inside the standard Astm E-647 (Method of Test Standard for the Measurement of Speed of Growth of Crack by fatigue) that is the one that indicates us as carrying out this test. The specifications that should complete the test tubes that are rehearsed as for their dimensions, it forms, finish and determination of mechanical properties (tenacity to the fracture mainly) they are contained inside the norm Astm E-399, the one which it is also based on the principles of the fracture mechanics. The obtained results were part of a database to be compared with those of other rehearsals under different conditions, Normal Chemistry of the Water (NWC) and it dilutes with high content of O 2 ; to determine the conditions that slow more the phenomena of stress corrosion cracking, as well as the effectiveness of the used chemistry and of the method of

  15. A reliability index for assessment of crack profile reconstructed from ECT signals using a neural-network approach

    International Nuclear Information System (INIS)

    Yusa, Noritaka; Chen, Zhenmao; Miya, Kenzo; Cheng, Weiying

    2002-01-01

    This paper proposes a reliability parameter to enhance an version scheme developed by authors. The scheme is based upon an artificial neural network that simulates mapping between eddy current signals and crack profiles. One of the biggest advantages of the scheme is that it can deal with conductive cracks, which is necessary to reconstruct natural cracks. However, it has one significant disadvantage: the reliability of reconstructed profiles was unknown. The parameter provides an index for assessment of the crack profile and overcomes this disadvantage. After the parameter is validated by reconstruction of simulated cracks, it is applied to reconstruction of natural cracks that occurred in steam generator tubes of a pressurized water reactor. It is revealed that the parameter is applicable to not only simulated cracks but also natural ones. (author)

  16. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

    Bergant M; Yawny, A; Perez Ipina, J

    2012-01-01

    The assessment of the structural integrity of steam generator tubes in nuclear power plants deserved increasing attention in the last years due to the negative impact related to their failures. In this context, elastic plastic fracture mechanics (EPFM) methodology appears as a potential tool for the analysis. The application of EPFM requires, necessarily, knowledge of two aspects, i.e., the driving force estimation in terms of an elastic plastic toughness parameter (e.g., J) and the experimental measurement of the fracture toughness of the material (e.g., the material J-resistance curve). The present work describes the development of a non standardized experimental technique aimed to determine J-resistance curves for steam generator tubes with circumferential through wall cracks. The tubes were made of Incoloy 800 (Ni: 30.0-35.0; Cr: 19.0-23.0; Fe: 35.5 min, % in weight). Due to its austenitic microstructure, this alloy shows very high toughness and is widely used in applications where a good corrosion resistance in aqueous environment or an excellent oxidation resistance in high temperature environment is required. Finally, a procedure for the structural integrity analysis of steam generator tubes with crack-like defects, based on a FAD diagram (Failure Assessment Diagram), is briefly described (author)

  17. Tube holding system

    International Nuclear Information System (INIS)

    Cunningham, R.C.

    1978-01-01

    A tube holding rig is described for the lateral support of tubes arranged in tight parcels in a heat exchanger. This tube holding rig includes not less than two tube supporting assemblies, with a space between them, located crosswise with respect to the tubes, each supporting assembly comprising a first set of parallel components in contact with the tubes, whilst a second set of components is also in contact with the tubes. These two sets of parts together define apertures through which the tubes pass [fr

  18. The effect of cracks on the limit load of pipe bends under in-plane bending

    International Nuclear Information System (INIS)

    Griffiths, J.E.

    1976-06-01

    The limit analysis of the in-plane bending of curved tubes had received attention previously, but the effect of defects in the tube has not been considered. A lower bound has been established which, with no defects present, is in agreement with previous theoretical work. The method of linear programming allows cracks to be introduced into analysis, and results have been obtained for various geometries of defect. The results show that the presence of cracks in the pipe bend can have a marked effect on the theoretical limit load: a part-through crack penetrating only half the wall thickness will reduce the limit moment by up to 10%. The worst possible case of a through-crack may reduce the limit load by 60%. (author)

  19. The effect of cracks on the limit load of pipe bends under in-plane bending

    International Nuclear Information System (INIS)

    Griffiths, J.E.

    1976-06-01

    The limit analysis of the in-plane bending of curved tubes had received attention previously, but the effect of defects in the tube has not been considered. A lower bound is established, which, with no defects present, is in agreement with previous theoretical work. The method of linear programming allows cracks to be introduced into the analysis. and results have been obtained for various geometries of defect. The results show that the presence of cracks in the pipe bend can have a marked effect on the theoretical limit load: a part-through crack penetrating only half the wall thickness will reduce the limit moment by up to 10%. The worst possible case of a through-crack may reduce the limit load by 60% (author)

  20. Life time estimation for irradiation assisted mechanical cracking of PWR RCCA rodlets

    Energy Technology Data Exchange (ETDEWEB)

    Matsuoka, Takanori; Yamaguchi, Youichirou [Nuclear Development Corp., Tokai, Ibaraki (Japan)

    1999-09-01

    Intergranular cracks of cladding tubes had been observed at the tips of the rodlets of PWR rod cluster control assemblies (RCCAs). Because RCCAs were important core components, an investigation was carried out to estimate their service lifetime. The reviews on their mechanism and the life time estimation are shown in this paper. The summaries are as follows. (1) The mechanism of the intergranular crack of the cladding tube was not IASCC but irradiation assisted mechanical cracking (IAMC) caused by an increase in hoop strain due to the swelling of the absorber and a decrease in elongation due to neutron irradiation. (2) The crack initiation limit of cylindrical shells made of low ductile material and subjected to internal pressure was determined in relation to the uniform strain of the material and was in accordance with that of the RCCA rodlets in an actual plant. (3) From the above investigation, the method of estimating the lifetime and countermeasures for its extension were obtained. (author)

  1. Branding on the Shop Floor

    Directory of Open Access Journals (Sweden)

    Szilvia Gyimóthy

    2010-09-01

    Full Text Available Service branding is a particular form of emotional management, where employees are regarded as adaptable media, who can be trained to convey corporate values while interacting with customers. This paper examines the identity work of butchers during the brand revitalisation campaign of Kvickly, a Danish supermarket chain. During the implementation of the “Best Butcher in Town”-project, Kvickly’s shop floor becomes an engineered servicescape where the norms of good salesmanship must be performed. By documenting the disloyal behaviour of butchers, we demonstrate that the affective commitment towards corporate brand values is closely related with self-enactment opportunities of occupational communities. Total service-orientation threatens butchers’ perception of autonomy and may therefore result in the emergence of resistant sub-cultures.

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

    International Nuclear Information System (INIS)

    Cizelj, L.; Riesch-Oppermann, H.

    1992-04-01

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

  3. Choked flow through cracks

    International Nuclear Information System (INIS)

    Feburie, V.; Giot, M.; Granger, S.; Seynhaeve, J.M.

    1992-06-01

    The leaks through steam-generator cracks are the subject of a research carried out in cooperation between EDF and UCL. A software called ECREVISSE to predict the mass flow rate has been developed and has been successfully validated. The purpose of the paper is to present the mathematical model used in ECREVISSE as well as some comparison between the results and the presently available data. The model takes into account the persistence of some metastable liquid in the crack and the special flow pattern which appears in such particular geometry. Although the model involves the use of several correlations (friction, heat transfer), no adjustment of parameters against the data has been needed, neither in the single-phase part of the flow, or in the two-phase part. (authors). 8 figs., 1 tab., 20 refs

  4. Flaw analysis in steam generator tube

    International Nuclear Information System (INIS)

    Hutin, J.P.; Billon, F.

    1985-08-01

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

  5. Fatigue Crack Topography.

    Science.gov (United States)

    1984-01-01

    alloys (2). [--I Fig. 6. Fatigue fracture in Nitrile- butadien rubber ( NBR ). Fig. 7. The characteristic features of fatigue fracture in press moulded...in plastics and even in rubber . It follows therefore, that fatigue fractures must also occur in the mineral layers of our earth or in the rock on...effective until the weakest point yields and forms a crack. To get a feeling for this process, you can imagine that the stressed article is made of rubber

  6. Distributed password cracking

    OpenAIRE

    Crumpacker, John R.

    2009-01-01

    Approved for public release, distribution unlimited Password cracking requires significant processing power, which in today's world is located at a workstation or home in the form of a desktop computer. Berkeley Open Infrastructure for Network Computing (BOINC) is the conduit to this significant source of processing power and John the Ripper is the key. BOINC is a distributed data processing system that incorporates client-server relationships to generically process data. The BOINC structu...

  7. Utopia Cracks and Polygons

    Science.gov (United States)

    2003-01-01

    MGS MOC Release No. MOC2-339, 23 April 2003This Mars Global Surveyor (MGS) Mars Orbiter Camera (MOC) image shows a pattern of polygonal cracks and aligned, elliptical pits in western Utopia Planitia. The picture covers an area about 3 km (about 1.9 mi) wide near 44.9oN, 274.7oW. Sunlight illuminates the scene from the left.

  8. Cracking hydrocarbons. [British patent

    Energy Technology Data Exchange (ETDEWEB)

    Heyl, G E

    1926-05-06

    The vapors from a still in which oils, coal tar, pitch, creosote, and c. or solid carbonaccous material such as coal or shale are cracked by being heated to 600/sup 0/ to 1000/sup 0/C. are passed through a fractionating column to remove high-boiling constituents which are passed into a second cracking still. The vapors from this still are treated to separate high-boiling fractions which are passed into a third still. The sills preferably contain removable troughs or liners, which are freed from carbon deposits either after removal from the still or by a scraping disc which is rotated in and moved along the trough. Oil to be cracked is forced by a pump through a preheater to a still. Vapours pass through a carbon separator and dephlegmator to a condenser. The reflux from the dephlegmator is forced by a pump to a still, the vapors from which pass through a carbon separator and a dephlegmator, the reflux from which is passed into a third still fitted with a separate carbon separator, dephlegmator and final condenser.

  9. Nondestructive evaluation of the QT on the SG tubes affected by chemical cleaning

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ki Seok Shin; Cheon, Keun Young; Kim, Wang Bae [Central Research Institute, Daejeon (Korea, Republic of); Min, Kyong Mahn [UMI, Daejeon (Korea, Republic of)

    2012-10-15

    The major mechanisms of flaws detected on the currently operating steam generator(SG) tubes are wear and stress corrosion cracking(SCC) defects. Wear defect has continuously occurred in the upper tube bundle imposed to the flow induced vibration at the interaction between tube and its support structure. Meanwhile, SCC has been formed by a variety of mixed mode, such as the corrosion susceptible material, residual stress and secondary side chemical environment of the SG tubes. Recently, corrosion related defects were detected in the domestic OPR 1000 model SG tubes especially in the egg crate tube support plate(TSP), as a form of axially oriented outer diameter stress corrosion cracking (ODSCC). Therefore, the need to take corrective measures against the corrosion defects is required and various studies have been conducted to clarify the main causes of the defects. In general, as a representing SG tube materials, Ni based alloy 600 tubes have been widely applied and also adversely shown weak properties on the corrosion cracking resistivity. According to the studies on the factors developing corrosion cracking, densely accumulated sludge pile on the secondary side of the SG tubes have been mainly attributed to the formation of the corrosion defects. Therefore, it is imperative to secure applicable and efficient sludge removal process. In this paper, the chemical cleaning processes to dissolve and remove the sludge, thus promote the integrity of the SG tubes were introduced and eddy current testing(ECT) results on the pre cracked SG tubes to determine the effectiveness of those processes were represented as well.

  10. 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.)

  11. 9 CFR 91.26 - Concrete flooring.

    Science.gov (United States)

    2010-01-01

    ... 9 Animals and Animal Products 1 2010-01-01 2010-01-01 false Concrete flooring. 91.26 Section 91.26... LIVESTOCK FOR EXPORTATION Inspection of Vessels and Accommodations § 91.26 Concrete flooring. (a) Pens aboard an ocean vessel shall have a 3 inch concrete pavement, proportioned and mixed to give 2000 psi...

  12. Laparoscopic Pelvic Floor Repair Using Polypropylene Mesh

    Directory of Open Access Journals (Sweden)

    Shih-Shien Weng

    2008-09-01

    Conclusion: Laparoscopic pelvic floor repair using a single piece of polypropylene mesh combined with uterosacral ligament suspension appears to be a feasible procedure for the treatment of advanced vaginal vault prolapse and enterocele. Fewer mesh erosions and postoperative pain syndromes were seen in patients who had no previous pelvic floor reconstructive surgery.

  13. Comfort analysis of lightweight floor system

    NARCIS (Netherlands)

    Zegers, S.F.A.J.G.; Herwijnen, van F.; Randall, B.

    2007-01-01

    During the past 60 years, floor systems used in housing and office-buildings in the Netherlands were mostly made of concrete or other similar materials, These floor systems, which can be characterized as heavy, normally posed little problems concerning vibrations. In recent years, in light of

  14. Physical distribution of oak strip flooring 1969

    Science.gov (United States)

    William C. Miller; William C. Miller

    1971-01-01

    As an aid to the marketing of oak strip flooring, a study was made of the distribution process for this product, from manufacture to consumer-where the flooring came from, where it went, how much was shipped, and who handled it.

  15. Biomechanics of the pelvic floor musculature

    NARCIS (Netherlands)

    Janda, S.

    2006-01-01

    The present thesis was motivated by two main goals. The first research goal of the thesis was to understand the complex biomechanical behaviour of the pelvic floor muscles. The second goal was to study the mechanism of the pelvic organ prolapse (genital prolapse). The pelvic floor in humans is a

  16. Building with electromagnetic shield structure for individual floors

    Energy Technology Data Exchange (ETDEWEB)

    Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

    1991-09-10

    This invention relates to a building having a floor-by-floor electromagnetic shield structure well-suited for application to an information network system in which an electromagnetically shielded space is divided by individual floors and electric waves are utilized within the building on a floor-by-floor basis. (author). 8 figs.

  17. Building with electromagnetic shield structure for individual floors

    International Nuclear Information System (INIS)

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1991-01-01

    This invention relates to a building having a floor-by-floor electromagnetic shield structure well-suited for application to an information network system in which an electromagnetically shielded space is divided by individual floors and electric waves are utilized within the building on a floor-by-floor basis. (author). 8 figs

  18. Investigation of Cracks Found in Helicopter Longerons

    Science.gov (United States)

    Newman, John A.; Baughman, James M.; Wallace, Terryl A.

    2009-01-01

    Four cracked longerons, containing a total of eight cracks, were provided for study. Cracked regions were cut from the longerons. Load was applied to open the cracks, enabling crack surface examination. Examination revealed that crack propagation was driven by fatigue loading in all eight cases. Fatigue crack initiation appears to have occurred on the top edge of the longerons near geometric changes that affect component bending stiffness. Additionally, metallurigical analysis has revealed a local depletion in alloying elements in the crack initiation regions that may be a contributing factor. Fatigue crack propagation appeared to be initially driven by opening-mode loading, but at a crack length of approximately 0.5 inches (12.7 mm), there is evidence of mixed-mode crack loading. For the longest cracks studied, shear-mode displacements destroyed crack-surface features of interest over significant portions of the crack surfaces.

  19. Modified Dugdale crack models - some easy crack relations

    DEFF Research Database (Denmark)

    Nielsen, Lauge Fuglsang

    1997-01-01

    the same strength as a plain Dugdale model. The critical energy release rates Gamma_CR, however, become different. Expressions (with easy computer algorithms) are presented in the paper which relate critical energy release rates and crack geometry to arbitrary cohesive stress distributions.For future...... lifetime analysis of viscoelastic materials strain energy release rates, crack geometries, and cohesive stress distributions are considered as related to sub-critical loads sigma stress-deformation tests......The Dugdale crack model is widely used in materials science to predict strength of defective (cracked) materials. A stable Dugdale crack in an elasto-plastic material is prevented from spreading by uniformly distributed cohesive stresses acting in narrow areas at the crack tips. These stresses...

