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

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.)

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

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

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)

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

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)

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.

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

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

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.

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.

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.

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

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)

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

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

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).

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

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.

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

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

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

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)

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

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)

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)

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

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

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

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...

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

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)

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)

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)

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.

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

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

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)

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.

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)

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

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.)

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

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

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.)

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

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

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.

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)

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

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.

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)

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)

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.)

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

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

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

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

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

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)

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)

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)

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)

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

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.

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

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

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)

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.)

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)

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

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

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

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)

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)

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

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

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.

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

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)

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.

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)

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.

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)

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

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

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

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

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

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.

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.

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)

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)

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

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

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

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

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

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.

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.

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

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.

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)

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)

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

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)

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

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.

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...

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

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.

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.

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

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)

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)

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.

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.

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

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.

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

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

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)

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

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)

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)

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)

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.

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

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

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.

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.

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.

Materials for the pulp and paper industry. Section 1: Development of materials for black liquor recovery boilers

Energy Technology Data Exchange (ETDEWEB)

Keiser, J.R.; Hubbard, C.R.; Payzant, E.A. [Oak Ridge National Lab., TN (United States)] [and others

1997-04-01

Black liquor recovery boilers are essential components of kraft pulp and paper mills because they are a critical element of the system used to recover the pulping chemicals required in the kraft pulping process. In addition, the steam produced in these boilers is used to generate a significant portion of the electrical power used in the mill. Recovery boilers require the largest capital investment of any individual component of a paper mill, and these boilers are a major source of material problems in a mill. The walls and floors of these boilers are constructed of tube panels that circulate high pressure water. Molten salts (smelt) accumulate on the floor of recovery boilers, and leakage of water into the boiler can result in a violent explosion when the leaked water instantly vaporizes upon contacting the molten smelt. Because corrosion of the conventionally-used carbon steel tubing was found to be excessive in the lower section of recovery boilers, use of stainless steel/carbon steel co-extruded tubing was adopted for boiler walls to lessen corrosion and reduce the likelihood of smelt/water explosions. Eventually, this co-extruded or composite (as it is known in the industry) tubing was selected for use as a portion or all of the floor of recovery boilers, particularly those operating at pressures > 6.2 MPa (900 psi), because of the corrosion problems encountered in carbon steel floor tubes. Since neither the cause of the cracking nor an effective solution has been identified, this program was established to develop a thorough understanding of the degradation that occurs in the composite tubing used for walls and floors. This is being accomplished through a program that includes collection and review of technical reports, examination of unexposed and cracked tubes from boiler floors, computer modeling to predict residual stresses under operating conditions, and operation of laboratory tests to study corrosion, stress corrosion cracking, and thermal fatigue.

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)

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)

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 ...

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

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.

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

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.

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

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)

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

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

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

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

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.

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.

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)

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

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

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)

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.

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

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

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

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

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

International Nuclear Information System (INIS)

Griesbach, T.; Cipolla, R.

1985-01-01

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

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

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

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.)

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.

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.)

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)

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.

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)

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.

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...

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

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

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)

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.

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

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.

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.)

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

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

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)

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

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

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.

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

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

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.

Steam generator tube integrity program. Semiannual report, August 1995--March 1996

International Nuclear Information System (INIS)

Diercks, D.R.; Bakhtiari, S.; Chopra, O.K.

1997-04-01

This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of that program in August 1995 through March 1996. The program is divided into five tasks, namely (1) Assessment of Inspection Reliability, (2) Research on ISI (in-service-inspection) Technology, (3) Research on Degradation Modes and Integrity, (4) Development of Methodology and Technical Requirements for Current and Emerging Regulatory Issues, and (5) Program Management. Under Task 1, progress is reported on the preparation of and evaluation of nondestructive evaluation (NDE) techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate burst pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Under Task 2, results are reported on closed-form solutions and finite element electromagnetic modeling of EC probe response for various probe designs and flaw characteristics. Under Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe accident conditions. In addition, crack behavior and stability are being modeled to provide guidance on test facility design, to develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and to predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the cracking and failure of tubes that have been repaired by sleeving, and with a review of literature on this subject

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

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

Fast fracture of a zirconium alloy pressure tube: cause and implications

International Nuclear Information System (INIS)

Price, E.G.; Cheadle, B.A.

1985-12-01

The cause of the unstable fracture of a Zircaloy-2 pressure tube in the core of a CANDU reactor is reviewed. Failure was associated with the presence of brittle zones of zirconium hydride which developed as a result of thermal gradient induced hydrogen diffusion. Unstable fracture occurred when the partial thickness crack reached an unstable length and the crack ran 2 meters along the tube and terminated by circumferential tearing. The partial thickness defect initiated and propagated to an unstable length by delayed hydride cracking is high compared to fatigue progression and increases exponentially with temperature. Delayed hydride cracking can be prevented by reducing residual stresses to a minimum and by high standards of non-destructive testing that ensures freedom from unacceptable defects. Future prevention of fast fracture is based upon the inspection of a limited number of fuel channels for the presence of defects and for conditions which can cause hydride build-up together with the periodic removal of Zr-2.5wt% Nb tubes to monitor their condition

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)

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

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)

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.

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

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

International Nuclear Information System (INIS)

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

2007-08-01

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

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

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)

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)

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.

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.

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)

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)

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

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)

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)

Inhibition of stress corrosion cracking of alloy 600 in 10% NaOH solutions with and with lead oxide at 315 C

International Nuclear Information System (INIS)

Hur, D.H.; Kim, J.S.; Baek, J.S.; Kim, J.G.

2002-01-01

Alloy 600 steam generator tube materials have experienced various degradations by corrosion such as stress corrosion cracking (SCC) on the inner and outer diameter surface of tube, intergranular attack and pitting, and by mechanical damage such as fretting-wear and fatigue. These tube degradations not only increase the costs for tube inspection, maintenance and repair but also reduce the operation safety and the efficiency of plants. Therefore, the methodologies have been extensively developed to mitigate them. The addition of inhibitors to the coolant is a feasible method to mitigate tube degradations in operating plants. In this paper, a new inhibitor is proposed to mitigate the secondary side stress corrosion cracking of alloy 600 tubes. The effect of inhibitors on the electrochemical behavior and the stress corrosion cracking resistance of alloy 600 was evaluated in 10% sodium hydroxide solution with and without lead oxide at 315 C. The specimens of a C-ring type for stress corrosion cracking test were polarized at 150 mV above the corrosion potential for 120 hours without and with inhibitors such as titanium oxide, titanium boride, cerium boride. The chemical compositions of the films formed on the crack tip in the C-ring specimens were analyzed using a scanning Auger electron spectroscopy. The cerium boride, the most effective inhibitors, was observed to decrease the crack propagation rate more than a factor of three compared with that obtained in pure 10% NaOH solution. Furthermore, no SCC was observed in lead contaminated 10% NaOH solution by the addition of the cerium boride. (authors)

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.

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)

A REVIEW ON THE ODSCC OF STEAM GENERATOR TUBES IN KOREAN NPPS

Directory of Open Access Journals (Sweden)

HANSUB CHUNG

2013-08-01

Full Text Available The ODSCC detected in the TSP position of Ulchin 3&4 SGs are typical ODSCC of Alloy 600MA tubes. The causative chemical environment is formed by concentration of impurities inside the occluded region formed by the tube surface, egg crate strips, and sludge deposit there. Most cracks are detected at or near the line contacts between the tube surface and the egg crate strips. The region of dense crack population, as defined as between 4th and 9th TSPs, and near the center of hot leg hemisphere plane, coincided well with the region of preferential sludge deposition as defined by thermal hydraulics calculation using SGAP computer code. The cracks developed homogeneously in a wide range of SGs, so that the number of cracks detected each outage increased very rapidly since the first detection in the 8th refueling outage. The root cause assessment focused on investigation of the difference in microstructure and manufacturing residual stress in order to reveal the cause of different susceptibilities to ODSCC among identical six units. The manufacturing residual stress as measured by XRD on OD surface and by split tube method indicated that the high residual stress of Alloy 600MA tube played a critical role in developing ODSCC. The level of residual stress showed substantial variations among the six units depending on details of straightening and OD grinding processes. Youngwang 3&4 tubes are less susceptible to ODSCC than U3 and U4 tubes because semi-continuous coarse chromium carbides are formed along the grain boundary of Y3&4 tubes, while there are finer less continuous chromium carbides in U3 and U4. The different carbide morphology is caused by the difference in cooling rate after mill anneal. There is a possibility that high chromium content in the Y3&4 tubes, still within the allowable range of Alloy 600, has made some contribution to the improved resistance to ODSCC. It is anticipated that ODSCC in Y5&6 SGs will be retarded more considerably

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)

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

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.

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)

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

Stress field determination in an alloy 600 stress corrosion crack specimen

International Nuclear Information System (INIS)

Rassineux, B.; Labbe, T.