  20. Imaging pelvic floor disorders. 2. rev. ed.

    International Nuclear Information System (INIS)

    Stoker, Jaap; Taylor, Stuart A.; DeLancey, John O.L.

    2008-01-01

    This volume builds on the success of the first edition of imaging pelvic floor disorders and is aimed at those practitioners with an interest in the imaging, diagnosis and treatment of pelvic floor dysfunction. Concise textual information from acknowledged experts is complemented by high-quality diagrams and images to provide a thorough update of this rapidly evolving field. Introductory chapters fully elucidate the anatomical basis underlying disorders of the pelvic floor. State of the art imaging techniques and their application in pelvic floor dysfunction are then discussed in detail. Additions since the first edition include consideration of the effect of aging and new chapters on perineal ultrasound, functional MRI and MRI of the levator muscles. The closing sections of the book describe the modern clinical management of pelvic floor dysfunction, including prolapse, urinary and faecal incontinence and constipation, with specific emphasis on the integration of diagnostic and treatment algorithms. (orig.)

  1. Imaging pelvic floor disorders. 2. rev. ed.

    Energy Technology Data Exchange (ETDEWEB)

    Stoker, Jaap [Amsterdam Univ. (Netherlands). Dept. of Radiology; Taylor, Stuart A. [University College Hospital, London (United Kingdom). Dept. of Specialist X-Ray; DeLancey, John O.L. (eds.) [Michigan Univ., Ann Arbor, MI (United States). L4000 Women' s Hospital

    2008-07-01

    This volume builds on the success of the first edition of imaging pelvic floor disorders and is aimed at those practitioners with an interest in the imaging, diagnosis and treatment of pelvic floor dysfunction. Concise textual information from acknowledged experts is complemented by high-quality diagrams and images to provide a thorough update of this rapidly evolving field. Introductory chapters fully elucidate the anatomical basis underlying disorders of the pelvic floor. State of the art imaging techniques and their application in pelvic floor dysfunction are then discussed in detail. Additions since the first edition include consideration of the effect of aging and new chapters on perineal ultrasound, functional MRI and MRI of the levator muscles. The closing sections of the book describe the modern clinical management of pelvic floor dysfunction, including prolapse, urinary and faecal incontinence and constipation, with specific emphasis on the integration of diagnostic and treatment algorithms. (orig.)

  2. Cryogenic testing of fluoropolymer-coated stainless steel tubing

    International Nuclear Information System (INIS)

    Dooley, J.B.

    1989-11-01

    Stainless steel tubing coated internally with two different types of fluorinated polymers were subjected to microscopic examination after a welding operation had been performed on the tubing. The welded assemblies were photographed and subjected to repeated cycles between liquid helium and room temperature. The green tetrafluoroethylene (TFE) coating peeled back in the area subjected to welding heat and displayed cracking all over its surface without regard to proximity to the weld area. The dark fluorinated ethylene propylene (FEP) coating showed a tendency to char or burn away progressively in the weld area. The dark (FEP) coating did not crack as extensively as the green TFE coating, but did show a few areas of ''crazing'' or cracking of the topmost surface after cryogenic exposure. 12 figs

  3. Numerical Analysis of Prefabricated Steel-Concrete Composite Floor in Typical Lipsk Building

    Directory of Open Access Journals (Sweden)

    Lacki Piotr

    2017-12-01

    Full Text Available The aim of the work was to perform numerical analysis of a steel-concrete composite floor located in a LIPSK type building. A numerical model of the analytically designed floor was performed. The floor was in a six-storey, retail and service building. The thickness of a prefabricated slab was 100 mm. The two-row, crisscrossed reinforcement of the slab was made from φ16 mm rods with a spacing of 150 x 200 mm. The span of the beams made of steel IPE 160 profiles was 6.00 m and they were spaced every 1.20 m. The steelconcrete composite was obtained using 80×16 Nelson fasteners. The numerical analysis was carried out using the ADINA System based on the Finite Element Method. The stresses and strains in the steel and concrete elements, the distribution of the forces in the reinforcement bars and cracking in concrete were evaluated. The FEM model was made from 3D-solid finite elements (IPE profile and concrete slab and truss elements (reinforcement bars. The adopted steel material model takes into consideration the plastic state, while the adopted concrete material model takes into account material cracks.

  4. 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

  5. Sleeving repair of heat exchanger tubes

    International Nuclear Information System (INIS)

    Street, Michael D.; Schafer, Bruce W.

    2000-01-01

    Defective heat exchanger tubes can be repaired using techniques that do not involve the cost and schedule penalties of component replacement. FTI's years of experience repairing steam generator tubes have been successfully applied to heat exchangers. Framatome Technologies heat exchanger sleeves can bridge defective areas of the heat exchanger tubes, sleeves have been designed to repair typical heat exchanger tube defects caused by excessive tube vibration, stress corrosion cracking, pitting or erosion. By installing a sleeve, the majority of the tube's heat transfer and flow capacity is maintained and the need to replace the heat exchanger can be delayed or eliminated. Both performance and reliability are improved. FTI typically installs heat exchanger tube sleeves using either a roll expansion or hydraulic expansion process. While roll expansion of a sleeve can be accomplished very quickly, hydraulic expansion allows sleeves to be installed deep within a tube where a roll expander cannot reach. Benefits of FTI's heat exchanger tube sleeving techniques include: - Sleeves can be positioned any where along the tube length, and for precise positioning of the sleeve eddy current techniques can be employed. - Varying sleeve lengths can be used. - Both the roll and hydraulic expansion processes are rapid and both produce joints that do not require stress relief. - Because of low leak rates and speed of installations, sleeves can be used to preventatively repair likely-to-fail tubes. - Sleeves can be used for tube stiffening and to limit leakage through tube defects. - Because of installation speed, there is minimal impact on outage schedules and budgets. FTI's recently installed heat exchanger sleeving at the Kori-3 Nuclear Power Station in conjunction with Korea Plant Service and Engineering Co., Ltd. The sleeves were installed in the 3A and 3B component cooling water heat exchangers. A total of 859 tubesheet and 68 freespan sleeves were installed in the 3A heat

  6. Crack retardation by load reduction during fatigue crack propagation

    International Nuclear Information System (INIS)

    Kim, Hyun Soo; Nam, Ki Woo; Ahn, Seok Hwan; Do, Jae Yoon

    2003-01-01

    Fracture life and crack retardation behavior were examined experimentally using CT specimens of aluminum alloy 5083. Crack retardation life and fracture life were a wide difference between 0.8 and 0.6 in proportion to ratio of load reduction. The wheeler model retardation parameter was used successfully to predict crack growth behavior. By using a crack propagation rule, prediction of fracture life can be evaluated quantitatively. A statistical approach based on Weibull distribution was applied to the test data to evaluate the dispersion in the retardation life and fracture life by the change of load reduction

  7. Ductile crack growth simulation from near crack tip dissipated energy

    International Nuclear Information System (INIS)

    Marie, S.; Chapuliot, S.

    2000-01-01

    A method to calculate ductile tearing in both small scale fracture mechanics specimens and cracked components is presented. This method is based on an estimation of the dissipated energy calculated near the crack tip. Firstly, the method is presented. It is shown that a characteristic parameter G fr can be obtained, relevant to the dissipated energy in the fracture process. The application of the method to the calculation of side grooved crack tip (CT) specimens of different sizes is examined. The value of G fr is identified by comparing the calculated and experimental load line displacement versus crack extension curve for the smallest CT specimen. With this identified value, it is possible to calculate the global behaviour of the largest specimen. The method is then applied to the calculation of a pipe containing a through-wall thickness crack subjected to a bending moment. This pipe is made of the same material as the CT specimens. It is shown that it is possible to simulate the global behaviour of the structure including the prediction of up to 90-mm crack extension. Local terms such as the equivalent stress or the crack tip opening angle are found to be constant during the crack extension process. This supports the view that G fr controls the fields in the vicinity near the crack tip. (orig.)

  8. Effect of crevice environment PH on corrosion damage of horizontal steam generator tubes

    International Nuclear Information System (INIS)

    Brozova, A.; Burda, J.; Splichal, K.

    2002-01-01

    In support of a project on lifetime calculation experiments were carried out to evaluate the resistance to environmentally assisted cracking (EAC) of steam generator tubes during operation. Estimations of the incubation period for crack initiation and the threshold K value, K Iscc , and the crack growth rate were made to predict evolution of damage in tube walls. The paper summarizes results of experiments of C ring specimen for the initiation testing and results of SENT (single edge notch tensile) specimen for the crack growth rate (CGR) testing. The specimens were exposed to concentrated environments at elevated temperatures simulating crevice environments in secondary side crevices in horizontal steam generators. The results show that the material of SG tubes is sensitive to transgranular environmentally assisted cracking in the three basic concentrated environments used, alkaline, neutral and acid. The most corrosive medium was the acid environment. The crack initiated practically immediately after acid environment exposure. The initiation process takes a long time in neutral and alkaline environments. The K Iscc values for environmentally assisted crack growth rate in alkaline and neutral concentrated environment were essentially the same. The crack growth rate was slightly higher for the neutral environment than for the alkaline one. Fracture patterns for the both environments were similar. (author)

  9. Mechanical strength evaluation of the glass base material in the JRR-3 neutron guide tube

    Energy Technology Data Exchange (ETDEWEB)

    Kobayashi, Tetsuya [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-02-01

    The lifetime of the thermal neutron guide tube installed JRR-3 was investigated after 6 years from their first installation. And it was confirmed that a crack had been piercing into the glass base material of the side plate of the neutron guide tube. The cause of the crack was estimated as a static fatigue of the guide tube where an inside of the tube had been evacuated and stressed as well as an embrittlement of the glass base material by gamma ray irradiation. In this report, we evaluate the mechanical strength of the glass base material and estimate the time when the base material gets fatigue fracture. Furthermore, we evaluate a lifetime of the neutron guide tube and confirm the validity of update timing in 2000 and 2001 when the thermal neutron guide tubes T1 and T2 were exchanged into those using the super mirror. (author)

  10. Dynamic Characteristics of Steam Generator Tubes with Defect due to Wear

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangjin; Rhee, Huinam [Sunchon National Univ., Sunchon (Korea, Republic of); Yoon, Doo Byung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    These defects may affect the dynamic characteristics of tubes, and therefore, the vibrational behavior of the tube due to flow-induced loads can be varied. Change in the vibrational response of a tube may result in different wear characteristics from the design condition, which must be checked for both safety and economic point of view. This paper deals with the study on the effect of wears or cracks on the dynamic characteristics of steam generator tubes using finite element analysis. In this paper the effect of defects on the surface due to wear on the variation of dynamic characteristics of steam generator tubes was studied using the finite element analysis. The changes of natural frequencies and mode shapes can directly affect the flow-induced vibration response characteristics, therefore, they must be evaluated appropriately. The results in this study can be a good basis to estimate the FIV characteristics of the steam generator tubes having defects such as wear or crack.

  11. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1983 and 1984

    International Nuclear Information System (INIS)

    Tatone, O.S.; Meindl, P.; Taylor, G.F.

    1986-06-01

    A review of the performance of steam generator tubes in water-cooled nuclear power reactors showed that tubes were plugged at 47 (35.6%) of the reactors in 1983 and at 63 (42.6%) of the reactors during 1984. In 1983 and 1984 3291 and 3335 tubes, respectively, were removed from service, about the same as in 1982. The leading causes assigned to tube failure were stress corrosion cracking from the primary side and stress corrosion cracking or intergranular attack from the secondary side. In addition 5668 tubes were repaired for further service by installation of internal sleeves. Most of these were believed to have deteriorated by one of the above mechanisms or by pitting. There is a continuing trend towards high-integrity condenser tube materials at sites cooled by brackish or sea water. 31 refs

  12. Creep collapse of thick-walled heat transfer tube subjected to external pressure at high temperature

    International Nuclear Information System (INIS)

    Ioka, Ikuo; Kaji, Yoshiyuki; Terunuma, Isao; Nekoya, Shin-ichi; Miyamoto, Yoshiaki

    1994-09-01

    A series of creep collapse tests of thick-walled heat transfer tube were examined experimentally and analytically to confirm an analytical method for creep deformation behavior of a heat transfer tube of an intermediate heat exchanger (IHX) at a depressurization accident of secondary cooling system of HTTR (High Temperature Engineering Test Reactor). The tests were carried out using thick-walled heat transfer tubes made of Hastelloy XR at 950degC in helium gas environment. The predictions of creep collapse time obtained by a general purpose FEM-code ABAQUS were in good agreement with the experimental results. A lot of cracks were observed on the outer surface of the test tubes after the creep collapse. However, the cracks did not pass through the tube wall and, therefore, the leak tightness was maintained regardless of a collapse deformation for all tubes tested. (author)

  13. Apparatus for treating the walls and floor of the pelvic cavity with radiation

    International Nuclear Information System (INIS)

    Clayton, R.S.

    1975-01-01

    An apparatus for reaing carcinoma of the walls and floor of the pelvic cavity is described. An elongated tube has an inner end adapted to be placed in the pelvic cavity and an outer end adapted to extend through to the outside of the body. Radioactive material is placed at the inner end. An inner balloon above the radioactive material is inflated to hold a body of liquid shielding material such as mercury. A lower balloon portion beneath the inner balloon spaces areas to be treated such as the walls and floor of the pelvic cavity from the radioactive material. An upper balloon portion above the inner balloon keeps the intestines out of the pelvic cavity and away from the radioactive material. The apparatus is inserted into the pelvic cavity through an abdominal incision. When treating a woman for carcinoma in the walls and floor of the pelvic cavity the tube is moved through the vaginal passage from the inside outwardly. When treating a woman with a closed vaginal passage, as may result from surgery, or when treating a man, such as for carcinoma of the bladder, the tube will pass out of the body through a lower abdominal incision. Following treatment, all balloons are deflated so that the apparatus can be withdrawn through the vaginal passage or the lower abdominal incision, as the case may be. (auth)

  14. Bender/Coiler for Tubing

    Science.gov (United States)

    Stoltzfus, J. M.

    1983-01-01

    Easy-to-use tool makes coils of tubing. Tubing to be bend clamped with stop post. Die positioned snugly against tubing. Operator turns handle to slide die along tubing, pushing tubing into spiral groove on mandrel.

  15. 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

  16. Corrosion fatigue cracking behavior of Inconel 690 (TT) in secondary water of pressurized water reactors

    International Nuclear Information System (INIS)

    Xiao Jun; Chen Luyao; Qiu Shaoyu; Chen Yong; Lin Zhenxia; Fu Zhenghong

    2015-01-01

    Inconel 690 (TT) is one of the key materials for tubes of steam generators for pressurized water reactors, where it is susceptible to corrosion fatigue cracking. In this paper, the corrosion fatigue cracking behavior of Inconel 690 (TT) was investigated under small scale yielding conditions, in the simulated secondary water of pressurized water reactor. It was observed that the fatigue crack growth rate was accelerated by a maximum factor up to 3 in the simulated secondary water, comparing to that in room temperature air. In addition, it was found that the accelerating effect was influenced by out-of-plane cracking of corrosion fatigue cracks and also correlated with stress intensity factor range, maximum stress intensity factor and stress ratio. (authors)

  17. 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)

  18. Material reliability of Ni alloy electrodeposition for steam generator tube repair

    International Nuclear Information System (INIS)

    Kim, Dong Jin; Kim, Myong Jin; Kim, Joung Soo; Kim, Hong Pyo

    2007-01-01

    Due to the occasional occurrences of Stress Corrosion Cracking (SCC) in steam generator tubing (Alloy 600), degraded tubes are removed from service by plugging or are repaired for re-use. Since electrodeposition inside a tube dose not entail parent tube deformation, residual stress in the tube can be minimized. In this work, tube restoration via electrodeposition inside a steam generator tubing was performed after developing the following: an anode probe to be installed inside a tube, a degreasing condition to remove dirt and grease, an activation condition for surface oxide elimination, a tightly adhered strike layer forming condition between the electroforming layer and the Alloy 600 tube, and the condition for an electroforming layer. The reliability of the electrodeposited material, with a variation of material properties, was evaluated as a function of the electrodeposit position in the vertical direction of a tube using the developed anode. It has been noted that the variation of the material properties along the electrodeposit length was acceptable in a process margin. To improve the reliability of a material property, the causes of the variation occurrence were presumed, and an attempt to minimize the variation has been made. A Ni alloy electrodeposition process is suggested as a Primary Water Stress Corrosion Cracking (PWSCC) mitigation method for various components, including steam generator tubes. The Ni alloy electrodeposit formed inside a tube by using the installed assembly shows proper material properties as well as an excellent SCC resistance

  19. Cracking of anisotropic cylindrical polytropes

    Energy Technology Data Exchange (ETDEWEB)

    Mardan, S.A. [University of the Management and Technology, Department of Mathematics, Lahore (Pakistan); Azam, M. [University of Education, Division of Science and Technology, Lahore (Pakistan)

    2017-06-15

    We study the appearance of cracking in charged anisotropic cylindrical polytropes with generalized polytropic equation. We investigate the existence of cracking in two different kinds of polytropes existing in the literature through two different assumptions: (a) local density perturbation with conformally flat condition, and (b) perturbing polytropic index, charge and anisotropy parameters. We conclude that cracking appears in both kinds of polytropes for a specific range of density and model parameters. (orig.)

  20. Validating eddy current array probes for inspecting steam generator tubes

    International Nuclear Information System (INIS)

    Sullivan, S.P.; Cecco, V.S.; Obrutsky, L.S.

    1997-01-01

    A CANDU nuclear reactor was shut down for over one year because steam generator (SG) tubes had failed with outer diameter stress-corrosion cracking (ODSCC) in the U-bend section. Novel, single-pass eddy current transmit-receive probes, denoted as C3, were successful in detecting all significant cracks so that the cracked tubes could be plugged and the unit restarted. Significant numbers of tubes with SCC were removed from a SG in order to validate the results of the new probe. Results from metallurgical examinations were used to obtain probability-of-detection (POD) and sizing accuracy plots to quantify the performance of this new inspection technique. Though effective, the above approach of relying on tubes removed from a reactor is expensive, in terms of both economic and radiation-exposure costs. This led to a search for more affordable methods to validate inspection techniques and procedures. Methods are presented for calculating POD curves based on signal-to-noise studies using field data. Results of eddy current scans of tubes with laboratory-induced ODSCC are presented with associated POD curves. These studies appear promising in predicting realistic POD curves for new inspection technologies. They are being used to qualify an improved eddy current array probe in preparation for field use. (author)

  1. 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

  2. On the heat exchange tube failures in steam generators at NPPs with WWER reactors

    International Nuclear Information System (INIS)

    Titov, V.F.; Banyuk, G.F.; Brykov, S.I.

    1992-01-01

    Data on dynamics of failed heat exchanging tube closing in steam generators of NPPs with WWER type reactors for the whole period of their operation are presented. It is shown that the main cause of the tube failures consists in their corrosion cracking under stresses. The effect of chlorine ions on tubes is intensified by the presence of porous sediments on heat exchaning surfaces in quantities exceeding 150 g/m 2

  3. Cryptography cracking codes

    CERN Document Server

    2014-01-01

    While cracking a code might seem like something few of us would encounter in our daily lives, it is actually far more prevalent than we may realize. Anyone who has had personal information taken because of a hacked email account can understand the need for cryptography and the importance of encryption-essentially the need to code information to keep it safe. This detailed volume examines the logic and science behind various ciphers, their real world uses, how codes can be broken, and the use of technology in this oft-overlooked field.

  4. Stress corrosion cracking

    International Nuclear Information System (INIS)

    Dietzel, W.; Turnbull, A.

    2007-01-01

    Comprehensive Structural Integrity is a reference work which covers all activities involved in the assurance of structural integrity. It provides engineers and scientists with an unparalleled depth of knowledge in the disciplines involved. The new online Volume 11 is dedicated to the mechanical characteristics of materials. This paper contains the chapter 11.03 and is structured as follows: General aspects of SCC testing; Non-precracked specimens; Precracked specimens - the fracture mechanics approach to SCC; Crack growth measurement; Limitations of the LEFM approach to SCC; The use of SCC data; Guide to selection of mechanical scc test method

  5. Analysis of WWER 1000 SG cold collector cracking

    International Nuclear Information System (INIS)

    Matocha, K.; Wozniak, J.

    2000-01-01

    Following the recommendations of the 1993 consultants' meeting on 'Steam Generator Collector Integrity of WWER 1000 Reactors', an extensive experimental program was started with the aim of finding the dominant damage mechanism responsible for cold collector cracking in steam generators, and of determining whether proper operating conditions can make the operation of VITKOVICE-produced steam generators safe throughout their lifetime. The experiments consisted of: a study of the effect of strain and thermal ageing and dissolved oxygen content on subcritical crack growth in 10GN2MFA steel; a study of the effect of high temperature water and tube expansion technology on the fracture behaviour of ligaments between holes for heat exchange tubes; a study of the effect of drilling, tube expansion technology and heat treatment on residual stresses on the surface of holes for heat exchange tubes. Details of the experimental techniques used are given as well as a discussion of the results obtained and presented in tables and graphs. (A.K.)

  6. Simulations of floor cooling system capacity

    International Nuclear Information System (INIS)

    Odyjas, Andrzej; Górka, Andrzej

    2013-01-01

    Floor cooling system capacity depends on its physical and operative parameters. Using numerical simulations, it appears that cooling capacity of the system largely depends on the type of cooling loads occurring in the room. In the case of convective cooling loads capacity of the system is small. However, when radiation flux falls directly on the floor the system significantly increases productivity. The article describes the results of numerical simulations which allow to determine system capacity in steady thermal conditions, depending on the type of physical parameters of the system and the type of cooling load occurring in the room. Moreover, the paper sets out the limits of system capacity while maintaining a minimum temperature of the floor surface equal to 20 °C. The results are helpful for designing system capacity in different type of cooling loads and show maximum system capacity in acceptable thermal comfort condition. -- Highlights: ► We have developed numerical model for simulation of floor cooling system. ► We have described floor system capacity depending on its physical parameters. ► We have described floor system capacity depending on type of cooling loads. ► The most important in the obtained cooling capacities is the type of cooling loads. ► The paper sets out the possible maximum cooling floor system capacity

  7. Pelvic floor and sexual male dysfunction

    Directory of Open Access Journals (Sweden)

    Antonella Pischedda

    2013-04-01

    Full Text Available The pelvic floor is a complex multifunctional structure that corresponds to the genito- urinary-anal area and consists of muscle and connective tissue. It supports the urinary, fecal, sexual and reproductive functions and pelvic statics. The symptoms caused by pelvic floor dysfunction often affect the quality of life of those who are afflicted, worsening significantly more aspects of daily life. In fact, in addition to providing support to the pelvic organs, the deep floor muscles support urinary continence and intestinal emptying whereas the superficial floor muscles are involved in the mechanism of erection and ejaculation. So, conditions of muscle hypotonia or hypertonicity may affect the efficiency of the pelvic floor, altering both the functionality of the deep and superficial floor muscles. In this evolution of knowledge it is possible imagine how the rehabilitation techniques of pelvic floor muscles, if altered and able to support a voiding or evacuative or sexual dysfunction, may have a role in improving the health and the quality of life.

  8. The development of crack measurement system using the direct current potential drop method for use in the hot cell

    International Nuclear Information System (INIS)

    Kim, Do-Sik; Ahn, Sang-Bok; Lee, Key-Soon; Kim, Yong-Suk; Kwon, Sang-Chul

    1999-01-01

    The crack length measurement system using the direct current potential drop (DCPD) method was developed for the detection of crack growth initiation and subsequent crack growth. The experimental precautions and data processing procedure required for its application were also described find discussed. The system presented herein was specially built for use in fracture toughness testing of unirradiated or irradiated pressure tube materials from nuclear reactor. The application of this system for fracture toughness determination was illustrated from the test of curved compact tension specimens removed from CANDU reactor pressure tubes. The crack extension was monitored using the DCPD method. It is found that the changes of the potential drop and the changes of the crack length have a linear relationship. The final crack front was marked by heat-tinting after the test and the specimen broken open for determination of the initial and final physical crack length. The physical crack lengths, obtained by the 9-point average method described in ASTM E1737-96 on heat-tinted fracture surface, were used to calibrate the DCPD method for each test on an individual basis by matching the change in voltage to the crack extension. It is found that this system can be recommended for determination of the J-integral resistance (J-R) curve of unirradiated or irradiation materials in the hot cell, especially when testing at elevated temperature and in the environment chamber or furnace. (author)

  9. Stress corrosion cracking of Zircaloys. Final report

    International Nuclear Information System (INIS)

    Cubicciotti, D.; Jones, R.L.; Syrett, B.C.

    1980-03-01

    The overall aim has been to develop an improved understanding of the stress corrosion cracking (SCC) mechanism considered to be responsible for pellet-cladding interaction (PCI) failures of nuclear fuel rods. The objective of the present phase of the project was to investigate the potential for improving the resistance of Zircaloy to iodine-induced SCC by modifying the manufacturing techniques used in the commercial production of fuel cladding. Several aspects of iodine SCC behavior of potential relevance to cladding performance were experimentally investigated. It was found that the SCC susceptibility of Zircaloy tubing is sensitive to crystallographic texture, surface condition, and residual stress distribution and that current specifications for Zircaloy tubing provide no assurance of an optimum resistance to SCC. Additional evidence was found that iodine-induced cracks initiate at local chemical inhomogeneities in the Zircaloy surface, but laser melting to produce a homogenized surface layer did not improve the SCC resistance. Several results were obtained that should be considered in models of PCI failure. The ratio of axial to hoop stress and the temperature were both shown to affect the SCC resistance whereas the difference in composition between Zircaloy-2 and Zircaloy-4 had no detectable effect. Damage accumulation during iodine SCC was found to be nonlinear: generally, a given life fraction at low stress was more damaging than the same life fraction at higher stress. Studies of the thermochemistry of the zirconium-iodine system (performed under US Department of Energy sponsorship) revealed many errors in the literature and provided important new insights into the mechanism of iodine SCC of Zircaloys

  10. 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

  11. Catalytic cracking of lignites

    Energy Technology Data Exchange (ETDEWEB)

    Seitz, M.; Nowak, S.; Naegler, T.; Zimmermann, J. [Hochschule Merseburg (Germany); Welscher, J.; Schwieger, W. [Erlangen-Nuernberg Univ. (Germany); Hahn, T. [Halle-Wittenberg Univ., Halle (Germany)

    2013-11-01

    A most important factor for the chemical industry is the availability of cheap raw materials. As the oil price of crude oil is rising alternative feedstocks like coal are coming into focus. This work, the catalytic cracking of lignite is part of the alliance ibi (innovative Braunkohlenintegration) to use lignite as a raw material to produce chemicals. With this new one step process without an input of external hydrogen, mostly propylene, butenes and aromatics and char are formed. The product yield depends on manifold process parameters. The use of acid catalysts (zeolites like MFI) shows the highest amount of the desired products. Hydrogen rich lignites with a molar H/C ratio of > 1 are to be favoured. Due to primary cracking and secondary reactions the ratio between catalyst and lignite, temperature and residence time are the most important parameter to control the product distribution. Experiments at 500 C in a discontinuous rotary kiln reactor show yields up to 32 wt-% of hydrocarbons per lignite (maf - moisture and ash free) and 43 wt-% char, which can be gasified. Particularly, the yields of propylene and butenes as main products can be enhanced four times to about 8 wt-% by the use of catalysts while the tar yield decreases. In order to develop this innovative process catalyst systems fixed on beads were developed for an easy separation and regeneration of the used catalyst from the formed char. (orig.)

  12. Environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Soppet, W.K.; Shack, W.J.

    1992-03-01

    This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking in high water reactors during the six months from April 1991 through September 1991. Topics that have been investigated during this period include (1) fatigue and stress corrosion cracking (SCC) of low-alloy steel used in piping and in steam generator and reactor pressure vessels; (2) role of chromate and sulfate in simulated boiling water reactor (BWR) water on SCC of sensitized Type 304 SS; and (3) radiation-induced segregation (RIS) and irradiation-assisted SCC of Type 304 SS after accumulation of relatively high fluence. Fatigue data were obtained on medium-S-content A533-Gr B and A106-Gr B steels in high-purity (HP) deoxygenated water, in simulated pressurized water reactor (PWR) water, and in air. Crack-growth-rates (CGRs) of composite specimens of A533-Gr B/Inconel-182/Inconel-600 (plated with nickel) and homogeneous specimens of A533-Gr B were determined under small- amplitude cyclic loading in HP water with ∼ 300 ppb dissolved oxygen. CGR tests on sensitized Type 304 SS indicate that low chromate concentrations in BWR water (25--35 ppb) may actually have a beneficial effect on SCC if the sulfate concentration is below a critical level. Microchemical and microstructural changes in HP and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy, and slow-strain,rate- tensile tests were conducts on tubular specimens in air and in simulated BWR water at 289 degrees C

  13. Crack closure and growth behavior of short fatigue cracks under random loading (part I : details of crack closure behavior)

    International Nuclear Information System (INIS)

    Lee, Shin Young; Song, Ji Ho

    2000-01-01

    Crack closure and growth behavior of physically short fatigue cracks under random loading are investigated by performing narrow-and wide-band random loading tests for various stress ratios. Artificially prepared two-dimensional, short through-thickness cracks are used. The closure behavior of short cracks under random loading is discussed, comparing with that of short cracks under constant-amplitude loading and also that of long cracks under random loading. Irrespective of random loading spectrum or block length, the crack opening load of short cracks is much lower under random loading than under constant-amplitude loading corresponding to the largest load cycle in a random load history, contrary to the behavior of long cracks that the crack opening load under random loading is nearly the same as or slightly higher than constant-amplitude results. This result indicates that the largest load cycle in a random load history has an effect to enhance crack opening of short cracks

  14. Crack Tip Parameters for Growing Cracks in Linear Viscoelastic Materials

    DEFF Research Database (Denmark)

    Brincker, Rune

    In this paper the problem of describing the asymptotic fields around a slowly growing crack in a linearly viscoelastic material is considered. It is shown that for plane mixed mode problems the asymptotic fields must be described by 6 parameters: 2 stress intensity factors and 4 deformation...... intensity factors. In the special case of a constant Poisson ratio only 2 deformation intensity factors are needed. Closed form solutions are given both for a slowly growing crack and for a crack that is suddenly arrested at a point at the crack extension path. Two examples are studied; a stress boundary...... value problem, and a displacement boundary value problem. The results show that the stress intensity factors and the displacement intensity factors do not depend explicitly upon the velocity of the crack tip....

  15. Leak-before-break assessment of RBMK-1500 fuel channel in case of delayed hydride cracking

    International Nuclear Information System (INIS)

    Klimasauskas, A.; Grybenas, A.; Makarevicius, V.; Nedzinskas, L.; Levinskas, R.; Kiselev, V.