1995-05-01

In the context of EDF studies on stress corrosion cracking rates in the Alloy 600 steam generators tubes, we studied the influence of strain hardened surface layers on the different stages of cracking for a tensile smooth specimen (TLT). The stress field was notably assessed to try and explain the slow/rapid-propagation change observed beyond the strain hardened layers. The main difficulty is to simulate in a finite element model the inner and outer surfaces of these strain hardened layers, produced by the final manufacturing stages of SG tubes which have not been heat treated. In the model, the strain hardening is introduced by simulating a multi-layer material. Residual stresses are simulated by an equivalent fictitious thermomechanical calculation, realigned with respect to X-ray measurements. The strain hardening introduction method was validated by an analytical calculation giving identical results. Stress field evolution induced by specimen tensile loading were studied using an elastoplastic 2D finite element calculations performed with the Aster Code. The stress profile obtained after load at 660 MPa shows no stress discontinuity at the boundary between the strain hardened layer and the rest of the tube. So we propose that a complementary calculation be performed, taking into account the multi-cracked state of the strain hardened zones by means of a damage variable. In fact, this state could induce stress redistribution in the un-cracked area, which would perhaps provide an explanation of the crack-ground rate change beyond the strain hardened zone. The calculations also evidence the harmful effects of plastic strains on a strain hardened layer due to the initial state of the tube (not heat-treated), to grit blasting or to shot peening. The initial compressive stress condition of this surface layer becomes, after plastic strain, a tensile stress condition. These results are confirmed by laboratory test. (author). 10 refs., 18 figs., 9 tabs., 2 appends

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).

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

Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

Energy Technology Data Exchange (ETDEWEB)

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

1997-10-01

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

Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

International Nuclear Information System (INIS)

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

1997-10-01

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

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

Operating envelope to minimize probability of fractures in Zircaloy-2 pressure tubes

International Nuclear Information System (INIS)

Azer, N.; Wong, H.

1994-01-01

The failure mode of primary concern with Candu pressure tubes is fast fracture of a through-wall axial crack, resulting from delayed hydride crack growth. The application of operating envelopes is demonstrated to minimize the probability of fracture in Zircaloy-2 pressure tubes based on Zr-2.5%Nb pressure tube experience. The technical basis for the development of the operating envelopes is also summarized. The operating envelope represents an area on the pressure versus temperature diagram within which the reactor may be operated without undue concern for pressure tube fracture. The envelopes presented address both normal operating conditions and the condition where a pressure tube leak has been detected. The examples in this paper are prepared to illustrate the methodology, and are not intended to be directly applicable to the operation of any specific reactor. The application of operating envelopes to minimized the probability of fracture in 80 mm diameter Zircaloy-2 pressure tubes has been discussed. Both normal operating and leaking pressure tube conditions have been considered. 3 refs., 4 figs

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

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

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

International Nuclear Information System (INIS)

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

1992-02-01

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

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

Settling time of dental x-ray tube head after positioning

International Nuclear Information System (INIS)

Yun, Suk Ja; Kang, Byung Cheol; Wang, Se Myung; Koh, Chang Sung

2002-01-01

The aim of this study was to introduce a method of obtaining the oscillation graphs of the dental x-ray tube heads relative to time using an accelerometer. An Accelerometer, Piezotron type 8704B25 (Kistler Instrument Co., Amherst, NY, USA) was utilized to measure the horizontal oscillation of the x-ray tube head immediately after positioning the tube head for an intraoral radiograph. The signal from the sensor was transferred to a dynamic signal analyzer, which displayed the magnitude of the acceleration on the Y-axis and time lapse on the X-axis. The horizontal oscillation of the tube head was measured relative to time, and the settling time was also determined on the basis of the acceleration graphs for 6 wall type, 5 floor-fixed type, and 4 mobile type dental x-ray machines. The oscillation graphs showed that tube head movement decreased rapidly over time. The settling time varied with x-ray machine types. Wall-type x-ray machines had a settling time of up to 6 seconds, 5 seconds for fixed floor-types, and 11 seconds for the mobile-types. Using an accelerometer, we obtained the oscillation graphs of the dental x-ray tube head relative to time. The oscillation graph with time can guide the operator to decide upon the optimum exposure moment after xray tube head positioning for better radiographic resolution.

Settling time of dental x-ray tube head after positioning

Energy Technology Data Exchange (ETDEWEB)

Yun, Suk Ja; Kang, Byung Cheol [Department of Oral and Maxillofacial Radiology, Chonnam National University, Gwangju (Korea, Republic of); Wang, Se Myung; Koh, Chang Sung [Department of Mechatronics, Kwangju Institute of Science and Technology, Gwangju (Korea, Republic of)

2002-09-15

The aim of this study was to introduce a method of obtaining the oscillation graphs of the dental x-ray tube heads relative to time using an accelerometer. An Accelerometer, Piezotron type 8704B25 (Kistler Instrument Co., Amherst, NY, USA) was utilized to measure the horizontal oscillation of the x-ray tube head immediately after positioning the tube head for an intraoral radiograph. The signal from the sensor was transferred to a dynamic signal analyzer, which displayed the magnitude of the acceleration on the Y-axis and time lapse on the X-axis. The horizontal oscillation of the tube head was measured relative to time, and the settling time was also determined on the basis of the acceleration graphs for 6 wall type, 5 floor-fixed type, and 4 mobile type dental x-ray machines. The oscillation graphs showed that tube head movement decreased rapidly over time. The settling time varied with x-ray machine types. Wall-type x-ray machines had a settling time of up to 6 seconds, 5 seconds for fixed floor-types, and 11 seconds for the mobile-types. Using an accelerometer, we obtained the oscillation graphs of the dental x-ray tube head relative to time. The oscillation graph with time can guide the operator to decide upon the optimum exposure moment after xray tube head positioning for better radiographic resolution.

Possible first occurrence of external corrosion on alloy 600TT tubes in France

International Nuclear Information System (INIS)

Boccanfuso, M.; Thebault, Y.; Massini, B.; Bigne, L.

2015-01-01

During the last decade, in different countries, several occurrences of external corrosion have been identified on steam generator (SG) tube bundles equipped with thermally treated 600 alloy. In France, this feedback leads EDF to enhance the SG inspection program. Nevertheless, until now, no damage of this type was reported. Recently, during in-service inspection at the Cattenom plant on a SG equipped with alloy 600TT tubes, Eddy current tests have highlighted a signal that could be related to external corrosion. The tube was removed and sent to the EDF hot laboratory for destructive examinations. Various exams were performed at different scales to characterize the causes of this NDT signal, the material properties and the residual stresses. The assessments carried out on the tube conclude that the source of the damage is external intergranular stress corrosion cracking, also called ODSCC (Outside Diameter Stress Corrosion Cracking) making it the first occurrence on the tube bundles made of alloy 600TT in the French fleet. This first case of 600 TT ODSCC in France is an unexpected and particular one, because of its altitude in the full mechanical rolling area. This is reinforced by the low number of occurrences noted to date (only one after nearly 30 years of operation of alloy 600TT tube bundles). International (Biblis) OPEX had identified recent IGSCC with cracks initiated and propagated in the tubesheet. For this case, the scenario considered requires highly restrictive conditions (tube in the sludge zone and on the periphery of the tube bundle, including the tube lane) and may explain the singular nature of the Cattenom tube

Steam generator tube integrity program: Annual report, August 1995--September 1996. Volume 2

International Nuclear Information System (INIS)

Diercks, D.R.; Bakhtiari, S.; Kasza, K.E.; Kupperman, D.S.; Majumdar, S.; Park, J.Y.; Shack, W.J.

1998-02-01

This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of the program in August 1995 through September 1996. The program is divided into five tasks: (1) assessment of inspection reliability, (2) research on ISI (inservice-inspection) technology, (3) research on degradation modes and integrity, (4) tube removals from steam generators, and (5) program management. Under Task 1, progress is reported on the preparation of facilities and evaluation of nondestructive evaluation techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate failure pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Results are reported in Task 2 on closed-form solutions and finite-element electromagnetic modeling of EC probe responses for various probe designs and flaw characteristics. In Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe-accident conditions. Crack behavior and stability are also being modeled to provide guidance for test facility design, develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the acquisition of tubes and tube sections from retired steam generators for use in the other research tasks. Progress on the acquisition of tubes from the Salem and McGuire 1 nuclear plants is reported

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)

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

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

Axial strain localization of CuCrZr tubes during manufacturing of ITER-like mono-block W/Cu components using HIP

International Nuclear Information System (INIS)

Zhao, S.X.; Peng, L.J.; Li, Q.; Wang, W.J.; Wei, R.; Qin, S.G.; Shi, Y.L.; Chang, S.P.; Xu, Y.; Liu, G.H.; Wang, T.J.; Luo, G.-N.