    2003-01-01

    One of the factors determining remaining lifetime of Zr-2.5% Nb fuel channel (FC) is the amount of hydrogen dissolved during corrosion process. When the concentration of hydrogen exceeds the terminal solid solubility limit zirconium hydrides are precipitated. As a result form necessary conditions for delayed hydride cracking (DHC). Data from the RBMK-1500 fuel channel tubes (removed from service) shows that hydrogen in some cases distributes unevenly and hydrogen concentration can differ several times between individual FC tubes or separate zones of the same tube and possibly, can reach dangerous levels in the future. Consequently, lacking statistical research data, it is difficult to forecast increase of hydrogen concentration and formation of DHC. So it is important to verify if under the most unfavorable situation leak before break condition will be satisfied in the case of DHC. To estimate possible DHC rates in RBMK 1500 FC pressure tubes experiments were done in the following order: hydriding of the Zr-2.5Nb pressure tube material to the required hydrogen concentration; hydrogen analysis; machining of specimens, fatigue crack formation in the axial direction, DHC testing; average crack length measurement and DHC velocity calculation. During the tests in average DHC values were determined at 283, 250 and 144 degC (with hydrogen concentrations correspondingly 76, 54 and 27 ppm). The fracture resistance dependence from hydrogen concentration was measured at 20 degC. To calculate leak through the postulated flaw, statistical distribution of DHC surface irregularity was determined. Leak before break analysis was carried out according to requirements of RBMK 1500 regulatory documents. J integral and crack opening were calculated using finite element method. Loading of the FC was determined using RELAP5 code. Critical crack length was calculated using R6 and J-integral methods. Coolant flow rate through the postulated crack was estimated using SQUIRT software

  16. 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

  17. Magnetic Resonance Imaging (MRI): Dynamic Pelvic Floor

    Science.gov (United States)

    ... to a CD or uploaded to a digital cloud server. Dynamic pelvic floor MRI provides detailed pictures ... with you. top of page What are the benefits vs. risks? Benefits MRI is a noninvasive imaging ...

  18. Pelvic floor electrophysiology patterns associated with faecal ...

    African Journals Online (AJOL)

    Hussein Al-Moghazy Sultan

    2012-12-28

    Dec 28, 2012 ... pelvic floor electrophysiological abnormalities associated with. FI were illustrated in ... detection of a localized anal sphincter defect clinically and ..... Woods R, Voyvodic F, Schloithe A, Sage M, Wattchow D. Anal sphincter ...

  19. Decontamination of polyvinylchloride- and rubber type flooring

    International Nuclear Information System (INIS)

    Kunze, S.

    1975-01-01

    These types, fabricated by mixing of the basic components, showed no relation between content of fillers and decontamination results. Decontamination results are partly poorer, if the flooring contains a high concentration of the filler, especially if the latter consists mainly of hydrophilic materials. The coloring of the floorings seems to have no influence on the decontamination but floorings with clearly separated patterns can not be recommended for nuclear facilities. Fabricated by chemical reactions between polymeres, vulcanization materials and fillers, the decontamination results depend definitely from the proper choice of the filler. Flooring types, containing lampblack, graphite, kaoline, barium sulfate and titanium oxide are easy to decontamine. Again, increasing contents of hydrophilic filler cause a fall off in the decontamination results. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1994-12-31

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

  1. Anonymous electronic trading versus floor trading

    OpenAIRE

    Franke, Günter; Hess, Dieter

    1995-01-01

    This paper compares the attractiveness of floor trading and anonymous electronic trading systems. It is argued that in times of low information intensity the insight into the order book of the electronic trading system provides more valuable information than floor trading, but in times of high information intensity the reverse is true. Thus, the electronic system's market share in trading activity should decline in times of high information intensity. This hypothesis is tested by data on BUND...

  2. Cracking in Drying Colloidal Films

    Science.gov (United States)

    Singh, Karnail B.; Tirumkudulu, Mahesh S.

    2007-05-01

    It has long been known that thick films of colloidal dispersions such as wet clays, paints, and coatings crack under drying. Although capillary stresses generated during drying have been recently identified as the cause for cracking, the existence of a maximum crack-free film thickness that depends on particle size, rigidity, and packing has not been understood. Here, we identify two distinct regimes for crack-free films based on the magnitude of compressive strain at the maximum attainable capillary pressure and show remarkable agreement of measurements with our theory. We anticipate our results to not only form the basis for design of coating formulations for the paints, coatings, and ceramics industry but also assist in the production of crack-free photonic band gap crystals.

  3. Crack tip stress and strain

    International Nuclear Information System (INIS)

    Francois, D.

    1975-01-01

    The study of potential energy variations in a loaded elastic solid containing a crack leads to determination of the crack driving force G. Generalization of this concept to cases other than linear elasticity leads to definition of the integral J. In a linear solid, the crack tip stress field is characterized by a single parameter: the stress-intensity factor K. When the crack tip plastic zone size is confined to the elastic singularity J=G, it is possible to establish relationship between these parameters and plastic strain (and in particular the crack tip opening displacement delta). The stress increases because of the triaxiality effect. This overload rises with increasing strain hardening. When the plastic zone size expands, using certain hypotheses, delta can be calculated. The plastic strain intensity is exclusively dependent on parameter J [fr

  4. Prediction of Crack Growth Aqueous Environments.

    Science.gov (United States)

    1983-06-01

    ORGANIZATION NAME AND ADDRESS 10. PROGRAM ELEMENT. PROJECT. TASK AREA & WORK UNIT NUMBERS SRI International 333 Ravenswood Avenue Menlo Park, CA 94025 II...34no crack" has at least a vestigial rupture, associated with cyclic loading of the oxide film at the crack tip. The curve labeled "crack" was obtained...be an effect of crack opening. For the data set labeled "crack", the vestigial crack, although short, is very tight and the impedance is large. Under

  5. A review of recent advances in the role of leak-before-break concept in assessments of flaws detected in CANDU pressure tubes

    International Nuclear Information System (INIS)

    Crespi, J.C.

    1994-01-01

    If a crack develops in a pressure tube, the leak is detected by monitoring the moisture in the gas annulus and the reactor shutdown before it becomes unstable. Because the delayed hydride cracking has been associated to date with all pressure tube failures at a rolled joints, the delayed hydride cracking is considered to be the dominant mecanism by which the flaws can grow to a size which exceeds the critical crack length. For the delayed hydride cracking failure mode leak-before-break is used as defense in depth against unstable rupture. The methodology depends on showing than the time available to detect a delayed hydride crack is much greater that the time required to detect it in the gas annulus. The time available is estimated from measurements of: (a) axial delayed hydride crack growth rates, (b) crack lengths at penetrations of the tube wall when leakage first occurs and (c) critical crack lengths at instability when a crack is growing by the delayed hydride cracking mechanism. A review of recent advances in the experimental data used in leak-before-break assessment are presented and discussed. (author). 17 refs, 6 figs, 2 tabs

  6. Tensile cracks in creeping solids

    International Nuclear Information System (INIS)

    Riedel, H.; Rice, J.R.

    1979-02-01

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

  7. X-ray tubes

    International Nuclear Information System (INIS)

    Young, R.W.

    1979-01-01

    A form of x-ray tube is described which provides satisfactory focussing of the electron beam when the beam extends for several feet from gun to target. Such a tube can be used for computerised tomographic scanning. (UK)

  8. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

    Purpose: To increase the safety of pressure tube type reactors by providing an additional ECCS system to an ordinary ECCS system and injecting heavy water in the reactor core tank into pressure tubes upon fractures of the tubes. Constitution: Upon fractures of pressure tubes, reduction of the pressure in the fractured tubes to the atmospheric pressure in confirmed and the electromagnetic valve is operated to completely isolate the pressure tubes from the fractured portion. Then, the heavy water in the reactor core tank flows into and spontaneously recycles through the pressure tubes to cool the fuels in the tube to prevent their meltdown. By additionally providing the separate ECCS system to the ordinary ECCS system, fuels can be cooled upon loss of coolant accidents to improve the safety of the reactors. (Moriyama, K.)

  9. Gastrostomy feeding tube - bolus

    Science.gov (United States)

    Feeding - gastrostomy tube - bolus; G-tube - bolus; Gastrostomy button - bolus; Bard Button - bolus; MIC-KEY - bolus ... KEY, 3 to 8 weeks after surgery. These feedings will help your child grow strong and healthy. ...

  10. Feeding tube insertion - gastrostomy

    Science.gov (United States)

    ... this page: //medlineplus.gov/ency/article/002937.htm Feeding tube insertion - gastrostomy To use the sharing features on this page, please enable JavaScript. A gastrostomy feeding tube insertion is the placement of a feeding ...

  11. Neural Tube Defects

    Science.gov (United States)

    Neural tube defects are birth defects of the brain, spine, or spinal cord. They happen in the ... that she is pregnant. The two most common neural tube defects are spina bifida and anencephaly. In ...

  12. A computational model for reliability calculation of steam generators from defects in its tubes

    International Nuclear Information System (INIS)

    Rivero, Paulo C.M.; Melo, P.F. Frutuoso e

    2000-01-01

    Nowadays, probability approaches are employed for calculating the reliability of steam generators as a function of defects in their tubes without any deterministic association with warranty assurance. Unfortunately, probability models produce large failure values, as opposed to the recommendation of the U.S. Code of Federal Regulations, that is, failure probabilities must be as small as possible In this paper, we propose the association of the deterministic methodology with the probabilistic one. At first, the failure probability evaluation of steam generators follows a probabilistic methodology: to find the failure probability, critical cracks - obtained from Monte Carlo simulations - are limited to have length's in the interval defined by their lower value and the plugging limit one, so as to obtain a failure probability of at most 1%. The distribution employed for modeling the observed (measured) cracks considers the same interval. Any length outside the mentioned interval is not considered for the probability evaluation: it is approached by the deterministic model. The deterministic approach is to plug the tube when any anomalous crack is detected in it. Such a crack is an observed one placed in the third region on the plot of the logarithmic time derivative of crack lengths versus the mode I stress intensity factor, while for normal cracks the plugging of tubes occurs in the second region of that plot - if they are dangerous, of course, considering their random evolution. A methodology for identifying anomalous cracks is also presented. (author)

  13. Buckling Analysis of Edge Cracked Sandwich Plate

    Directory of Open Access Journals (Sweden)

    Rasha Mohammed Hussein

    2016-07-01

    Full Text Available This work presents mainly the buckling load of sandwich plates with or without crack for different cases. The buckling loads are analyzed experimentally and numerically by using ANSYS 15. The experimental investigation was to fabricate the cracked sandwich plate from stainless steel and PVC to find mechanical properties of stainless steel and PVC such as young modulus. The buckling load for different aspect ratio, crack length, cracked location and plate without crack found. The experimental results were compared with that found from ANSYS program. Present of crack is decreased the buckling load and that depends on crack size, crack location and aspect ratio.

  14. Nickel electroplating of steam generator tubes (kiss sleeving process)

    International Nuclear Information System (INIS)

    Michaut, B.

    1988-01-01

    This process, the nickel electroplating of steam generator tubes, has been jointly developed under a Belgatom (Laborelec) and Framatome agreement with shared experience gained by both companies, industrial applications being under the responsibility of Framatome. Application of the coating in zones where residual stresses or cracks are present prevents contact between the primary water and the tube, which stops the stress corrosion process. In the Doel 2 plant, 91 tubes have been plated since 1985, and different sets of parameters have been used for comparison purposes. Among these tubes, 9 have been preventively plugged because of defective plating, 9 have been pulled out for laboratory examinations, 2 just after plating and 7 after 1 or 2 yr of service. There are 73 plated tubes still in service. From the tests that were performed, it was possible to select an optimized set of parameters guaranteeing the following properties: bridging of existing cracks and good behavior of the coating in relevant zones, good adhesion to the Inconel tube, high ductility, low residual stresses, thermal shock resistance, corrosion resistance, erosion resistance, and low cobalt content. The licensability of this process is being completed. It is based first on the leak-before-break concept to determine the characteristics of the nickel plating, thickness in particular, and second on the inspectability of ultrasonic testing methods

  15. Stress Corrosion Cracking of alloy 600 in high temperature water: a study of mechanisms

    International Nuclear Information System (INIS)

    Boursier, J.M.; Bouvier, O. de; Gras, J.M.; Noel, D.; Vaillant, F.; Rios, R.

    1992-12-01

    Investigations of the stress corrosion cracking behaviour of Alloy 600 tubing in high temperature water were performed in order to get a precise knowledge of the different stages of the cracking and their dependence on various parameters. The compatibility of the results with the main mechanisms to be considered was examined. Results showed three stages in the cracking: a true incubation time, a slow-rate propagation period followed by a rapid-propagation stage. Tests separating stress and strain rate contributions show that the strain rate is the main parameter which controls the crack propagation. The hydrogen overpressure was found to increase the crack growth rate up to 1-4 bar, but a strong decrease is observed from 4 to 20 bar. Analysis of the hydrogen ingress in the metal showed that it is neither correlated to the hydrogen overpressure nor to the severity of cracking; so cracking resulting from an hydrogen-model is unlikely. No detrimental effect of oxygen (4 bar) was noticed both in the mill-annealed and the sensitized conditions. Finally, none of the classical mechanisms, neither hydrogen-assisted cracking nor slip-step dissolution, can correctly describe the observed behaviour. Some fractographic examinations, and an influence of primary water on the creep rate of Alloy 600, lead to consider that other recent mechanisms, involving an interaction between dissolution and plasticity, have to be considered

  16. Evaluation of options for life cycle management of feeder cracking at the Point Lepreau Generating Station

    International Nuclear Information System (INIS)

    Gendron, T.S.; Slade, J.P.

    2003-01-01

    The CANDU industry has a predictive capability for most Heat Transport System (HTS) degradation issues that allows utilities to apply cost-effective maintenance programs. The standard approach for maintenance programs is focussed inspection and planned replacement. Some examples of degradation issues with deterministic failure rates are feeder thinning, and pressure tube elongation and deuterium ingress. However, the cracking observed in Point Lepreau Generating Station (PLGS) outlet feeder first bends is one notable exception to this behaviour. A predictive capability for feeder cracking does not currently exist for several reasons. First, the mechanism of feeder cracking, stress corrosion cracking (SCC), has to some degree a random nature. Second, although a probable environment causing cracking has been identified, the precise stress and environmental conditions for feeder crack initiation and propagation have not been defined. Finally, the very low incidence of feeder cracking observed to-date (four, all at PLGS) precludes a probabilistic or statistical prediction of failure rate. Generally, utilities select a Life Cycle Management Plan that ensures safe operation and has the lowest Net Present Value cost. In preparing a Feeder Life Cycle Management Plan, New Brunswick Power (NBP) has recognized that the Net Present Value cost is very sensitive to failure rate. Since the failure rate for feeder cracking is not well defined, the following three scenarios were considered to bound the probability of future failures at PLGS. (author)

  17. Crack growth rate of PWR piping

    International Nuclear Information System (INIS)

    Bethmont, M.; Doyen, J.J.; Lebey, J.

    1979-01-01

    The Aquitaine 1 program, carried out jointly by FRAMATOME and the CEA is intended to improve knowledge about cracking mechanisms in AISI 316 L austenitic stainless steel under conditions similar to those of the PWR environment (irradiation excluded). Experiments of fatigue crack growth are performed on piping elements, scale 1/4 of primary pipings, by means of internal hydraulic cyclic pressure. Interpretation of results requires a knowledge of the stress intensity factor Ksub(I) at the front of the crack. Results of a series of calculations of Ksub(I) obtained by different methods for defects of finite and infinite length (three dimensional calculations) are given in the paper. The following have been used: calculations by finite elements, calculations by weight function. Notches are machined on the test pipes, which are subjected to internal hydraulic pressure cycles, under cold conditions, to initiate a crack at the tip of the notch. They are then cycled at a frequency of 4 cycles/hour on on water demineralised loop at a temperature of 280 0 C, the pressure varying at each cycle between approximately 160 bars and 3 bars. After each test, a specimen containing the defect is taken from the pipe for micrographic analysis. For the first test the length of the longitudinal external defect is assumed infinite. The number of cycles carried out is 5880 cycles. Two defects are machined in the tube for the second test. The number of cycles carried out is N = 440. The tests are performed under hot conditions (T = 280 0 C). For the third test two defects are analysed under cold and hot conditions. The number of cycles carried out for the external defect is 7000 when hot and 90000 when cold. The number of cycles for the internal defect is 1650 when hot and 68000 when cold. In order to interpret the results, the data da/dN are plotted on a diagram versus ΔK. Comparisons are made between these results and the curves from laboratory tests

  18. Evaluation of the residual stress field in a steam generator end tube after hydraulic expansion

    International Nuclear Information System (INIS)

    Thiel, F.; Kang, S.; Chabrerie, J.