2014-01-01

Highlights: • Axial cracking and denting of CuCrZr tubes were observed. • Annealing the as-received tubes can alleviate cracking. • Denting results in the formation bonding flaws at the Cu/CuCrZr interfaces. - Abstract: Two forms of axial strain localization of CuCrZr tubes, i.e., cracking and denting, were observed during the manufacturing of ITER-like mono-block W/Cu components for EAST employing hot isostatic pressing (HIP). Microscopic investigations indicate that the occurrence of axial strain localization correlates to the heavily deformed Cu grains and elongated Cr-rich precipitates as well as highly anisotropic microstructures, which impair the circumferential ductility. Annealing the as-received tubes at 600 °C alleviates cracking due to partial recrystallization of Cu grains. However, the annealed tubes are still sensitive to wall thinning (caused by non-uniform polishing or tube bending), which results in denting. Denting may cause bonding flaws at CuCrZr/Cu interfaces and the underlying mechanisms are discussed. To some extent, denting seems do not affect the high heat flux performance of components. In this paper, we demonstrate that testing only the axial mechanical properties is not enough for manufacturers who use HIP or hot radial pressing technologies, especially for those anisotropic tubes

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.)

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

Effects of metallurgical factors on stress corrosion cracking of Ni-base alloys in high temperature water

International Nuclear Information System (INIS)

Yonezawa, T.; Sasaguri, N.; Onimura, K.

1988-01-01

Nickel-base Alloy 600 is the principal material used for the steam generator tubes of PWRs. Generally, this alloy has been proven to be satisfactory for this application, however when it is subjected to extremely high stress level in PWR primary water, it may suffer from stress corrosion cracking. The authors have systematically studied the effects of test temperature and such metallurgical factors as cold working, chemical composition and heat treatment on the stress corrosion cracking of Alloy 600 in high temperature water, and also on that of Alloy 690 which is a promising material for the tubes and may provide improved crrosion resistance for steam generators. The test materials, the stress corrosion cracking test and the test results are reported. When the test temperature was raise, the stress corrosion cracking of the nickel-base alloys was accelerated. The time of stress corrosion cracking occurrence decreased with increasing applied stress, and it occurred at the stress level higher than the 0.2 % offset proof stress of Alloy 600. In Alloy 690, stress corrosion cracking was not observed at such stress level. Cold worked Alloy 600 showed higher resistance to stress corrosion cracking than the annealed alloy. (Kako, I.)

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

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)

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

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.

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....

Evaluation of ECT reliability for axial ODSCC in steam generator tubes

International Nuclear Information System (INIS)

Lee, Jae Bong; Park, Jai Hak; Kim, Hong Deok; Chung, Han Sub

2010-01-01

The integrity of steam generator tubes is usually evaluated based on eddy current test (ECT) results. Because detection capacity of the ECT is not perfect, all of the physical flaws, which actually exist in steam generator tubes, cannot be detected by ECT inspection. Therefore it is very important to analyze ECT reliability in the integrity assessment of steam generators. The reliability of an ECT inspection system is divided into reliability of inspection technique and reliability of quality of analyst. And the reliability of ECT results is also divided into reliability of size and reliability of detection. The reliability of ECT sizing is often characterized as a linear regression model relating true flaw size data to measured flaw size data. The reliability of detection is characterized in terms of probability of detection (POD), which is expressed as a function of flaw size. In this paper the reliability of an ECT inspection system is analyzed quantitatively. POD of the ECT inspection system for axial outside diameter stress corrosion cracks (ODSCC) in steam generator tubes is evaluated. Using a log-logistic regression model, POD is evaluated from hit (detection) and miss (no detection) binary data obtained from destructive and non-destructive inspections of cracked tubes. Crack length and crack depth are considered as variables in multivariate log-logistic regression and their effects on detection capacity are assessed using two-dimensional POD (2-D POD) surface. The reliability of detection is also analyzed using POD for inspection technique (POD T ) and POD for analyst (POD A ).

TEM characterisation of stress corrosion cracks in nickel based alloys: effect of chromium content and chemistry of environment

International Nuclear Information System (INIS)

Delabrouille, F.

2004-11-01

Stress corrosion cracking (SCC) is a damaging mode of alloys used in pressurized water reactors, particularly of nickel based alloys constituting the vapour generator tubes. Cracks appear on both primary and secondary sides of the tubes, and more frequently in locations where the environment is not well defined. SCC sensitivity of nickel based alloys depends of their chromium content, which lead to the replacement of alloy 600 (15 % Cr) by alloy 690 (30 % Cr) but this phenomenon is not yet very well understood. The goal of this thesis is two fold: i) observe the effect of chromium content on corrosion and ii) characterize the effect of environment on the damaging process of GV tubes. For this purpose, one industrial tube and several synthetic alloys - with controlled chromium content - have been studied. Various characterisation techniques were used to study the corrosion products on the surface and within the SCC cracks: SIMS; TEM - FEG: thin foil preparation, HAADF, EELS, EDX. The effect of chromium content and surface preparation on the generalised corrosion was evidenced for synthetic alloys. Moreover, we observed the penetration of oxygen along triple junctions of grain boundaries few micrometers under the free surface. SCC tests show the positive effect of chromium for contents varying from 5 to 30 % wt. Plastic deformation induces a modification of the structure, and thus of the protective character, of the internal chromium rich oxide layer. SCC cracks which developed in different chemical environments were characterised by TEM. The oxides which are formed within the cracks are different from what is observed on the free surface, which reveals a modification of medium and electrochemical conditions in the crack. Finally we were able to evidence some structural characteristics of the corrosion products (in the cracks and on the surface) which turn to be a signature of the chemical environment. (author)

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)

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)

Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

Science.gov (United States)

Huang, F. H.

1993-12-01

Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

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)

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.

Impact Fretting Wear Behavior of Alloy 690 Tubes in Dry and Deionized Water Conditions

Institute of Scientific and Technical Information of China (English)

Zhen-Bing Cai; Jin-Fang Peng; Hao Qian; Li-Chen Tang; Min-Hao Zhu

2017-01-01

The impact fretting wear has largely occurred at nuclear power device induced by the flow-induced vibration,and it will take potential hazards to the service of the equipment.However,the present study focuses on the tangential fretting wear of alloy 690 tubes.Research on impact fretting wear of alloy 690 tubes is limited and the related research is imminent.Therefore,impact fretting wear behavior of alloy 690 tubes against 304 stainless steels is investigated.Deionized water is used to simulate the flow environment of the equipment,and the dry environment is used for comparison.Varied analytical techniques are employed to characterize the wear and tribochemical behavior during impact fretting wear.Characterization results indicate that cracks occur at high impact load in both water and dry equipment;however,the water as a medium can significantly delay the cracking time.The crack propagation behavior shows a jagged shape in the water,but crack extended disorderly in dry equipment because the water changed the stress distribution and retarded the friction heat during the wear process.The SEM and XPS analysis shows that the main failure mechanisms of the tube under impact fretting are fatigue wear and friction oxidation.The effect of medium(water) on fretting wear is revealed,which plays a potential and promising role in the service of nuclear power device and other flow equipments.

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)

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

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

International Nuclear Information System (INIS)

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

1985-07-01

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

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.

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)

Metallographical procedure for detecting carburized layer in the 12Kh18N10T steel tubes

International Nuclear Information System (INIS)

Najmark, S.I.; Shmalenyuk, R.P.

1977-01-01

The various causes of the tendency to intercrystalline corrosion of stainless steel tubing were studied. 12Kh18N10T steel tube specimens were metallographically investigated to reveal the presence and depth of the carburized layer. Unlubricated specimens and specimens with the graphite-saltpeter lubrication applied on them were austenized in the temperature range from 950 to 1250 deg C and then exposed to a provoking anneal at 650 deg C for 24 h. All the lubricated specimens showed a carbide network along grain boundaries and intercrystalline cracks. The specimens treated without lubrication had no cracks. The method offered can differentiate poor tube cleaning and the undesirable effect of other processing steps on the tubes

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)

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

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.

Fracture inspection by BHTV logging through a vinyl chloride tube; Enbi pipe ga sonyusareta kosei deno BHTV ni yoru kiretsu chosa

Energy Technology Data Exchange (ETDEWEB)

Horikawa, H.; Okabe, T.; Takasugi, S. [GERD Geothermal Energy Research and Development Co. Ltd., Tokyo (Japan); Yamauchi, Y.; Imanishi, H. [Osaka Soil Test, Osaka (Japan)

1997-05-27

A borehole televiewer (BHTV) logging was conducted to investigate cracks in bored wells. The logging process shoots ultrasonic waves onto bore walls and utilizes reflection waves from the bore walls. If the bores are filled with muddy water, or the ultrasonic waves can pass through even if vinyl chloride tubes have been inserted after excavation, the state of the bore walls can be investigated in principle. Conventional optical scanners are, however, incapable of making this investigation. The BHTV logging can be used for identifying lithofacies from reflection intensities from bore walls (it depends on sound impedance of rocks), not to speak of finding cracks. As a result of making logging upon setting time windows from 105 to 145 {mu} sec in a bored well inserted with a vinyl chloride tube, cracks in the bore wall were identified clearly through the vinyl chloride tube. If the distance between the bore wall and the vinyl chloride tube becomes smaller, setting the time window becomes difficult, making reflection data acquisition more difficult. Impacts of decentering in the vinyl chloride tube on the data and the crack analysis result are small. The reflection intensity has decreased by about half as a result of having the circular vinyl chloride tube inserted. 5 refs., 9 figs., 1 tab.

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

Fatigue life prediction of autofrettage tubes using actual material behaviour

International Nuclear Information System (INIS)

Jahed, Hamid; Farshi, Behrooz; Hosseini, Mohammad

2006-01-01

There is a profound Bauschinger effect in the behaviour of high-strength steels used in autofrettaged tubes. This has led to development of methods capable of considering experimentally obtained (actual) material behaviour in residual stress calculations. The extension of these methods to life calculations is presented here. To estimate the life of autofrettaged tubes with a longitudinal surface crack emanating from the bore more accurately, instead of using idealized models, the experimental loading-unloading stress-strain behaviour is employed. The resulting stresses are then used to calculate stress intensity factors by the weight function method as input to fatigue life determination. Fatigue lives obtained using the actual material behaviour are then compared with the results of frequently used ideal models including those considering Bauschinger effect factors and strain hardening in unloading. Using standard fatigue crack growth relationships, life of the vessel is then calculated based on recommended initial and final crack length. It is shown that the life gain due to autofrettage above 70% overstrain is considerable

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.

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

Overview of magnetic bias X-probe qualification and inspection of PNGS Monel 400 steam generator tubing

International Nuclear Information System (INIS)

Lepine, B.A.; Van Langen, J.; Obrutsky, L.

2006-01-01

This paper presents an overview of the X-probe MB 350, the qualification for detection of open OD axial crack-like flaws, and a selection of inspection results from the subsequent field inspections performed with this probe during the 2003 and 2004 period at Pickering Nuclear Generating Station A and B. Examples of the field indications to be presented are axial cracking, OD pitting at top of tubesheet location (TTS), and flow assisted corrosion (top hats). During the 2003 in-service eddy current inspection results of Pickering Nuclear Generating Station A (PNGS-A) Unit 2, a 13 mm (0.5 inch) long axial indication was detected by the CTR1 bobbin and CTR2-C4 array probes in Tube R25-C52 of Steam Generator (SG) 11 in the hot leg sludge pile region. An experimental magnetic bias X-probe, especially designed by Zetec for inspection of Monel 400 tubing, was deployed and the indication was characterized as a potential outer diameter (OD) axially oriented crack. Posterior tube pulling and destructive examination confirmed the presence of an Environmentally Assisted Crack (EAC), approximately 80% deep and 13 mm long. Due to the significance of this discovery, Ontario Power Generation (OPG) requested AECL to initiate a program for qualification of the X-Probe MB 350 for the detection of OD axial cracks in medium to high magnetic permeability (μ r ) Monel 400 PNGS-A and B steam generator tubing at different locations. The X-probe MB 350 subsequently has been deployed as a primary inspection probe for crack detection for PNGS steam generators. (author)

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.

Stress corrosion cracking behaviour of Alloy 600 in high temperature water

International Nuclear Information System (INIS)

Webb, G.L.; Burke, M.G.

1995-01-01

The stress corrosion cracking (SCC) susceptibility of Alloy 600 in deaerated water at 360 deg. C, as measured with statistically-loaded U-bend specimens, is dependent upon microstructure and whether the material was cold-worked and annealed (CWA) or hot-worked and annealed (HWA). All cracking was intergranular, and materials lacking grain boundary carbides were most susceptible to SCC initiation. CWA tubing materials are more susceptible to SCC initiation than HWA ring-rolled forging materials with similar microstructures, as determined by light optical metallography (LOM). In CWA tubing materials one crack dominated and grew to a large size that was observable by visual inspection. HWA materials with a low hot-working finishing temperature (below 925 deg. C) and final anneals at temperatures ranging from 1010 deg. C to 1065 deg. C developed both large cracks, similar to those found in CWA materials, and also small intergranular microcracks, which are detectable only by destructive metallographic examination. HWA materials with a high hot-working finishing temperature (above 980 deg. C) and high-temperature final anneal (above 1040 deg. C), with grain boundaries that are fully decorated, developed only microcracks, which were observed in all specimens examined. These materials developed no large, visually detectable cracks, even after more than 300 weeks exposure. A low-temperature thermal treatment (610 deg. C for 7h), which reduced or eliminates SCC in Alloy 600, did not eliminate microcrack formation in the high temperature processed HWA materials. Detailed microstructural characterization using conventional metallographic and analytical electron microscopy (AEM) techniques was performed on selected materials to identify the factors responsible for the observed differences in cracking behaviour. 11 refs, 12 figs, 3 tabs

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.

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)

Effects of deposited pyrolytic carbon on some mechanical properties of zircaloy-4 tubes. Vol. 3

Energy Technology Data Exchange (ETDEWEB)

Shrkawy, S W; Abdel-razek, I D; El-Sayed, H A [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

Zircaloy cladding tubes are not compatible with the uranium fuel pellets as they suffer from failure due to pelletclad interaction (PCI). A carbon coating, as used in the canadian CANLUB fuel elements, is thought to improve the cladding performance with respect to the PCI problem. In this paper pyrolytic carbon coating was deposited on zircaloy-4 cladding tubes by the thermal cracking of commercial butant gas at the temperature range 250-450 degree C. In order to evaluate the effect of gaseous species on the mechanical properties of the tubes tensile and microhardness testing measurements were performed on samples prepared from the coated tubes. The fractured surface of the tensile zircaloy tubes and the deposited carbon coating, both, were examined by the SEM. The results of the tensile tests of zircaloy-4 tubes indicated that the coating process has insignificant effect on the ultimate strength of the tubes tested. The values of Vickers hardness numbers were not significantly changed across the tubes thickness. The microstructure of deposited carbon, due to the cracking process, was granular in all the temperature range (250-450 degree C) studied. 9 figs., 1 tab.

Trends of degradation in steam generator tubes of Krsko NPP before the last planned inspection

International Nuclear Information System (INIS)

Cizelj, L.; Dvorsek, T.; Androjna, F.

1998-01-01

Full-length inspection of all active tubes in both Krsko steam generators resulted in a huge amount of inspection records. A computerized database was developed by Reactor Engineering division to accelerate the management of about 200.000 records. The database was designed to support the development and decision related to the plugging criteria for damaged tubes and is utilized to gain as much experience concerning the degradation of SG tube as possible. In this paper, two prevailing group of data are statistically analyzed: the axial cracks in expansion transitions at the top of tube sheet (TTS) and Outside Diameter Stress Corrosion Cracking at tube support plates (TSP). Especially ODSCC caused a vast majority of repaired tubes (e.g., plugs and sleeves). The influence of plant startups involving oxidizing transient on the repair rates of tubes affected by ODSCC is analyzed in some detail. The results are promising and show excellent correlation in SG 2 and reasonable fit in SG 1. Predictions of maximum expected number of tubes repaired due to ODSCC at the last planned inspection is given as 67 in SG 1 and 400 in SG 2. (author)

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)

Heat Exchanger Tube Inspection of Nuclear Power Plants using IRIS Technique

International Nuclear Information System (INIS)

Yoon, Byung Sik; Yang, Seung Han; Song, Seok Yoon; Kim, Yong Sik; Lee, Hee Jong

2005-01-01

Inspection of heat exchange tubing include steam generator of nuclear power plant mostly performed with eddy current method. Recently, various inspection technique is available such as remote field eddy current, flux leakage and ultrasonic methods. Each of these techniques has its merits and limitations. Electromagnetic techniques are very useful to locate areas of concern but sizing is hard because of the difficult interpretation of an electric signature. On the other hand, ultrasonic methods are very accurate in measuring wall loss damage, and are reliable for detecting cracks. Additionally ultrasound methods is not affected by support plates or tube sheets and variation of electrical conductivity or permeability. Ultrasound data is also easier to analyze since the data displayed is generally the remaining wall thickness. It should be emphasized that ultrasound is an important tool for sizing defects in tubing. In addition, it can be used in situations where eddy current or remote field eddy current is not reliable, or as a flaw assessment tool to supplement the electromagnetic data. The need to develop specialized ultrasonic tools for tubing inspection was necessary considering the limitations of electromagnetic techniques to some common inspection problems. These problems the sizing of wall loss in carbon steel tubes near the tube sheet or support plate, sizing internal erosion damage, and crack detection. This paper will present an IRIS(Internal Rotating Inspection System) ultrasonic tube inspection technique for heat exchanger tubing in nuclear power plant and verify inspection reliability for artificial flaw embedded to condenser tube

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

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

Effect of boric acid on intergranular corrosion and on hideout return efficiency of sodium in the tube support plate crevices

International Nuclear Information System (INIS)

Paine, J.P.N.; Shoemaker, C.E.; Campan, J.L.; Brunet, J.P.; Schindler, P.; Stutzmann, A.