    1994-01-01

    This paper presents a finite element elastoplastic model of a nuclear steam generator end tube, used to evaluate the residual stress field existing after hydraulic expansion of the tube into the tubesheet of the heat exchanger. This model has been tested against an experimental hydraulic expansion, carried out on full scale end tubes. The operation was monitored thanks to strain gages localized on the outer surface of the tubes, subjected to elastoplastic deformations. After a presentation of the expansion test and the description of the numerical model, the authors compare the stress fields issues from the gages and from the model. The comparison shows a good agreement. These results allow them to calculate the stress field resulting from normal operating conditions, while taking into account a correct initial state of stress. Therefore the authors can improve the understanding of the behavior of a steam generator end tube, with respect to stress corrosion cracking and crack growth

  19. Characterization of flaws in a tube bundle mock-up for reliability studies

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1997-01-01

    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

  20. Characterization of flaws in a tube bundle mock-up for reliability studies

    International Nuclear Information System (INIS)

    Kupperman, D.S.; Bakhtiari, S.

    1996-10-01

    As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

  1. 75 FR 66126 - Multilayered Wood Flooring From China

    Science.gov (United States)

    2010-10-27

    ...)] Multilayered Wood Flooring From China AGENCY: United States International Trade Commission. ACTION: Institution... flooring, provided for in subheadings 4409.10, 4409.29, 4412.31, 4412.32, 4412.39, 4412.94, 4412.99, 4418... multilayered wood flooring. The following companies are members of the CAHP: Anderson Hardwood Floors, LLC...

  2. 75 FR 79019 - Multilayered Wood Flooring From China

    Science.gov (United States)

    2010-12-17

    ...)] Multilayered Wood Flooring From China Determinations On the basis of the record \\1\\ developed in the subject... imports from China of multilayered wood flooring, provided for in subheadings 4409.10, 4409.29, 4412.31... multilayered wood flooring. The following companies are members of the CAHP: Anderson Hardwood Floors, LLC...

  3. 76 FR 76435 - Multilayered Wood Flooring From China

    Science.gov (United States)

    2011-12-07

    ...)] Multilayered Wood Flooring From China Determinations On the basis of the record \\1\\ developed in the subject... multilayered wood flooring, provided for in subheadings 4409.10, 4409.29, 4412.31, 4412.32, 4412.39, 4412.94... flooring. The following companies are members of the CAHP: Anderson Hardwood Floors, LLC, Fountain Inn, SC...

  4. Weld solidification cracking in 304 to 304L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Hochanadel, Patrick W [Los Alamos National Laboratory; Lienert, Thomas J [Los Alamos National Laboratory; Martinez, Jesse N [Los Alamos National Laboratory; Martinez, Raymond J [Los Alamos National Laboratory; Johnson, Matthew Q [Los Alamos National Laboratory

    2010-01-01

    A series of annulus welds were made between 304 and 304L stainless steel coaxial tubes using both pulsed laser beam welding (LBW) and pulsed gas tungsten arc welding (GTAW). In this application, a change in process from pulsed LBW to pulsed gas tungsten arc welding was proposed to limit the possibility of weld solidification cracking since weldability diagrams developed for GTAW display a greater range of compositions that are not crack susceptible relative to those developed for pulsed LBW. Contrary to the predictions of the GTAW weldability diagram, cracking was found. This result was rationalized in terms of the more rapid solidification rate of the pulsed gas tungsten arc welds. In addition, for the pulsed LBW conditions, the material compositions were predicted to be, by themselves, 'weldable' according to the pulsed LBW weldability diagram. However, the composition range along the tie line connecting the two compositions passed through the crack susceptible range. Microstructurally, the primary solidification mode (PSM) of the material processed with higher power LBW was determined to be austenite (A), while solidification mode of the materials processed with lower power LBW apparently exhibited a dual PSM of both austenite (A) and ferrite-austenite (FA) within the same weld. The materials processed by pulsed GT A W showed mostly primary austenite solidification, with some regions of either primary austenite-second phase ferrite (AF) solidification or primary ferrite-second phase austenite (FA) solidification. This work demonstrates that variations in crack susceptibility may be realized when welding different heats of 'weldable' materials together, and that slight variations in processing can also contribute to crack susceptibility.

  5. Weld solidification cracking in 304 to 204L stainless steel

    Energy Technology Data Exchange (ETDEWEB)

    Hochanadel, Patrick W [Los Alamos National Laboratory; Lienert, Thomas J [Los Alamos National Laboratory; Martinez, Jesse N [Los Alamos National Laboratory; Johnson, Matthew Q [Los Alamos National Laboratory

    2010-09-15

    A series of annulus welds were made between 304 and 304L stainless steel coaxial tubes using both pulsed laser beam welding (LBW) and pulsed gas tungsten arc welding (GTAW). In this application, a change in process from pulsed LBW to pulsed gas tungsten arc welding was proposed to limit the possibility of weld solidification cracking since weldability diagrams developed for GTAW display a greater range of compositions that are not crack susceptible relative to those developed for pulsed LBW. Contrary to the predictions of the GTAW weldability diagram, cracking was found.This result was rationalized in terms of the more rapid solidification rate of the pulsed gas tungsten arc welds. In addition, for the pulsed LBW conditions, the material compositions were predicted to be, by themselves, 'weldable' according to the pulsed LBW weldability diagram. However, the composition range along the tie line connecting the two compositions passed through the crack susceptible range. Microstructurally, the primary solidification mode (PSM) of the material processed with higher power LBW was determined to be austenite (A), while solidification mode of the materials processed with lower power LBW apparently exhibited a dual PSM of both austenite (A) and ferrite-austenite (FA) within the same weld. The materials processed by pulsed GTAW showed mostly primary austenite solidification, with some regions of either primary austenite-second phase ferrite (AF) solidification or primary ferrite-second phase austenite (FA) solidification. This work demonstrates that variations in crack susceptibility may be realized when welding different heats of 'weldable' materials together, and that slight variations in processing can also contribute to crack susceptibility.

  6. Investigation of floor Nusselt number in floor heating system for insulated ceiling conditions

    International Nuclear Information System (INIS)

    Karadag, Refet; Teke, Ismail

    2007-01-01

    In this study, in a floor heated room, natural convection heat transfer over the floor is analysed numerically for different thermal conditions. An equation relevant to Nusselt number over the floor has been obtained by using the numerical data. Different equations are given in the literature. They consider the effect of floor Rayleigh number while neglecting the effect of wall and ceiling thermal conditions. Numerical data obtained in this study show that the Nusselt number over the floor depends on not only the floor Rayleigh number but also the wall Rayleigh number (for insulated ceiling conditions). The equations given in the literature are different from each other due to their not considering the effect of wall and ceiling Rayleigh numbers. This difference between the equations may be eliminated by obtaining an equation containing the effect of floor, wall and ceiling Rayleigh numbers. In this new approach, an equation relevant to the floor Nusselt number that depends on the floor and wall Rayleigh numbers has been obtained in the floor heating system for insulated ceiling conditions. The equation obtained in this study has been compared with the equations given in the literature. It has been seen that the equation obtained in this study matches the numerical values under more extensive thermal conditions than the equations given in the literature. The maximum deviation for the equations given in the literature is 35%, but in the current study, the maximum deviation has been found to be 10%. As a result, it is more convenient to use the equation found in the new approach as a function of Rayleigh number over the floor and wall for insulated ceiling conditions

  7. Effect of pelvic floor rehabilitation technique in preventing the postpartum pelvic floor dysfunction

    Directory of Open Access Journals (Sweden)

    Shi-Qiong Li

    2017-04-01

    Full Text Available Objective: To explore the effect of pelvic floor rehabilitation technique in preventing the postpartum pelvic floor dysfunction and on the sexual life quality. Methods: A total of 286 puerpera with pelvic floor dysfunction who were admitted in our hospital from May, 2014 to May, 2015 42 d after delivery were included in the study, and randomized into the treatment group and the control group with 143 cases in each group. After guidance, the puerpera in the control group were given pelvic floor muscle training by themselves at home. On this basis, the puerpera in the treatment group were treated by the pelvic floor rehabilitation apparatus. The puerpera in the two groups were treated for 4 weeks. The pelvic floor function before treatment, 6 months and 1 year after delivery was detected. The color Doppler ultrasound apparatus was used to detect BSD, PUVA, UVJ-M, and BND 3 months after delivery. Results: BND, PUVA-R, PUVA-S, and UVJ-M 3 months after delivery in the treatment groups were significantly lower than those in the control group, while BSD-S was significantly higher than that in the control group. The improvement of type I and II muscle fiber fatigue (%, POP-Q degree, AP indication point (cm, and vaginal dynamic pressure (cmH2O was significantly superior to that in the control group. The comparison of pelvic floor muscle strength classification before treatment between the two groups was not statistically significant. After treatment, the pelvic floor muscle in the two groups was significantly strengthened, and the proportion of V grade patients was significantly increased when compared with before treatment. Conclusions: The postpartum early pelvic floor rehabilitation technique can effectively enhance the pelvic floor function, and prevent the postpartum pelvic floor dysfunction, with an accurate efficacy; therefore, it deserves to be widely recommended in the clinic.

  8. Influence of metallurgical variables on the velocity of crack propagation by delayed hydride cracking (DHC) in Zr-Nb

    International Nuclear Information System (INIS)

    Cirimelo, Pablo G.

    2002-01-01

    In the present thesis work the propagation of cracks due to the delayed hydride cracking (DHC) mechanism in Zr-2,5 % Nb pressure tubes is analyzed. For this purpose two different type of tubes of different origin were used: CANDU type (Canada) and RBMK type (Russia). The analyzed figurative parameters were: critical temperature Tc (highest temperature at which DHC phenomenon could occur) and crack propagation velocity by DHC, Vp, in the axial direction. The influence of the memory effect (phenomenon proper of hydride precipitation) was studied, as well as the type of cracks (fatigue or DHC) on Tc. However, no influence of these effects was found. Instead, it was found that Tc varies with the hydrogen content of the specimen, in agreement with previous works. Samples obtained from tubes with different microstructures and similar amounts of hydrogen presented similar Tc values. It was also shown that DHC propagation could occur without precipitated hydrides in the volume. Besides, Vp determinations were performed in temperature ranges and hydrogen amounts of technological importance. Two techniques were set up in order to determine Vp at different temperatures in a single specimen, thus saving time and material. An Arrhenius type variation was found for Vp vs. temperature, for temperatures lower than that corresponding to precipitation. For higher temperatures, but lower than the critical one, velocity decreases with temperature. Determination of Vp vs. temperature was performed for the two above-mentioned materials, whose microstructure and hardness were previously characterized. For RBMK material, which presents a spheroidal β phase, the velocity was lower than the corresponding to CANDU material, in which β phase is formed by continuous plates. In addition, yield stress σ Y is lower in RBMK material, which presents lower Vp. However, it is considered that the effect of microstructure is more important on Vp since it highly affects diffusion of hydrogen from the

  9. Analysis of short and long crack behavior and single overload effect by crack opening stress

    International Nuclear Information System (INIS)

    Song, Sam Hong; Lee, Kyeong Ro

    1999-01-01

    The study analyzed the behaviors of short and long crack as well as the effect of single tensile overload on the crack behaviors by using fatigue crack opening behavior. Crack opening stress is measured by an elastic compliance method which may precisely and continuously provide many data using strain gages during experiment. The unusual growth behaviors of short crack and crack after the single tensile overload applied, was explained by the variations of crack opening stress. In addition, fatigue crack growth rate was expressed as a linear form for short crack as for long crack by using effective stress intensity factor range as fracture mechanical parameter, which is based on crack closure concept. And investigation is performed with respect to the relation between plastic zone size formed at the crack tip and crack retardation, crack length and the number of cycles promoted or retarded, and the overload effect on the fatigue life

  10. Fatigue crack growth and endurance data on 9% Cr 1% Mo steels for AGR applications

    International Nuclear Information System (INIS)

    Priddle, E.K.

    1987-01-01

    Experimental investigations have been carried out on 9%Cr 1%Mo steels to examine: (1) The significance of carburisation on the fatigue endurance of plain and welded boiler tubes, and tube spacer strip; (2) the high cycle fatigue endurance of spacer strip and spacer weld metal; (3) fatigue crack growth rates in spacer strip and spacer weld metal. This report summarises the results of these investigations and where necessary compares the data to that in current data sheets. The effects of carburisation are variable depending on the structure and type of carburisation. The fatigue endurance properties of spacer strip and spacer weld metal are also similar and need not be considered separately for assessment or design purposes. Fatigue crack growth rates in spacer strip and space weld metal are similar and are influenced by both stress ratio and temperature. A design curve from a fast reactor data sheet may be used as an upper bound to these fatigue crack growth results. (author)

  11. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  12. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  13. 21 CFR 868.5800 - Tracheostomy tube and tube cuff.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Tracheostomy tube and tube cuff. 868.5800 Section... (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Therapeutic Devices § 868.5800 Tracheostomy tube and tube cuff. (a) Identification. A tracheostomy tube and tube cuff is a device intended to be placed into a...

  14. Probabilistic Analysis of Crack Width

    Directory of Open Access Journals (Sweden)

    J. Marková

    2000-01-01

    Full Text Available Probabilistic analysis of crack width of a reinforced concrete element is based on the formulas accepted in Eurocode 2 and European Model Code 90. Obtained values of reliability index b seem to be satisfactory for the reinforced concrete slab that fulfils requirements for the crack width specified in Eurocode 2. However, the reliability of the slab seems to be insufficient when the European Model Code 90 is considered; reliability index is less than recommended value 1.5 for serviceability limit states indicated in Eurocode 1. Analysis of sensitivity factors of basic variables enables to find out variables significantly affecting the total crack width.

  15. Multispecimen fatigue crack propagation testing

    International Nuclear Information System (INIS)

    Ermi, A.M.; Bauer, R.E.; Chin, B.A.; Straalsund, J.L.

    1981-01-01

    Chains of miniature center-cracked-tension specimens were tested on a conventional testing machine and on a prototypic in-reactor fatigue machine as part of the fusion reactor materials alloy development program. Annealed and 20 percent cold-worked 316 stainless steel specimens were cycled under various conditions of temperature, frequency, stress ratio and chain length. Crack growth rates determined from multispecimen visual measurements and from an electrical potential technique were consistent with those obtained by conventional test methods. Results demonstrate that multispecimen chain testing is a valid method of obtaining fatigue crack propagation information for alloy development. 8 refs

  16. Monitoring crack growth using thermography

    International Nuclear Information System (INIS)

    Djedjiga, Ait Aouita; Abdeldjalil, Ouahabi

    2008-01-01

    The purpose of this work is to present a novel strategy for real-time monitoring crack growth of materials. The process is based on the use of thermal data extracted along the horizontal axis of symmetry of single edge notch tension (SENT) specimens, during fatigue tests. These data are exploited using an implemented program to detect in situ the growth of fatigue crack, with the critical size and propagation speed of the crack. This technique has the advantage to be applicable to a wide range of materials regardless of their electrical conductivity and their surface texture. (authors)

  17. Password Cracking Using Sony Playstations

    Science.gov (United States)

    Kleinhans, Hugo; Butts, Jonathan; Shenoi, Sujeet

    Law enforcement agencies frequently encounter encrypted digital evidence for which the cryptographic keys are unknown or unavailable. Password cracking - whether it employs brute force or sophisticated cryptanalytic techniques - requires massive computational resources. This paper evaluates the benefits of using the Sony PlayStation 3 (PS3) to crack passwords. The PS3 offers massive computational power at relatively low cost. Moreover, multiple PS3 systems can be introduced easily to expand parallel processing when additional power is needed. This paper also describes a distributed framework designed to enable law enforcement agents to crack encrypted archives and applications in an efficient and cost-effective manner.