1995-01-01

Sodium hydroxide is one of the main causes of intergranular attack/stress corrosion cracking (IGA/SCC) of alloy 600 steam generator (S.G.) tubes. Boric acid appears to be one of the possible remedies for intergranular corrosion process inhibition. In order to obtain data on boric acid injection efficiency, an experimental program was performed on previously corroded tubes. To prevent premature tube wall cracking, samples were sleeved on alloy 690 tubes. The objective of the tests was to evaluate, on a statistically valid number of samples, the effectiveness of boric acid and tube sleeving as possible remedies for IGA/SCC extension. Another independent experimental program was initiated to determine the hideout return efficiency in the tube support plate (TSP) and tubesheet (TS) crevices after a significant duration (≤ 180 hours) of sodium hideout. The main objective of the first tests being a statistical evaluation of the efficiency of boric acid treatment, was not achieved. The tests did demonstrate that sleeving effectively reduces IGA/SCC growth. In an additional program, cracks were obtained on highly susceptible tubes when specimens were not sleeved. The companion tests performed in the same conditions but with an addition of boric acid did not show any IGA or cracks. These results seem to demonstrate the possible effect of boric acid in preventing the corrosion process. Results of the second tests did not demonstrate any difference in the amount of sodium piled up in the crevices before and after boric acid injection. They however showed an increase of the hideout return efficiency at the tube support plate level from 78 % without boric acid to 95 % when boric acid is present in the feed water

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.)

Study on the manufacturing process, causes of the pressure tube failure and methods for improving its performance

Energy Technology Data Exchange (ETDEWEB)

You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-07-01

Manufacturing processes of Zr-2.5Nb pressure tube used in CANDU reactor, effects of impurities on the properties of the pressure tube, experiences and causes of the pressure tube cracking accident and the development programs on the fuel channel at AECL have been described. Fabrication processes on the pressure tube have been explained in detail from the sponge production step to the final product. Test methods that are performed to verify the integrity of the final product have also been described. Most of the pressure tube rupture accidents were caused by DHC (Delayed Hydride Cracking). In cases of Pickering units 3 and 4 and Bruce unit 2, excessive residual stresses induced by improper rolled joint process had played a role to cause DHC. In Pickering unit 2, cracks formed by contact between pressure and calandria tubes due to the movement of garter spring were direct cause of failure. After the accidents, a lot of R and D programs on each component of the fuel channel have been carried out. The study on the improvement of manufacturing processes such as increasing cold working rate, performing the intermediate and final annealing and adding the third element like Fe, V, Cr for enhancing the pressure tube performance are on progress. To suppress hydrogen uptake into the pressure tube, the methods such as zirconia coating on the pressure tube, Cr-plating on the end fitting and placing the yttrium getter on the pressure tube are considered. Experiments on each test specimen are currently under way. Owing to such an effort, more advanced fuel channel can be installed in the next CANDU reactor. 6 tabs., 20 figs., 20 refs. (Author).

Influence of hydrogen content on fracture toughness of CWSR Zr-2.5Nb pressure tube alloy

Science.gov (United States)

Singh, R. N.; Bind, A. K.; Srinivasan, N. S.; Ståhle, P.

2013-01-01

In this work, influence of hydrogen and temperature on the fracture toughness parameters of unirradiated, cold worked and stress relieved (CWSR) Zr-2.5Nb pressure tube alloys used in Indian Pressurized Heavy Water Reactor is reported. The fracture toughness tests were carried out using 17 mm width curved compact tension specimens machined from gaseously hydrogen charged tube-sections. Metallography of the samples revealed that hydrides were predominantly oriented along axial-circumferential plane of the tube. Fracture toughness tests were carried out in the temperature range of 30-300 °C as per ASTM standard E-1820-06, with the crack length measured using direct current potential drop (DCPD) technique. The fracture toughness parameters (JQ, JMax and dJ/da), were determined. The critical crack length (CCL) for catastrophic failure was determined using a numerical method. It was observed that for a given test temperature, the fracture toughness parameters representing crack initiation (JQ) and crack propagation (JMax, and dJ/da) is practically unaffected by hydrogen content. Also, for given hydrogen content, all the aforementioned fracture toughness parameters increased with temperature to a saturation value.

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.)

Eddy current magnetic bias x-probe qualification and inspection of steam generator Monel 400 tubing in Pickering Nuclear Generating Station

International Nuclear Information System (INIS)

Lepine, B.A.; Van Langen, J.; Obrutsky, L.

2006-01-01

This paper presents an overview of the x-probe MB 350 eddy current inspection array probe, for detection of open OD axial crack-like flaws in Monel 400 tubes at Pickering Nuclear Generating Station. This report contains a selection of inspection results from the field inspections performed with this probe during the 2003 and 2004 period at Pickering Nuclear Generating Station A and B. During the 2003 in-service eddy current inspection results of Pickering Nuclear Generating Station A (PNGS-A) Unit 2, a 13 mm (0.5 inch) long axial indication was detected by the CTR1 bobbin and CTR2-C4 array probes in Tube R25-C52 of Steam Generator (SG) 11 in the hot leg sludge pile region. An experimental magnetic bias X-probe, specially designed by Zetec for inspection of Monel 400 tubing, was deployed and the indication was characterized as a potential out diameter (OD) axially oriented crack. Post-inspection tube pulling and destructive examination confirmed the presence of an Environmentally Assisted Crack (EAC), approximately 80% deep and 13mm long. Due to the significance of this discovery, Ontario Power Generation (OPG) requested AECL to initiate a program for qualification of the X-probe MB 350 for the detection of OD axial cracks in medium to high magnetic permeability μ r Monel 400 PNGS-A and B steam generator tubing at different locations. The X-probe MB 350 subsequently has been deployed as a primary inspection probe for crack detection for PNGS steam generators. (author)

Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

Energy Technology Data Exchange (ETDEWEB)

Groenwall, B; Ljungberg, L; Huebner, W; Stuart, W

1966-08-15

Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 {mu}g/cm{sup 2}). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in

Intercrystalline Stress Corrosion of Inconel 600 Inspection Tubes in the Aagesta Reactor

International Nuclear Information System (INIS)

Groenwall, B.; Ljungberg, L.; Huebner, W.; Stuart, W.

1966-08-01

Intercrystalline stress corrosion cracking has occurred in the Aagesta reactor in three so-called inspection tubes made of Inconel 600. The tubes had been exposed to 217 deg C light water, containing 1-4 ppm LiOH (later KOH) but only small amounts of oxygen, chloride and other impurities. Some of the circumferential cracks developed in or at crevices on the outside surface. At these positions constituents dissolved in the water may have concentrated. The crevices are likely to have contained a gas phase, mainly nitrogen. Local boiling in the crevices may also have occurred. Some few cracks were also found outside the crevice region. Irradiation effects can be neglected. No surface contamination could be detected except for a very minor fluoride content (1 μg/cm 2 ). The failed tubes had been subjected to high stresses, partly remaining from milling, partly induced by welding operations. The possibility that stresses slightly above the 0.2 per cent offset yield strength have occurred at the operating temperature cannot be excluded. The cracked tube material contained a large amount of carbide particles and other precipitates, both at grain boundaries and in the interior of grains. The particles appeared as stringers in circumferential zones. Zones depleted in precipitates were found along grain boundaries. The failed tube turned out to have an unusually high mechanical strength, likely due to a combination of some kind of ageing process and cold work (1.0 - 1.3 per cent plastic strain). Laboratory exposures of stressed surplus material in high purity water and in 1 M LiOH at 220 deg C showed some pitting but no cracking after 6800 h and 5900 h respectively. Though the encountered failures may have developed because of influence of some few or several of the above-mentioned detrimental factors, the actual cause cannot be stated with certainty. In the literature information is given concerning intercrystalline stress corrosion cracking of Inconel 600 both in caustic

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

The impact of a chest tube management protocol on the outcome of trauma patients with tube thoracostomy.

Science.gov (United States)

de Abreu, Emanuelle Maria Sávio; Machado, Carla Jorge; Pastore Neto, Mario; de Rezende Neto, João Baptista; Sanches, Marcelo Dias

2015-01-01

to investigate the effect of standardized interventions in the management of tube thoracostomy patients and to assess the independent effect of each intervention. A chest tube management protocol was assessed in a retrospective cohort study. The tube thoracostomy protocol (TTP) was implemented in August 2012, and consisted of: antimicrobial prophylaxis, chest tube insertion in the operating room (OR), admission post chest tube thoracostomy (CTT) in a hospital floor separate from the emergency department (ED), and daily respiratory therapy (RT) sessions post-CTT. The inclusion criteria were, hemodynamic stability, patients between the ages of 15 and 59 years, and injury severity score (ISS) chest wall, lung, and pleura. During the study period 92 patients were managed according to the standardized protocol. The outcomes of those patients were compared to 99 patients treated before the TTP. Multivariate logistic regression analysis was performed to assess the independent effect of each variable of the protocol on selected outcomes. Demographics, injury severity, and trauma mechanisms were similar among the groups. As expected, protocol compliance increased after the implementation of the TTP. There was a significant reduction (pchest tube days. Respiratory therapy was independently linked to significant reduction (pchest tube insertion was associated with significant (ptrauma patients.