  18. 17 CFR 3.11 - Registration of floor brokers and floor traders.

    Science.gov (United States)

    2010-04-01

    ... 17 Commodity and Securities Exchanges 1 2010-04-01 2010-04-01 false Registration of floor brokers and floor traders. 3.11 Section 3.11 Commodity and Securities Exchanges COMMODITY FUTURES TRADING... a contract market or registered as a derivatives transaction execution facility by the Commission...

  19. Flooring-systems and their interaction with usage of the floor

    DEFF Research Database (Denmark)

    Pedersen, Lars; Frier, Christian; Andersen, Lars Vabbersgaard

    2017-01-01

    Some flooring-system designs might be sensitive to their vibrational performance, as there might be the risk that serviceability-limit-state problems may be encountered. For evaluating the vibrational performance of the flooring-system at the design stage, decisions need to be made by the enginee...

  20. No bulging of floor heating pipes to be expected in case of incomplete floor plastering

    Energy Technology Data Exchange (ETDEWEB)

    Radtke, U

    1983-02-01

    According to advertising slogans floor heating pipes are said to be damaged prematurely by bulges if they are not completely surrounded by flooring plaster. The author has thoroughly dealt with this problem and made the respective measurements. He found out that there are so few bulges occurring that they cannot lead to damages.

  1. Fatigue and environmentally assisted cracking in light water reactors

    International Nuclear Information System (INIS)

    Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

    1991-12-01

    Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

  2. Structural integrity evaluations of CANDU pressure tubes

    International Nuclear Information System (INIS)

    Radu, Vasile

    2003-01-01

    The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (K r and L r ). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables K r and L r , the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

  3. Heat exchanger tube tool

    International Nuclear Information System (INIS)

    Gugel, G.

    1976-01-01

    Certain types of heat-exchangers have tubes opening through a tube sheet to a manifold having an access opening offset from alignment with the tube ends. A tool for inserting a device, such as for inspection or repair, is provided for use in such instances. The tool is formed by a flexible guide tube insertable through the access opening and having an inner end provided with a connector for connection with the opening of the tube in which the device is to be inserted, and an outer end which remains outside of the chamber, the guide tube having adequate length for this arrangement. A flexible transport hose for internally transporting the device slides inside of the guide tube. This hose is long enough to slide through the guide tube, into the heat-exchanger tube, and through the latter to the extent required for the use of the device. The guide tube must be bent to reach the end of the heat-exchanger tube and the latter may be constructed with a bend, the hose carrying anit-friction elements at interspaced locations along its length to make it possible for the hose to negotiate such bends while sliding to the location where the use of the device is required

  4. Glazed Tiles as Floor Finish in Nigeria

    Directory of Open Access Journals (Sweden)

    Toyin Emmanuel AKINDE

    2013-09-01

    Full Text Available Tile is no doubt rich in antiquity; its primordial  show, came as mosaic with primary prospect in sacred floor finish before its oblivion, courtesy of, later consciousness towards wall finish in banquets, kitchens, toilets, restaurants and even bars. Today, its renaissance as floor finish is apparent in private and public architectural structures with prevalence in residential, recreational, commercial, governmental and other spaces. In Nigeria, the use of glazed tiles as floor finish became apparent, supposedly in mid-twentieth century; and has since, witnessed ever increasing demands from all sundry; a development that is nascent and has necessitated its mass  production locally with pockets of firms in the country. The latter however, is a resultant response to taste cum glazed tiles affordability, whose divergent sophistication in design, colour, size and shape is believed preferred to terrazzo, carpet and floor flex tile. Accessible as glazed tile and production is, in recent times; its dearth of a holistic literature in Nigeria is obvious. In the light of the latter, this paper examine glazed tiles as floor finish in Nigeria, its advent, usage, production, challenge, benefit and prospect with the hope of opening further frontier in discipline specifics.

  5. Global floor planning approach for VLSI design

    International Nuclear Information System (INIS)

    LaPotin, D.P.

    1986-01-01

    Within a hierarchical design environment, initial decisions regarding the partitioning and choice of module attributes greatly impact the quality of the resulting IC in terms of area and electrical performance. This dissertation presents a global floor-planning approach which allows designers to quickly explore layout issues during the initial stages of the IC design process. In contrast to previous efforts, which address the floor-planning problem from a strict module placement point of view, this approach considers floor-planning from an area planning point of view. The approach is based upon a combined min-cut and slicing paradigm, which ensures routability. To provide flexibility, modules may be specified as having a number of possible dimensions and orientations, and I/O pads as well as layout constraints are considered. A slicing-tree representation is employed, upon which a sequence of traversal operations are applied in order to obtain an area efficient layout. An in-place partitioning technique, which provides an improvement over previous min-cut and slicing-based efforts, is discussed. Global routing and module I/O pin assignment are provided for floor-plan evaluation purposes. A computer program, called Mason, has been developed which efficiently implements the approach and provides an interactive environment for designers to perform floor-planning. Performance of this program is illustrated via several industrial examples

  6. WOODEN FLOORING – BETWEEN PRESENT AND FUTURE

    Directory of Open Access Journals (Sweden)

    Ivan CISMARU

    2015-06-01

    Full Text Available The paper aims at presenting a systematization of the wood floors, both in terms of the areas of application, and in terms of the fastening solutions and structures in constructions. In this respect, an extensive bibliographic research was achieved, on the researchers’ preoccupations. Starting from the current situation and forecasting the future, from the point of view of the chances held by wooden flooring, in competition with other types of materials, we dare say the wooden flooring or the wood in combination with other materials are not likely to be eliminated from the “civil-engineering market”. The wood floors are likely to develop as an application, especially in the area of the “special floors”, specific to the indoor sports or social halls; and even for some industrial sectors, with strict operating conditions (elasticity, thermal insulation, soundproofing that cannot be provided by other types of materials or structures. Starting from this last observation, the paper also aims at submitting current opinions with respect to this type of floors, both in the light of the current databases and in the light of the future researches, to this end

  7. Subsurface metals fatigue cracking without and with crack tip

    Directory of Open Access Journals (Sweden)

    Andrey Shanyavskiy

    2013-07-01

    Full Text Available Very-High-Cycle-Fatigue regime for metals was considered and mechanisms of the subsurface crack origination were introduced. In many metals first step of crack origination takes place with specific area formation because of material pressing and rotation that directed to transition in any volume to material ultra-high-plasticity with nano-structure appearing. Then by the border of the nano-structure takes place volume rotation and fracture surface creates with spherical particles which usually named Fine-Granular-Area. In another case there takes place First-Smooth-Facet occurring in area of origin due to whirls appearing by the one of the slip systems under discussed the same stress-state conditions. Around Fine-Granular-Area or First-Smooth-Facet there plastic zone appeared and, then, subsurface cracking develops by the same manner as for through cracks. In was discussed quantum-mechanical nature of fatigue crack growth in accordance with Yang’s modulus quantization for low level of deformations. New simply equation was considered for describing subsurface cracking in metals out of Fine-Granular-Area or Fist-Smooth-Facet.

  8. Evaluation of a steam generator tube repair process using an explosive expansion techniuqe at TMI-1

    International Nuclear Information System (INIS)

    Rajan, J.; Shook, T.A.; Leonard, L.

    1983-01-01

    After a planned shutdown of Unit No. 1 at Three Mile Island, cracks were discovered in the primary side of steam generator tubes in the vicinity of the upper surface of the upper tubesheet. The nature of these cracks was later characterized as intergranular stress corrosion. The licensee, General Public Utilities Nuclear (GPUN), proposed to form a new tube-to-tubesheet seal below the cracks using a repair process wherein a detonating cord and polyethylene cartridge assembly inserted into the tube explosively expand the tube against the tubesheet. The explosive expansion process has had numerous applications over the years in the initial fabrication of heat exchanger tube-to-tubesheet assemblies and in repair processes using sleeving. However, this is the first use of this process in a steam generator to expand a previously rolled tube and to form a new seal between it and the tubesheet below a defective region in the tube. The seal obtained between the tube and tubesheet depends on the magnitude of explosive energy released in the detonating process. In this application, it is desired to obtain a mechanical bond rather than a metallurgical welding of the tube and tubesheet. A number of critical variables must be taken into account in order to obtain a successful mechanical seal. These include the explosive power of the detonating cord, the number of expansion shots used, the length of tube which is expanded, cartridge and tube diameters, the diameter of the tubesheet hole, the materials of the tube and tubesheet, and the condition of the surfaces at the time of repair. (orig./GL)

  9. Effects of Floor Covering Resistance of a Radiant Floor on System Energy and Exergy Performances

    DEFF Research Database (Denmark)

    Kazanci, Ongun Berk; Shukuya, Masanori; Olesen, Bjarne W.

    2016-01-01

    Floor covering resistance (material and thickness) can be influenced by subjective choices (architectural design, interior design, texture, etc.) with significant effects on the performance of a radiant heating and cooling system. To study the effects of floor covering resistance on system...... performance, a water-based radiant floor heating and cooling system (dry, wooden construction) was considered to be coupled to an air-to-water heat pump, and the effects of varying floor covering resistances (0.05 m2K/W, 0.09 m2K/W and 0.15 m2K/W) on system performance were analyzed in terms of energy...... and exergy. In order to achieve the same heating and cooling outputs, higher average water temperatures are required in the heating mode (and lower temperatures in the cooling mode) with increasing floor covering resistance. These temperature requirements decrease the heat pump’s performance (lower...

  10. Intercostal drainage tube or intracardiac drainage tube?

    Directory of Open Access Journals (Sweden)

    N Anitha

    2016-01-01

    Full Text Available Although insertion of chest drain tubes is a common medical practice, there are risks associated with this procedure, especially when inexperienced physicians perform it. Wrong insertion of the tube has been known to cause morbidity and occasional mortality. We report a case where the left ventricle was accidentally punctured leading to near-exsanguination. This report is to highlight the need for experienced physicians to supervise the procedure and train the younger physician in the safe performance of the procedure.

  11. Intercostal drainage tube or intracardiac drainage tube?

    Science.gov (United States)

    Anitha, N; Kamath, S Ganesh; Khymdeit, Edison; Prabhu, Manjunath

    2016-01-01

    Although insertion of chest drain tubes is a common medical practice, there are risks associated with this procedure, especially when inexperienced physicians perform it. Wrong insertion of the tube has been known to cause morbidity and occasional mortality. We report a case where the left ventricle was accidentally punctured leading to near-exsanguination. This report is to highlight the need for experienced physicians to supervise the procedure and train the younger physician in the safe performance of the procedure.

  12. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... YouTube Videos » NEI YouTube Videos: Amblyopia Listen NEI YouTube Videos YouTube Videos Home Age-Related Macular Degeneration ... Retinopathy of Prematurity Science Spanish Videos Webinars NEI YouTube Videos: Amblyopia Embedded video for NEI YouTube Videos: ...

  13. 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

  14. Super oil cracking update

    International Nuclear Information System (INIS)

    Mulraney, D.

    1997-01-01

    The conversion of residual fuel oil to usable middle distillates was discussed. The residue conversion processing paths are usually based on separation, carbon rejection, or hydrogen addition principles. Super Oil Cracking (SOC) uses a slurry catalyst system in a new, tubular reactor to achieve high levels of hydrothermal conversion. SOC can upgrade a variety of heavy, high metals residue feedstocks with high yields of middle distillates. The SOC products can also be further treated into feedstocks for FCC or hydrocracking. The SOC process can be incorporated easily into a refinery to obtain incremental residue conversion directly. It can also be integrated with other residue processes, acting as a demetallization and decarbonization step which results in enhanced overall conversion. The relative rate of coke formation and its handling are distinguishing characteristics between residue upgrading technologies. The SOC process operates at higher temperatures that other residue hydrocracking processes resulting in higher rates of thermal decomposition, thus preventing coke formation. SOC process can operate as a stand-alone upgrader or can be integrated with other bottoms processing steps to extend the refiner's range of options for increasing bottoms conversion.3 tabs., 14 figs

  15. Metallurgy of stress corrosion cracking

    International Nuclear Information System (INIS)

    Donovan, J.A.

    1973-01-01

    The susceptibility of metals and alloys to stress corrosion is discussed in terms of the relationship between structural characteristics (crystal structure, grains, and second phases) and defects (vacancies, dislocations, and cracks) that exist in metals and alloys. (U.S.)

  16. Peridynamic model for fatigue cracking.

    Energy Technology Data Exchange (ETDEWEB)

    Silling, Stewart Andrew; Abe Askari (Boeing)

    2014-10-01

    The peridynamic theory is an extension of traditional solid mechanics in which the field equations can be applied on discontinuities, such as growing cracks. This paper proposes a bond damage model within peridynamics to treat the nucleation and growth of cracks due to cyclic loading. Bond damage occurs according to the evolution of a variable called the "remaining life" of each bond that changes over time according to the cyclic strain in the bond. It is shown that the model reproduces the main features of S-N data for typical materials and also reproduces the Paris law for fatigue crack growth. Extensions of the model account for the effects of loading spectrum, fatigue limit, and variable load ratio. A three-dimensional example illustrates the nucleation and growth of a helical fatigue crack in the torsion of an aluminum alloy rod.

  17. Shapes formed by interacting cracks

    Science.gov (United States)

    Daniels, Karen

    2012-02-01

    Brittle failure through multiple cracks occurs in a wide variety of contexts, from microscopic failures in dental enamel and cleaved silicon to geological faults and planetary ice crusts. In each of these situations, with complicated stress geometries and different microscopic mechanisms, pairwise interactions between approaching cracks nonetheless produce characteristically curved fracture paths. We investigate the origins of this widely observed ``en passant'' crack pattern by fracturing a rectangular slab which is notched on each long side and subjected to quasi-static uniaxial strain from the short side. The two cracks propagate along approximately straight paths until they pass each other, after which they curve and release a lens-shaped fragment. We find that, for materials with diverse mechanical properties, each curve has an approximately square-root shape, and that the length of each fragment is twice its width. We are able to explain the origins of this universal shape with a simple geometrical model.

  18. Stress-Assisted Corrosion in Boiler Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Preet M Singh; Steven J Pawel

    2006-05-27

    A number of industrial boilers, including in the pulp and paper industry, needed to replace their lower furnace tubes or decommission many recovery boilers due to stress-assisted corrosion (SAC) on the waterside of boiler tubes. More than half of the power and recovery boilers that have been inspected reveal SAC damage, which portends significant energy and economic impacts. The goal of this project was to clarify the mechanism of stress-assisted corrosion (SAC) of boiler tubes for the purpose of determining key parameters in its mitigation and control. To accomplish this in-situ strain measurements on boiler tubes were made. Boiler water environment was simulated in the laboratory and effects of water chemistry on SAC initiation and growth were evaluated in terms of industrial operations. Results from this project have shown that the dissolved oxygen is single most important factor in SAC initiation on carbon steel samples. Control of dissolved oxygen can be used to mitigate SAC in industrial boilers. Results have also shown that sharp corrosion fatigue and bulbous SAC cracks have similar mechanism but the morphology is different due to availability of oxygen during boiler shutdown conditions. Results are described in the final technical report.