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

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

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)

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%

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

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

Structural integrity assessments of steam generator tubes using the FAD methodology

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)

2015-12-15

Highlights: • The Failure Assessment Diagram (FAD) is used to assess cracked steam generator tubes. • Typical loading conditions and reported tensile and fracture properties are used. • The FAD is capable to predict the failure mode for different cracks and loads. • The FAD can be used to reduce the conservatism of the current plugging criteria. • Appropriate tensile and fracture properties at operating conditions are required. - Abstract: Steam generator tubes (SGTs) represents up to 60% of the total primary pressure retaining boundary area of a nuclear power plant. They have been found susceptible to diverse degradation mechanisms during service. Due to the significance of a SGT failure on the plant safe operation, nuclear regulatory authorities have established tube plugging or repairing criteria which are based on the defect depth. The widespreadly used “40% criterion” proposed in the 70s is an example whose use is still recommended in the last editions of the ASME Boiler and Pressure Vessel Code. In the present work, an alternative, more realistic and less conservative methodology for SGT integrity evaluation is proposed. It is based on the Failure Assessment Diagram (FAD) and takes advantage of the recent developments in non-destructive techniques which allow a more comprehensive characterization of tube defects, i.e., depth, length, orientation and type. The proposed approach has been applied to: the study of the influence of primary and secondary stresses on tube integrity; the prediction of failure mode (i.e., ductile fracture or plastic collapse) of defective SGTs for varied crack geometries and loading conditions; the analysis of the sensibility of tensile and fracture properties with temperature. The potentiality of the FAD as a comprehensive methodology for predicting the failure loads and failure modes of flawed SGTs is highlighted.

Structural integrity assessments of steam generator tubes using the FAD methodology

International Nuclear Information System (INIS)

Bergant, Marcos A.; Yawny, Alejandro A.; Perez Ipiña, Juan E.

2015-01-01

Highlights: • The Failure Assessment Diagram (FAD) is used to assess cracked steam generator tubes. • Typical loading conditions and reported tensile and fracture properties are used. • The FAD is capable to predict the failure mode for different cracks and loads. • The FAD can be used to reduce the conservatism of the current plugging criteria. • Appropriate tensile and fracture properties at operating conditions are required. - Abstract: Steam generator tubes (SGTs) represents up to 60% of the total primary pressure retaining boundary area of a nuclear power plant. They have been found susceptible to diverse degradation mechanisms during service. Due to the significance of a SGT failure on the plant safe operation, nuclear regulatory authorities have established tube plugging or repairing criteria which are based on the defect depth. The widespreadly used “40% criterion” proposed in the 70s is an example whose use is still recommended in the last editions of the ASME Boiler and Pressure Vessel Code. In the present work, an alternative, more realistic and less conservative methodology for SGT integrity evaluation is proposed. It is based on the Failure Assessment Diagram (FAD) and takes advantage of the recent developments in non-destructive techniques which allow a more comprehensive characterization of tube defects, i.e., depth, length, orientation and type. The proposed approach has been applied to: the study of the influence of primary and secondary stresses on tube integrity; the prediction of failure mode (i.e., ductile fracture or plastic collapse) of defective SGTs for varied crack geometries and loading conditions; the analysis of the sensibility of tensile and fracture properties with temperature. The potentiality of the FAD as a comprehensive methodology for predicting the failure loads and failure modes of flawed SGTs is highlighted.

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)

Root cause analysis of SG tube leakage at Fessenheim unit 2 in 2008

International Nuclear Information System (INIS)

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

2015-01-01

In February 2008, a primary-to-secondary leak caused an unscheduled shutdown at Fessenheim Unit 2 NPP. A circumferential crack was observed just above the top support plate of Row 12 Column 62 U-bend tube on Steam Generator (SG) number 3, which has been attributed to high cycle fatigue. This tube was pulled out in 2011, just before the SG replacement at the third decenal outage, in order to perform exhaustive metallurgical investigations. The destructive examinations revealed that the circumferential crack (70 degrees of extension) was due to high cycle fatigue, with several external initiation areas associated with the presence of small piles of Intergranular Attack (IGA) (600 MA tube) and with very low stress intensity factors ΔK (close to the non-propagating threshold region). This paper complements the metallurgical investigations by carrying out numerical analyses (thermal-hydraulic computation, fluid-elastic instability evaluation, tube vibratory response analysis and fatigue evaluation). The first objective of the study is to attempt to clarify the effect of IGA and the role of several competing factors that could be involved in the tube vibration induced fatigue failure. From these results, a root cause analysis of the R12C62 tube fatigue failure is then provided. It appears that a combination of various factors led to the failure of the tube

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)

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.

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)

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

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

Mechanical design of the storage tubes in the HWVP canister storage building

International Nuclear Information System (INIS)

Divona, C.J.; Fages, R.; Janicek, G.P.; Mullally, J.A.

1993-01-01

Canisters of high-level waste from the Hanford Waste Vitrification Plant (HWVP) will be stored in an adjacent facility, the Canister Storage Building (CSB). The canisters are stored vertically in an array of tubes within the shielded vault area of the CSB. This paper describes the mechanical design of the storage tubes, the shield floor plugs that confine the waste within the tubes and the impact absorber system used to assure that the canisters are not breached in the event of an accidental drop. Installation and testing of the components is also discussed

Study on the Leak Rate Estimation of SG Tubes and Residual Stress Estimation based on Plastic Deformation

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Chang, Yoon Suk; Lee, Dock Jin; Lee, Tae Rin; Choi, Shin Beom; Jeong, Jae Uk; Yeum, Seung Won [Sungkyunkwan University, Seoul (Korea, Republic of)

2009-02-15

In this research project, a leak rate estimation model was developed for steam generator tubes with through wall cracks. The modelling was based on the leak data from 23 tube specimens. Also, the procedure of finite element analysis was developed for residual stress calculation of dissimilar metal weld in a bottom mounted instrumentation. The effect of geometric variables related with the residual stress in penetration weld part was investigated by using the developed analysis procedure. The key subjects dealt in this research are: 1. Development of leak rate estimation model for steam generator tubes with through wall cracks 2. Development of the program which can perform the structure and leakage integrity evaluation for steam generator tubes 3. Development of analysis procedure for bottom mounted instrumentation weld residual stress 4. Analysis on the effects of geometric variables on weld residual stress It is anticipated that the technologies developed in this study are applicable for integrity estimation of steam generator tubes and weld part in NPP.

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)

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

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

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.

Short intergranular cracks in the piecewise anisotropic continuum model of the microstructure

International Nuclear Information System (INIS)

Cizelj, L.; Kovse, I.

2001-01-01

Computational algorithms aiming at modeling and visualization of the initiation and growth of intergranular stress corrosion cracks (e.g., in the steam generator tubes) on the grain-size scale have already been proposed [6]. The main focus of the paper is given to the influence of randomly oriented neighboring grains on the microscopic stress fields at crack tips. The incompatibility strains, which develop along the boundaries of randomly oriented anisotropic grains, are shown to influence the local stress fields at crack tips significantly. Special attention has been paid to the implementation and comparison of different numerical methods estimating the local stress fields at crack tips, aiming at optimizing the computational time and the numerical accuracy of the results. The limited number of calculations indicate that the anisotropic arrangement of grains with local incompatibility strains causes on average about 10% (plane strain) and 26% (plane stress) higher J-integral values at the crack tips than expected from the calculations in the isotropic case.(author)

Eddy current inspection of weld defects in tubing

Science.gov (United States)

Katragadda, G.; Lord, W.

1992-01-01

An approach using differential probes for the inspection of weld defects in tubing is studied. Finite element analysis is used to model the weld regions and defects. Impedance plane signals are predicted for different weld defect types and compared wherever possible with signals from actual welds in tubing. Results show that detection and sizing of defects in tubing is possible using differential eddy current techniques. The phase angle of the impedance plane trajectory gives a good indication of the sizing of the crack. Data on the type of defect can be obtained from the shape of the impedance plane trajectory and the phase. Depending on the skin depth, detection of outer wall, inner wall, and subsurface defects is possible.

Susceptibility of cold-worked zirconium-2.5 wt% niobium alloy to delayed hydrogen cracking

International Nuclear Information System (INIS)

Coleman, C.E.

1976-01-01

Notched tensile specimens of cold-worked zirconium-2.5 wt% niobium alloy have been stressed at 350 K and 520 K. At 350 K, above a possible threshold stress of 200 MPa, specimens exhibited delayed failure which was attributed to hydride cracking. Metallography showed that hydrides accumulated at notches and tips of growing cracks. The time to failure appeared to be independent of hydrogen content over the range 7 to 100 ppm hydrogen. Crack growth rates of about 10 -10 m/s deduced from fractography were in the same range as those necessary to fracture pressure tubes. The asymptotic stress intensity for delayed failure, Ksub(1H), appeared to be about 5 MPa√m. With this low value of Ksub(1H) small surface flaws may propagate in pressure tubes which contain large residual stresses. Stress relieving and modified rolling procedures will reduce the residual stresses to such an extent that only flaws 12% of the wall thickness or greater will grow. At 520 K no failures were observed at times a factor of three greater than times to failure at 350 K. Zirconium-2.5 wt% niobium appears to be safe from delayed hydrogen cracking at the reactor operating temperature. (author)

Slit-burst testing of cold-worked Zr-2.5 wt.% Nb pressure tubing for CANDU-PHW reactors

International Nuclear Information System (INIS)

Wilkins, B.J.S.; Barrie, J.N.; Zink, R.J.

1978-12-01

This report documents the available data on critical crack length of cold-worked Zr-2.5 wt.% Nb pressure tubing in CANDU reactors. In particular, it includes data for tubing removed from the Pickering 3 and 4 reactors. (author)

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

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.

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

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)

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,…

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.

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)

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

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)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-04-01

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

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

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.)

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)

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.

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

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

Layer texture of hot-rolled BCC metals and its significance for stress-corrosion cracking of main gas pipelines

Science.gov (United States)

Perlovich, Yu. A.; Isaenkova, M. G.; Krymskaya, O. A.; Morozov, N. S.

2016-10-01

Based on data of X-ray texture analysis of hot-rolled BCC materials it was shown that the layerwise texture inhomogeneity of products is formed during their manufacturing. The effect can be explained by saturation with interstitial impurities of the surface layer, resulting in dynamical deformation aging (DDA). DDA prevents the dislocation slip under rolling and leads to an increase of lattice parameters in the external layer. The degree of arising inhomogeneity correlates with the tendency of hot-rolled sheets and obtained therefrom tubes to stress-corrosion cracking under exploitation, since internal layers have a compressive effect on external layers, and prevents opening of corrosion cracks at the tube surface.

Secondary side corrosion in steam generator tubes: lessons learned in France from the in-service inspection results

International Nuclear Information System (INIS)

Comby, R.

1997-01-01

Non-destructive testing (NDT) has proved to be very important in the maintenance of steam generator tubing. This is particularly true in the case of secondary side corrosion, because this type of degradation leads to various morphologies which are often complex (intergranular attack) (IGA), intergranular stress corrosion cracking (IGSCC), or a mixture of both. Their detection and characterization by the usual NDT techniques have been achieved through numerous laboratory studies, which were conducted in order to determine the performance and limitations of NDT. Pulled tube examination in a hot laboratory was very valuable, for both NDT and fracture mechanics aspects. The eddy current bobbin coil probe, used for multipurpose inspection of tubes, allows the detection of IGA-SCC at the tube support plate elevation. In France, the use of rotating probes is not required for that type of degradation, since the repair criterion is based on bobbin coil results only. The bobbin coil is also used for detection of IGSCC occurring in free spans, within sludge deposits. The eddy current rotating probe allows, in that case, characterization of main cracks. Concerning the outer diameter initiated circumferential cracks which occur at the top of the tube sheet, only the rotating probe is used. An ultrasonic (UT) inspection was performed several times, in order to obtain information on UT capabilities. The goal of tube inspection is obviously knowledge of the status of steam generators, but also to follow up degradations and to estimate their revolution, and to verify the beneficial effect of some corrective measures, e.g. boric acid injection. (orig.)

Oxidation study by Moessbauer and optic microscopy of steels from boiler tubes used in sugar industry

International Nuclear Information System (INIS)

Fajardo, M.; Perez Alcazar, G.A.; Aguilar, Y.

1998-01-01

Optic microscopy and Moessbauer spectroscopy were used to study the fail and the inner rusted surface of two boiler tubes used in the sugar industry, respectively. The studied tubes, of the type ASTM A 192, were found to have cracks. By optic microscopy it was observed that the failure begins in the inner surface with circumferential cracking. Also, inside and around the surface close to the cracks a rusted layer was detected. Powder from these layers was collected for Moessbauer spectroscopy analysis. By this method the presence of two or three types of Fe oxides such as wuestite, magnetite and hematite, was proved. These results permit to conclude that the failure mechanism was the thermal fatigue due to a hot work in an O 2 -rich vapor atmosphere. The rusted products are stable at high temperatures

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

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)

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...

Final report on development evaluation of Task Group 3 pressure tubes

International Nuclear Information System (INIS)

Fleck, R.G.; Price, E.G.; Cheadle, B.A.

1983-11-01

This report describes the production and evaluation of pressure tubes manufactured to the recommendations of Task Group 3 (TG3) of the Creep Engineering Design Plan. The Zr-2.5 wt percent Nb tubes were manufactured by modified production route to change their metallurgical structure and so reduce the in-service elongation rates. Three modified routes were investigated and a total of twenty-eight tubes produced. There were no difficulties in manufacture and the tubes satisfied the quality assurance and design specifications of reactor grade tubes. Metallurgical evaluation showed that the expected changes in microstructure had occurred but not to the extent anticipated. The TG3 tubes were found to have comparable properties to current tubes when tested for: tensile strength (irradiated and unirradiated); hydride cracking; stress to reorient hydrides; hydrogen diffusion; flaw tolerance; corrosion (irradiated and unirradiated); wear; rolled joint characteristics; irradiation creep and growth. Lower in-service elongation rates are expected for tubes produced by two of the modified routes

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

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...

Failure investigation of super heater tubes of coal fired power plant

Directory of Open Access Journals (Sweden)

A.K. Pramanick

2017-10-01

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

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)

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...

Field trial of a fast single-pass transmit-receive probe during Gentilly II steam generator tube inspection

International Nuclear Information System (INIS)

Obrutsky, L.; Cantin, M.; Renaud, J.; Cecco, V.; Lakhan, R.; Sullivan, S.

2000-01-01

A new generation of transmit-receive single-pass probes, denoted as C6 or X probe, was field tested during the Gentilly II, 2000 steam generator tube inspection. This probe has a performance equivalent to rotating probes and can be used for tubesheet and full-length inspection at an inspection speed equivalent to that of bobbin probes. Existing C3 transmit-receive probes have been demonstrated to be effective in detecting circumferential cracks. The C5 probe can detect both circumferential and axial cracks and volumetric defects but cannot discriminate between them. The C6 probe expands on the capabilities of both probes in a single probe head. It can simultaneously detect and discriminate between circumferential and axial cracks to satisfy different plugging criteria. It has excellent coverage, good defect detectability, and improved sizing and characterization. Probe data is displayed in C-scan format so that the amount of data to be analyzed is similar to rotating probes. The C6 probe will significantly decrease inspection time and the need for re-inspection and tube pulling. This paper describes the advantages of the probe and demonstrates its capabilities employing signals from tube samples with calibration flaws and laboratory induced cracks. It shows the results from the field trial of the probe at Gentilly II and describes the instrumentation, hardware and software used for the inspection. (author)

Field trial of a fast single-pass transmit-receive probe during Gentilly II steam generator tube inspection

International Nuclear Information System (INIS)

Obrutsky, L.; Cantin, M.; Renaud, J.; Cecco, V.; Lakhan, R.; Sullivan, S.

2000-01-01

A new generation of transmit-receive single-pass probes, denoted as C6 or X probe, was field-tested during the Gentilly II, 2000 steam generator tube inspection. This probe has a performance equivalent to rotating probes and can be used for tubesheet and full-length inspection at an inspection speed equivalent to that of bobbin probes. Existing C3 transmit-receive probes have been demonstrated to be effective in detecting circumferential cracks. The C5 probe can detect both circumferential and axial cracks and volumetric defects but cannot discriminate between them. The C6 probe expands on the capabilities of both probes in a single probe head. It can simultaneously detect and discriminate between circumferential and axial cracks to satisfy different plugging criteria. It has excellent coverage, good defect detectability, and improved sizing and characterization. Probe data is displayed in C-scan format so that the amount of data to be analyzed is similar to rotating probes. The C6 probe will significantly decrease inspection time and the need for re-inspection and tube pulling. This paper describes the advantages of the probe and demonstrates its capabilities employing signals from tube samples with calibration flaws and laboratory induced cracks. It shows the results from the field trial of the probe at Gentilly II and describes the instrumentation, hardware and software used for the inspection. (author)

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

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.

Oxidation study by Mössbauer and optic microscopy of steels from boiler tubes used in sugar industry

Science.gov (United States)

Fajardo, M.; Pérez Alcázar, G. A.; Aguilar, Y.

1998-08-01

Optic microscopy and Mössbauer spectroscopy were used to study the fail and the inner rusted surface of two boiler tubes used in the sugar industry, respectively. The studied tubes, of the type ASTM A 192, were found to have cracks. By optic microscopy it was observed that the failure begins in the inner surface with circumferential cracking. Also, inside and around the surface close to the cracks a rusted layer was detected. Powder from these layers was collected for Mössbauer spectroscopy analysis. By this method the presence of two or three types of Fe oxides such as wüstite, magnetite and hematite, was proved. These results permit to conclude that the failure mechanism was the thermal fatigue due to a hot work in an O2 -rich vapor atmosphere. The rusted products are stable at high temperatures.

Research perspectives on the evaluation of steam generator tube integrity

International Nuclear Information System (INIS)

Muscara, J.; Diercks, D.R.; Majumdar, S.; Kupperman, D.S.; Bakhtiari, S.; Shack, W.J.