  19. Heat exchanger tube inspection using ultrasonic arrays

    International Nuclear Information System (INIS)

    Meyer, P.A.; Carodiskey, T.J.

    1986-01-01

    Tubing used in industrial heat exchangers is often subject to failure caused by corrosion and cracking. Technical conferences are used as a forum in the steam generator industry to ensure that the failure mechanisms are well understood and that the quality of the heat exchanger is maintained. The quality of a heat exchanger can be thought of as its ability to operate to design specifications over its intended life. This is the motivation to inspect and evaluate these devices periodically. Inspection, however, normally requires shutdown of the heat exchanger which is costly but is much more acceptable than an unscheduled shutdown due to failure of a tube. Therefore, the degree of inspection is established by balancing the cost of inspection with the risk of a tube failure. Any method of reducing the cost of inspection will permit a higher degree of inspection and, therefore, improve heat exchanger quality. This paper reviews the design and performance of an improved method of ultrasonic inspection of heat exchanger tubing with emphasis on applications in the nuclear industry

  20. The crack growth mechanism in asphaltic mixes

    NARCIS (Netherlands)

    Jacobs, M.M.J.; Hopman, P.C.; Molenaar, A.A.A.

    1995-01-01

    The crack growth mechanism in asphalt concrete (Ac) mixes is studied. In cyclic tests on several asphaltic mixes crack growth is measured, both with crack foils and with cOD-gauges. It is found that crack growth in asphaltic mixes is described by three processes which are parallel in time: cohesive

  1. Dynamic Crack Branching - A Photoelastic Evaluation,

    Science.gov (United States)

    1982-05-01

    0.41 mPai and a 0.18 MPa, and predicted a theoretical kinking angle of 84°whichagreed well with experimentally measured angle. After crack kinking...Consistent crack branching’at KIb = 2.04 MPaI -i- and r = 1.3 mm verified this crack branching criterion. The crack branching angle predicted by--.’ DD

  2. 21 CFR 137.190 - Cracked wheat.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Cracked wheat. 137.190 Section 137.190 Food and... Related Products § 137.190 Cracked wheat. Cracked wheat is the food prepared by so cracking or cutting into angular fragments cleaned wheat other than durum wheat and red durum wheat that, when tested by...

  3. Structural evaluation of electrosleeved tubes under severe accident transients

    International Nuclear Information System (INIS)

    Majumdar, S.

    1999-01-01

    A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients

  4. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    International Nuclear Information System (INIS)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

    2012-01-01

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  5. Nickel electroplating as a remedy to steam generator tubing PWSCC

    International Nuclear Information System (INIS)

    Michaut, B.; Steltzlen, F.; Sala, B.; Laire, Ch.; Stubbe, J.

    1993-01-01

    Nickel plating appears to be a versatile process, as the application field, even if always used against PWSCC, is different from plant-to-plant. Its usage has been from a purely preventive action on tubes without defects, to a corrective action on through-wall cracked and leaking tubes. As a background for the large scale on-site operations of Doel 2 in 1990 (345 tubes) and Tihange 2 in 1992 (600 tubes), studies on four points are outlined, i.e. corrosion tests, stress measurements, sulfamate bath quality control, and in-service inspection. In conclusion, it appears that the nickel plating technique, following a case-by-case study, can often be a convenient remedy against Alloy 600 stress corrosion problems. New applications, in locations other than the steam generator field are under consideration

  6. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced.

  7. Comparison of evaluation method for planar flaw in pressure tube

    International Nuclear Information System (INIS)

    Choi, Sung Nam; Kim, Hyung Nam; Yoo, Hyun Joo; Hwang, Won Gul

    2009-01-01

    CSA N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA N285.8-05, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of FFSG(Fitness For Service Guideline for Zirconium alloy pressure in operation CANDU) used now. The object of this paper is to address the fracture initiation and plastic collapse evaluation for the planar flaw as it applies to the pressure tube on Wolsong NPP.

  8. PWSCC in the tube expansion zone - an overview

    International Nuclear Information System (INIS)

    Hernalsteen, P.

    1993-01-01

    Most of the PWR Steam Generators (SG) with tubes in Inconel 600 alloy are affected by Primary Water Stress Corrosion Cracking (PWSCC) in the Expansion Zone (mainly the Roll Transition) of tubes mechanically expanded in the tube sheet. After a description of the defect mechanism and characterization methods, the paper reviews the various measures that can be used to prevent the problem. In-Service Inspection results are presented to illustrate the actual field experience; prediction tools are available to forecast the further SG degradation. Degraded tubes are eventually removed from service; this plugging policy undergoes presently a major evolution towards a mechanism specific approach, taking into account both structural and leakage requirements. The paper reviews various repair techniques that can be used as an alternate to plugging. Ultimately repair has to be weighed against SG replacement with a comprehensive problem management approach. (orig.)

  9. Dermoid cyst in the mouth floor

    International Nuclear Information System (INIS)

    Portelles Masso, Ayelen Maria; Torres Inniguez, Ailin Tamara.

    2010-01-01

    The Dermoid cyst account for the 0.01 % of all cysts of buccal cavity. Its more frequent location is in the mouth floor. This is the case of a female patient aged 19 who approximately 7 years noted an increase of volume under tongue growing gradually and noting outside face and the discomfort at to speak and to chew. Complementary studies were conducted and under general anesthesia a surgical exeresis was carried out by intrabuccal approach achieving excellent esthetic and functional results. Histopathologic diagnosis matched with a dermoid cyst of mouth floor. Patient has not lesion recurrence after three years after operation. We conclude that the Dermoid cyst of mouth floor appear as benign tumor of middle line. The intrabuccal exeresis demonstrates esthetic and functional benefits. (author)

  10. Imaging of the posterior pelvic floor

    International Nuclear Information System (INIS)

    Stoker, Jaap; Bartram, Clive I.; Halligan, Steve

    2002-01-01

    Disorders of the posterior pelvic floor are relatively common. The role of imaging in this field is increasing, especially in constipation, prolapse and anal incontinence, and currently imaging is an integral part of the investigation of these pelvic floor disorders. Evacuation proctography provides both structural and functional information for rectal voiding and prolapse. Dynamic MRI may be a valuable alternative as the pelvic floor muscles are visualised, and it is currently under evaluation. Endoluminal imaging is important in the management of anal incontinence. Both endosonography and endoanal MRI can be used for detection of anal sphincter defects. Endoanal MRI has the advantage of simultaneously evaluating external sphincter atrophy, which is an important predictive factor for the outcome of sphincter repair. Many aspects of constipation and prolapse remain incompletely understood and treatment is partly empirical; however, imaging has a central role in management to place patients into treatment-defined groups. (orig.)

  11. Development of rationalized system treating floor drain

    International Nuclear Information System (INIS)

    Nakamura, Yasuyuki; Serizawa, Kenichi; Komatsu, Akihiro; Shimizu, Takayuki

    1998-01-01

    Radioactive liquid wastes generated at BWR plants are collected and treated as required. These days, however, generation of floor drain has deceased and HFF (Hollow Fiber Filter) has experienced a wide applicability to several kinds of liquid wastes. We should consider that the floor drain can be mixed and diluted with equipment drain and be purified by HFF. That enables some of the sumps and long priming pipes to be combined. From this point of view, we have developed a highly rationalized waste liquid system. We have evaluated the applicability of this system after an investigation into the generation and properties of floor drain and equipment drain at the latest BWR'S and an on-site test at a typical BWR. (author)

  12. Imaging of the posterior pelvic floor

    Energy Technology Data Exchange (ETDEWEB)

    Stoker, Jaap [Department of Radiology, Academic Medical Center, University of Amsterdam (Netherlands); Bartram, Clive I.; Halligan, Steve [Intestinal Imaging Centre, St. Mark' s Hospital, London (United Kingdom)

    2002-04-01

    Disorders of the posterior pelvic floor are relatively common. The role of imaging in this field is increasing, especially in constipation, prolapse and anal incontinence, and currently imaging is an integral part of the investigation of these pelvic floor disorders. Evacuation proctography provides both structural and functional information for rectal voiding and prolapse. Dynamic MRI may be a valuable alternative as the pelvic floor muscles are visualised, and it is currently under evaluation. Endoluminal imaging is important in the management of anal incontinence. Both endosonography and endoanal MRI can be used for detection of anal sphincter defects. Endoanal MRI has the advantage of simultaneously evaluating external sphincter atrophy, which is an important predictive factor for the outcome of sphincter repair. Many aspects of constipation and prolapse remain incompletely understood and treatment is partly empirical; however, imaging has a central role in management to place patients into treatment-defined groups. (orig.)

  13. Crack propagation in dynamic thermoelasticity

    International Nuclear Information System (INIS)

    Bui, H.D.

    1980-01-01

    We study the singular thermoelastic fields near the crack tip, in the linear strain assumption. The equations are coupled and non linear. The asymptotic expansions of the displacement and the temperature are given for the first and the second order. It is shown that the temperature is singular when the crack propagates. However, this field does not change the dominant singularity of the mechanical field which is the same as that obtained in the theory of isothermal elasticity [fr

  14. H2S cracking resistance of type 420 stainless steel tubulars

    International Nuclear Information System (INIS)

    Klein, L.J.

    1984-01-01

    Type 420 stainless steel (13Cr) production tubing is being used successfully in deep sour gas wells in the Tuscaloosa Trend. Despite their reputation for poor H 2 S cracking resistance in laboratory tests, 12-13% Cr steels continue to perform well in sour environments. NACE Tensile Test and Shell bent beam test results indicate Type 420 is more resistant to H 2 S cracking than Type 410, but is not as resistant as carbon steel, at to 586-690 MPa (85-100 ksi) yield strength level. In addition to evaluating Type 420 stainless steel in the standard NACE Tensile and Shell bent beam tests, the effects on cracking tendency of chloride concentration, pH, and H 2 S gas concentration in the NACE Test solution were also examined. Type 420 appears to be more resistant to H 2 S cracking than is indicated by standard laboratory tests, at least in low H 2 S level sour environments

  15. A model of single and two-phase flow (critical or not) through cracks

    International Nuclear Information System (INIS)

    Seynhaeve, J.M.; Giot, M.; Granger, S.; Pages, D.

    1995-07-01

    The leaks through steam-generator cracks are the subject of research carried out in cooperation between EDF and UCL. A model to predict the mass flow rate with inlet subcooling has been developed, validated and published. The model takes into account the persistence of some metastable liquid in the crack. The present paper improves and extends the model, by making it applicable to all kinds of conditions prevailing in the S.G. tubes: not only subcooled water, but also saturated water, steam-water mixtures, saturated dry steam or superheated steam. Therefore, the flow at the crack inlet is analyzed and appropriate methods to initialize the numerical integration of the flow equations along the crack are proposed. The extensions of the model are still in the process of validation. However, a sensitivity analysis of its results has been made and is presented. (author)

  16. Method for the protection of the cladding tubes of fuel rods

    International Nuclear Information System (INIS)

    Steinberg, E.

    1978-01-01

    To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. (DG) [de

  17. Fretting wear damage of steam generator tubes and its prediction modeling

    International Nuclear Information System (INIS)

    Che Honglong; Lei Mingkai

    2013-01-01

    The steam generator is the key equipment used for the energy transition in nuclear power plant. Since the high-temperature and high-pressure fluid flows with high speed, the steam generator tubes will be excited and vibrate, leading to the tremendous fretting wear problem on the tubes, sometimes even leading to tube cracking. This paper introduces typical fretting wear cases, the result of corresponding simulation wear experiment and damage mechanism which combining mechanical wear and erosion-corrosion. Work rate model could give a reasonable life prediction about the steam generator tube, and this predictive model has been used in nuclear power plant safety assessment. (authors)

  18. Properties and application study of Inconel alloy tube made in China

    International Nuclear Information System (INIS)

    Yang Xiang; Su Xingwan; Wen Yan

    1997-01-01

    The mech-physical properties and the corrosion resistance properties of the SG tube of Inconel alloy made in China under any conditions are briefly presented, and the test and research for bending and expending the tubes have been performed. In the process of corrosion experiments the Inconel alloy tubes were compared with that of the same kind of materials made in foreign countries. The Inconel alloy tubes have better stress corrosion resistance cracking prosperities than Inconel 600 and Incoloy 800 when they were in the solutions which contained high concentrated chlorine ion and alkali at high temperature

  19. The resistance to PWSCC of explosively expanded Alloy 600 tube-to-tubesheet joints

    International Nuclear Information System (INIS)

    Gold, R.E.; Pement, F.W.; Tarabek, S.A.; Economy, G.

    1992-01-01

    Experimental evaluations were performed to determine the approximate magnitude of the residual stresses associated with explosively expanded steam generator tubing, and to assess the resistance to primary water stress corrosion cracking (PWSCC) of these expansions. Indexing of residual stresses was performed by means of magnesium chloride exposures of surrogate stainless steel mockups. The PWSCC resistance was evaluated by the testing of pressurized mockups of explosively expanded mill annealed Alloy 600 tubing in a highly accelerated Alloy 600 tubing in a highly accelerated steam test environment. Shot peening of the inside tube surfaces was demonstrated to be effective in modifying the residual stresses, providing additional resistance to PWSCC

  20. Environmentally assisted cracking of light-water reactor materials

    International Nuclear Information System (INIS)

    Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

    1996-02-01

    Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used

  1. Evaluation of crack interaction effect for in-plane surface cracks using elastic finite element analyses

    International Nuclear Information System (INIS)

    Huh, Nam Su; Choi, Suhn; Park, Keun Bae; Kim, Jong Min; Choi, Jae Boong; Kim, Young Jin

    2008-01-01

    The crack-tip stress fields and fracture mechanics assessment parameters, such as the elastic stress intensity factor and the elastic-plastic J-integral, for a surface crack can be significantly affected by adjacent cracks. Such a crack interaction effect due to multiple cracks can magnify the fracture mechanics assessment parameters. There are many factors to be considered, for instance the relative distance between adjacent cracks, crack shape and loading condition, to quantify a crack interaction effect on the fracture mechanics assessment parameters. Thus, the current guidance on a crack interaction effect (crack combination rule), including ASME Sec. XI, BS7910, British Energy R6 and API RP579, provide different rules for combining multiple surface cracks into a single surface crack. The present paper investigates a crack interaction effect by evaluating the elastic stress intensity factor of adjacent surface cracks in a plate along the crack front through detailed 3-dimensional elastic finite element analyses. The effects of the geometric parameters, the relative distance between cracks and the crack shape, on the stress intensity factor are systematically investigated. As for the loading condition, only axial tension is considered. Based on the elastic finite element results, the acceptability of the crack combination rules provided in the existing guidance was investigated, and the relevant recommendations on a crack interaction for in-plane surface cracks in a plate were discussed

  2. Pediatric cuffed endotracheal tubes

    Directory of Open Access Journals (Sweden)

    Neerja Bhardwaj

    2013-01-01

    Full Text Available Endotracheal intubation in children is usually performed utilizing uncuffed endotracheal tubes for conduct of anesthesia as well as for prolonged ventilation in critical care units. However, uncuffed tubes may require multiple changes to avoid excessive air leak, with subsequent environmental pollution making the technique uneconomical. In addition, monitoring of ventilatory parameters, exhaled volumes, and end-expiratory gases may be unreliable. All these problems can be avoided by use of cuffed endotracheal tubes. Besides, cuffed endotracheal tubes may be of advantage in special situations like laparoscopic surgery and in surgical conditions at risk of aspiration. Magnetic resonance imaging (MRI scans in children have found the narrowest portion of larynx at rima glottides. Cuffed endotracheal tubes, therefore, will form a complete seal with low cuff pressure of <15 cm H 2 O without any increase in airway complications. Till recently, the use of cuffed endotracheal tubes was limited by variations in the tube design marketed by different manufacturers. The introduction of a new cuffed endotracheal tube in the market with improved tracheal sealing characteristics may encourage increased safe use of these tubes in clinical practice. A literature search using search words "cuffed endotracheal tube" and "children" from 1980 to January 2012 in PUBMED was conducted. Based on the search, the advantages and potential benefits of cuffed ETT are reviewed in this article.

  3. Modal analysis for floors in lightweight buildings

    DEFF Research Database (Denmark)

    Sjökvist, Lars-Göran; Brunskog, Jonas

    2007-01-01

    of acoustical prediction methods for those houses. The calculation standard EN 12354 is under evaluation since it cannot include most of the wooden houses that are built. It is important during such a work to have a great understanding of the acoustical behaviour for the wooden houses. The floors in lightweight...... constructions usually consist of plates that are stiffened by beams and by the dividing walls. In this study the wave equation for a plate is expanded by Fourier series and an analytical solution in terms of the eigenmodes of the entire system is presented. The studied system consists of one lightweigt floor...

  4. Coatings and floor covers for nuclear applications

    International Nuclear Information System (INIS)

    Kunze, S.

    1998-01-01

    To prevent damage to, or even the destruction of, components of very sensitive electrical equipment in rooms in which unsealed radioactive emitters are handled, floors must be antistatic and capable of being decontaminated. Conductive additives to the cover compounds achieve the desired leakage resistance of 5.10 4 to 10 6 Ω. Investigations have shown the decontamination capability of all floor covers and coatings to be excellent in most cases, and good in a few cases. Except for one coating, the coatings examined after radiation exposure also meet the requirements applying to nuclear installations. (orig.) [de

  5. Some Passive Damping Sources on Flooring Systems besides the TMD

    DEFF Research Database (Denmark)

    Pedersen, Lars

    2010-01-01

    Impulsive loads and walking loads can generate problematic structural vibrations in flooring-systems. Measures that may be taken to mitigate the problem would often be to consider the implementation of a tuned mass damper or even more advanced vibration control technologies; this in order to add...... damping to the structure. Basically also passive humans on a floor act as a damping source, but it also turns out from doing system identification tests with a floor strip that a quite simple set-up installed on the floor (cheap and readily at hand) might do a good job in terms of reducing vertical floor...... vibrations for some floors. The paper describes the tests with the floor strip, and the results, in terms of dynamic floor behaviour, are compared with what would be expected had the floor instead been equipped with a tuned mass damper....

  6. Lessons learned from tubes pulled from French steam generators

    International Nuclear Information System (INIS)

    Berge, Ph.; Boursier, J.M.; Dallery, D.; De Keroulas, F.; Rouillon, Y.

    1998-01-01

    Since 1981, the Chinon Hot Laboratory has completed more than 380 metallurgical examinations of pulled French steam generator tubes. Electricite de France decided to perform such investigations from the very outset of the French nuclear program, in order to contribute to nuclear power plant safety. The main reasons for withdrawing tubes are to evaluate the degradation, to validate non destructive examination (NDE) techniques, to gain a better understanding of cracking phenomena, and to ensure that the criteria on which plugging operations are based remain conservative. Considerable experience has been accumulated in the field of primary water stress corrosion cracking (PWSCC), OD (secondary) side corrosion, leak and burst tests, and various tube plugging techniques. This paper focuses on the PWSCC phenomenon and on the secondary side corrosion process, and in particular, attempts to correlate French data from pulled tubes with the results of fundamental R and D studies. Finally, within the framework of the Nuclear Power Plant Safety and Maintenance Policy, all these results are discussed in terms of optimization of the field inspection of tube bundles and plugging criteria. (author)

  7. Evaluation of reliability of EC inspection of VVER SG tubes

    International Nuclear Information System (INIS)

    Stanic, D.

    2001-01-01

    Evaluation of eddy current data collected during inspection of VVER steam generators is very complex task because of numerous parameters which have affect on eddy current signals. That was the reason that recently ago INETEC has started related scientific project in order to evaluate the reliability of eddy current (EC) inspection of VVER steam generator (SG) tubing. In the scope of project the following objectives will be investigated: 1. Determination of POD (Probability of detection) of various types degradation cracks, where their basic parameters are variables (basic parameters are depth, length, width, orientation, number) on three different sets of tubes (clean ideal tubes, tubes with pilgering, tubes electroplated with copper) 2. Sizing quality (accuracy, repeatability) (same data sets as defined in 1.) 3. Effect of fill factor on POD and sizing quality. 4. Effect of tube bends on POD and sizing quality. 5. Effect of other tube geometry variations on POD and sizing quality (tube ovality, transition zone region, expanded (rolled) part of tube, dents, dings). Investigation will start with bobbin probe technique which is the most used technique for general purpose VVER tube examination. Since INETEC is the only world company which successfully developed and applied rotating probe technique for VVER SG tubes, scope of the project will be extended on rotating probe technique utilizing 'pancake' and 'point' coil. Method reliability will be investigated first on the huge set of EDM notches representing various defect morphologies and simulating different factors, and the second part will be investigated on sets of degradation defects obtained by artificial corrosion. In the scope of the project the measures for enhancing the method reliability have to be determined. This considers the proper definition of parameters of examination system, as well as establishment of the suitable analysis procedures. This article presents the temporary results of the first part of

  8. Compressive failure with interacting cracks

    International Nuclear Information System (INIS)

    Yang Guoping; Liu Xila

    1993-01-01

    The failure processes in concrete and other brittle materials are just the results of the propagation, coalescence and interaction of many preexisting microcracks or voids. To understand the real behaviour of the brittle materials, it is necessary to bridge the gap from the relatively matured one crack behaviour to the stochastically distributed imperfections, that is, to concern the crack propagation and interaction of microscopic mechanism with macroscopic parameters of brittle materials. Brittle failure in compression has been studied theoretically by Horii and Nemat-Nasser (1986), in which a closed solution was obtained for a preexisting flaw or some special regular flaws. Zaitsev and Wittmann (1981) published a paper on crack propagation in compression, which is so-called numerical concrete, but they did not take account of the interaction among the microcracks. As for the modelling of the influence of crack interaction on fracture parameters, many studies have also been reported. Up till now, some researcher are working on crack interaction considering the ratios of SIFs with and without consideration of the interaction influences, there exist amplifying or shielding effects of crack interaction which are depending on the relative positions of these microcracks. The present paper attempts to simulate the whole failure process of brittle specimen in compression, which includes the complicated coupling effects between the interaction and propagation of randomly distributed or other typical microcrack configurations step by step. The lengths, orientations and positions of microcracks are all taken as random variables. The crack interaction among many preexisting random microcracks is evaluated with the help of a simple interaction matrix (Yang and Liu, 1991). For the subcritically stable propagation of microcracks in mixed mode fracture, fairly known maximum hoop stress criterion is adopted to compute branching lengths and directions at each tip of the crack

  9. Status of the steam generator tube circumferential ODSCC degradation experienced at the Doel 4 plant

    International Nuclear Information System (INIS)

    Roussel, G.

    1997-01-01

    Since the 1991 outage, the Doel Unit 4 nuclear power plant is known to be affected by circumferential outside diameter intergranular stress corrosion cracking at the hot leg tube expansion transition. Extensive non destructive examination inspections have shown the number of tubes affected by this problem as well as the size of the cracks to have been increasing for the three cycles up to 1993. As a result of the high percentage of tubes found non acceptable for continued service after the 1993 in-service inspection, about 1,700 mechanical sleeves were installed in the steam generators. During the 1994 outage, all the tubes sleeved during the 1993 outage were considered as potentially cracked to some extent at the upper hydraulic transition and were therefore not acceptable for continued service. They were subsequently repaired by laser welding. Furthermore all the tubes not sleeved during the 1993 outage were considered as not acceptable for continued service and were repaired by installing laser welded sleeves. During the 1995 outage, some unexpected degradation phenomena were evidenced in the sleeved tubes. This paper summarizes the status of the circumferential ODSCC experienced in the SG tubes of the Doel 4 plant as well as the other connected degradation phenomena

  10. Eddy current technology for heat exchanger and steam generator tube inspection

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.; Lepine, B.; Lu, J.; Cassidy, R.; Carter, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2004-07-01

    A variety of degradation modes can affect the integrity of both heat exchanger (HX) and balance of plant tubing, resulting in expensive repairs, tube plugging or replacement of tube bundles. One key component for ensuring tube integrity is inspection and monitoring for detection and characterization of the degradation. In-service inspection of HX and balance of plant tubing is usually carried out using eddy current (EC) bobbin coils, which are adequate for the detection of volumetric degradations. However, detection and quantification of additional modes of degradation such as pitting, intergranular attack (IGA), axial cracking and circumferential cracking require specialized probes. The need for timely, reliable detection and characterization of these modes of degradation is especially critical in Nuclear Generating Stations. Transmit-receive single-pass array probes, developed by AECL, offer high defect detectability in conjunction with fast and reliable inspection capabilities. They have strong directional properties, permitting probe optimization for circumferential or axial crack detection. Compared to impedance probes, they offer improved performance in the presence of variable lift-off. This EC technology can help resolve critical detection issues at susceptible areas, such as the rolled-joint transitions at the tubesheet, U-bends and tube-support intersections. This paper provides an overview of the operating principles and the capabilities of advanced ET inspection technology available for HX tube inspection. Examples of recent application of this technology in Nuclear Generating Stations (NGSs) are discussed. (author)

  11. Eddy current technology for heat exchanger and steam generator tube inspection

    International Nuclear Information System (INIS)

    Obrutsky, L.; Lepine, B.; Lu, J.; Cassidy, R.; Carter, J.

    2004-01-01

    A variety of degradation modes can affect the integrity of both heat exchanger (HX) and balance of plant tubing, resulting in expensive repairs, tube plugging or replacement of tube bundles. One key component for ensuring tube integrity is inspection and monitoring for detection and characterization of the degradation. In-service inspection of HX and balance of plant tubing is usually carried out using eddy current (EC) bobbin coils, which are adequate for the detection of volumetric degradations. However, detection and quantification of additional modes of degradation such as pitting, intergranular attack (IGA), axial cracking and circumferential cracking require specialized probes. The need for timely, reliable detection and characterization of these modes of degradation is especially critical in Nuclear Generating Stations. Transmit-receive single-pass array probes, developed by AECL, offer high defect detectability in conjunction with fast and reliable inspection capabilities. They have strong directional properties, permitting probe optimization for circumferential or axial crack detection. Compared to impedance probes, they offer improved performance in the presence of variable lift-off. This EC technology can help resolve critical detection issues at susceptible areas, such as the rolled-joint transitions at the tubesheet, U-bends and tube-support intersections. This paper provides an overview of the operating principles and the capabilities of advanced ET inspection technology available for HX tube inspection. Examples of recent application of this technology in Nuclear Generating Stations (NGSs) are discussed. (author)

  12. Lunar Lava Tube Sensing

    Science.gov (United States)

    York, Cheryl Lynn; Walden, Bryce; Billings, Thomas L.; Reeder, P. Douglas

    1992-01-01

    Large (greater than 300 m diameter) lava tube caverns appear to exist on the Moon and could provide substantial safety and cost benefits for lunar bases. Over 40 m of basalt and regolith constitute the lava tube roof and would protect both construction and operations. Constant temperatures of -20 C reduce thermal stress on structures and machines. Base designs need not incorporate heavy shielding, so lightweight materials can be used and construction can be expedited. Identification and characterization of lava tube caverns can be incorporated into current precursor lunar mission plans. Some searches can even be done from Earth. Specific recommendations for lunar lava tube search and exploration are (1) an Earth-based radar interferometer, (2) an Earth-penetrating radar (EPR) orbiter, (3) kinetic penetrators for lunar lava tube confirmation, (4) a 'Moon Bat' hovering rocket vehicle, and (5) the use of other proposed landers and orbiters to help find lunar lava tubes.

  13. Categorising YouTube

    DEFF Research Database (Denmark)

    Simonsen, Thomas Mosebo

    2011-01-01

    This article provides a genre analytical approach to creating a typology of the User Generated Content (UGC) of YouTube. The article investigates the construction of navigation processes on the YouTube website. It suggests a pragmatic genre approach that is expanded through a focus on YouTube......’s technological affordances. Through an analysis of the different pragmatic contexts of YouTube, it is argued that a taxonomic understanding of YouTube must be analysed in regards to the vacillation of a user-driven bottom-up folksonomy and a hierarchical browsing system that emphasises a culture of competition...... and which favours the already popular content of YouTube. With this taxonomic approach, the UGC videos are registered and analysed in terms of empirically based observations. The article identifies various UGC categories and their principal characteristics. Furthermore, general tendencies of the UGC within...

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  15. Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum, E-mail: bahn@anl.gov [Argonne National Laboratory, Argonne, IL 60439 (United States); Majumdar, Saurin [Argonne National Laboratory, Argonne, IL 60439 (United States); Harris, Charles [United States Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

    2011-10-15

    Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low {Delta}K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed. - Highlights: > Leak rate of flawed Alloy 600 tubing increased at constant pressure condition. > Fractography revealed two cases: ductile tearing and crystallographic facets. > Crystallographic facets are typical features of fatigue crack growth at low {Delta}K. > Fatigue source could be water jet-induced vibration and/or high-pressure pump pulsation.

  16. Divergent axial morphogenesis and early shh expression in vertebrate prospective floor plate

    Directory of Open Access Journals (Sweden)

    Stanislav Kremnyov

    2018-01-01

    Full Text Available Abstract Background The notochord has organizer properties and is required for floor plate induction and dorsoventral patterning of the neural tube. This activity has been attributed to sonic hedgehog (shh signaling, which originates in the notochord, forms a gradient, and autoinduces shh expression in the floor plate. However, reported data are inconsistent and the spatiotemporal development of the relevant shh expression domains has not been studied in detail. We therefore studied the expression dynamics of shh in rabbit, chicken and Xenopus laevis embryos (as well as indian hedgehog and desert hedgehog as possible alternative functional candidates in the chicken. Results Our analysis reveals a markedly divergent pattern within these vertebrates: whereas in the rabbit shh is first expressed in the notochord and its floor plate domain is then induced during subsequent somitogenesis stages, in the chick embryo shh is expressed in the prospective neuroectoderm prior to the notochord formation and, interestingly, prior to mesoderm immigration. Neither indian hedgehog nor desert hedgehog are expressed in these midline structures although mRNA of both genes was detected in other structures of the early chick embryo. In X. laevis, shh is expressed at the beginning of gastrulation in a distinct area dorsal to the dorsal blastopore lip and adjacent to the prospective neuroectoderm, whereas the floor plate expresses shh at the end of gastrulation. Conclusions While shh expression patterns in rabbit and X. laevis embryos are roughly compatible with the classical view of “ventral to dorsal induction” of the floor plate, the early shh expression in the chick floor plate challenges this model. Intriguingly, this alternative sequence of domain induction is related to the asymmetrical morphogenesis of the primitive node and other axial organs in the chick. Our results indicate that the floor plate in X. laevis and chick embryos may be initially

  17. Rectangular drift tube characteristics

    International Nuclear Information System (INIS)

    Denisov, D.S.; Musienko, Yu.V.

    1985-01-01

    Results on the study of the characteristics of a 50 x 100 mm aluminium drift tube are presented. The tube was filled with argon-methane and argon-isobutane mixtures. With 16 per cent methane concentration the largest deviation from a linear relation between the drift time and the drift path over 50 mm is less than 2 mm. The tube filled with argon-isobutane mixture is capable of operating in a limited streamer mode

  18. 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...

  19. Pressure tube reactor

    International Nuclear Information System (INIS)

    Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

    1993-01-01

    In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

  20. Heated Tube Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Heated Tube Facility at NASA GRC investigates cooling issues by simulating conditions characteristic of rocket engine thrust chambers and high speed airbreathing...