2002-01-01

Industry efforts have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, stress corrosions cracking (SCC) and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding, there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by 'plug or repair on detection' because current NDE techniques for characterization of flaws and the knowledge of SCC crack growth rates are not accurate enough to permit continued operation. Replacement generators with improved designs and materials have performed well to date, but previous experience with the appearance of some types of SCC in Alloy 600 after 10 years or more of operation and laboratory results suggest additional understanding of corrosion performance of these materials is needed. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators. (author)

Standard test method for crack-tip opening displacement (CTOD) fracture toughness measurement

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers the determination of critical crack-tip opening displacement (CTOD) values at one or more of several crack extension events, and may be used to measure cleavage crack initiation toughness for materials that exhibit a change from ductile to brittle behavior with decreasing temperature, such as ferritic steels. This test method applies specifically to notched specimens sharpened by fatigue cracking. The recommended specimens are three-point bend [SE(B)], compact [C(T)], or arc-shaped bend [A(B)] specimens. The loading rate is slow and influences of environment (other than temperature) are not covered. The specimens are tested under crosshead or clip gage displacement controlled loading. 1.1.1 The recommended specimen thickness, B, for the SE(B) and C(T) specimens is that of the material in thicknesses intended for an application. For the A(B) specimen, the recommended depth, W, is the wall thickness of the tube or pipe from which the specimen is obtained. Superficial surface machini...

Iodine-induced stress corrosion cracking of fixed deflection stressed slotted rings of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Sejnoha, R.; Wood, J.C.

1978-01-01

Stress corrosion cracking of Zircaloy fuel cladding by fission products is thought to be an important mechanism influencing power ramping defects of water-reactor fuels. We have used the fixed-deflection stressed slotted-ring technique to demonstrate cracking. The results show both the sensitivity and limitations of the stressed slotted-ring method in determining the responses of tubing to stress corrosion cracking. They are interpreted in terms of stress relaxation behavior, both on a microscopic scale for hydrogen-induced stress-relief and on a macroscopic scale for stress-time characteristics. Analysis also takes account of nonuniform plastic deformation during loading and residual stress buildup on unloading. 27 refs

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

Premature failure of dissimilar metal weld joint at intermediate temperature superheater tube

OpenAIRE

Al Hajri, Mohammed; Malik, Anees U.; Meroufel, Abdelkader; Al-Muaili, Fahd

2015-01-01

Dissimilar metal weld (DMW) joint between alloyed steel (AS) and stainless steel (SS) failed at one of intermediate temperature superheater (ITSH) tube in steam/power generation plant boiler. The premature failure was detected after a relatively short time of operation (8 years) where the crack propagated circumferentially from AS side through the ITSH tube. Apart from physical examination, microstructural studies based on optical microscopy, SEM and EDX analysis were performed. The results o...

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

Analytical Tem Comparisons of Stress-Corrosion-Crack Microstructures in Alloy 600 under Steam-Generator Service and Laboratory Test Conditions

International Nuclear Information System (INIS)

Thomas, L.E.; Bruemmer, S.M.; Scott, P.M.

2002-01-01

High-resolution analytical transmission electron microscopy (ATEM) has been used to characterize stress-corrosion cracks (SCC) in Alloy 600 steam-generator (SG) tubing from tests with caustic and acid-sulfate solutions. The aim of this work was to identify the microstructural and microchemical signatures of intergranular attack and cracking produced under well-controlled test conditions in order to determine the local environments promoting degradation in service. Cross-sectioned cracks and crack tips were examined in samples of mill-annealed alloy 600 tested in concentrated caustic and acid-sulfate solutions at 320 C. Characteristic microstructures observed in the caustic (10% NaOH) test sample included deeply penetrative attack along crack-intersected grain boundaries, with Cr-rich spinel and NiO structure oxides ranging from random nanocrystalline to oriented epitaxial films filling cracks up to the tips. Sodium was readily detectable in the oxides (up to 5 wt.% in the spinel corrosion product) along with S and Cu enrichment at crack-wall metal/oxide interfaces and local attack of the metal matrix around IG carbide particles. In the sulfate (Na 2 SO 4 + FeSO 4 ) test sample, the grain boundaries were also deeply attacked/cracked. Epitaxial NiO-structure oxide formed on the crack walls and S, sometimes with Cu, was concentrated between the oriented oxide layers rather than along the metal/oxide interfaces. Carbides were attacked and partially converted to fine-grained oxide containing up to several percent S. Observations of crack tips in the acid sulfate sample also revealed nm-wide cracks preceding the oxide along grain boundaries. The SCC structures produced in the laboratory tests differed in most details from the secondary-side SCC structures observed in pulled SG tubes. Important differences included the oxide morphologies, the presence of easily detectable Na and absence of sulfides in the test samples, different types of attack on IG carbide particles

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.

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.

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

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)

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.

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)

The stress corrosion cracking of hard 2 1/4 CrMo steel in water at 2000 and 3000C

International Nuclear Information System (INIS)

Hurst, P.; Appleton, D.A.; Hurley, J.R.F.; Pennington, C.

1983-01-01

An account is given of experiments performed in 200 0 or 300 0 C water to evaluate the susceptibility of the quench-hardened steel to stress corrosion cracking. The work has covered self-stressed specimens (U-bends and C-rings), and constant load tests tensile specimens and tube/tube plate welds of the type used for the UK Prototype Fast Reactor. At 200 0 C, the effects have been examined of strength, stress and oxygen level; at 300 0 C the effect of quenching temperature (1400 or 1050 0 C) has been studied. Different mechanisms may be responsible at the two test temperatures. Hydrogen absorption in the region of any localised corrosion is believed to be mechanistically significant in the case of 200 0 C cracking, but general embrittlement does not occur. At 300 0 C the cracking has been linked with the increased probability of grain boundary segregation arising from the higher quenching temperature. The value of shot-peening as a means of inducing surface compressive stress, and hence reducing the risk of cracking, has been demonstrated and the factors that could counter-act its usefulness have been identified. (author)

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

Fracture toughness of irradiated Zr-2.5Nb pressure tube from KAPS-2 evaluated using disk compact tension specimens

International Nuclear Information System (INIS)

Shah, Priti Kotak; Dubey, J.S.; Balakrishnan, K.S.; Shriwastaw, R.S.; Dhotre, M.P.; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

The report gives the results of the fracture toughness tests carried out over the range of temperatures on specimens prepared from the irradiated S-07Zr-2.5Nb pressure tube removed from Kakrapar Atomic Power Station-2 (KAPS-2) as a part of materials surveillance programme. The pressure tube had experienced ∼ 8 effective full power years (EFPY) of reactor operation and had hydrogen equivalent (H eq ) content less than 20 ppm along the tube length. The fracture toughness tests have been carried out using 30 mm Disk Compact Tension (DCT) specimens, that were punched out of the irradiated pressure tube. The disk punching was carried out using specially made shielded enclosure and hydraulic press. Fatigue pre-cracking and fracture toughness tests were performed using servo-hydraulic universal testing machine with Direct Current Potential Drop (DCPD) equipment to monitor the crack length. The tests were carried out at different test temperature from ambient to 300℃. The fracture toughness values have been used to estimate the critical pressure for the tube. The fracture properties indicate that such tubes have sufficient toughness to satisfy the Leak-Before-Break (LBB) criterion for in-reactor operation. (author)

Investigations On Crack Propagation Under Cyclical Isothermal And Thermo-mechanical Loadings For A Type 304-L Stainless Steel Used For Pressurized Water Reactor

Directory of Open Access Journals (Sweden)

Gourdin Cédric

2018-01-01

Full Text Available The integrity of structures exhibiting flaws in Pressurized Water Reactor (PWR has to be assessed to meet safety criteria. This paper deals with crack-propagation under cyclic thermo-mechanical loadings, as encountered in class I austenitic pipes of PWR’s. To have a conservative and reliable assessment of the crack propagation due to the in-service loading, various codes and standards use simplified method. For example, the RSE-M Code introduces a plastic correction depending on the proportion of the mechanical loading. An improvement of the current method requires additional investigations. Moreover, components loaded with transient or thermal fluctuations are not really in loadcontrolled conditions. To this end, a device called PROFATH was designed. The specimen is a pre-cracked thick-walled tube undergoing a set of thermal cycles and loaded with a static mechanical force. During the first part of the thermal cycle, a high frequency induction coil heats the external wall of the tube. Then, the heating system stops and the specimen is cooled down by running water inside the tube. Finite element calculations show that only a region half-way along the tube should be heated to ensure adequate structural effect. In the heated zone, the machining of a sharp circumferential groove ensures the propagation of a unique crack. An electro-mechanical jack controls the level of the mechanical static load. Tests have been carried out, and these tests allow having an evaluation of the pertinence of the correction proposed by the RSE-M Code for a significant plasticity.

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

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...

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

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

Evaluation of the eddy-current method of inspecting steam generator tubing

International Nuclear Information System (INIS)

Flora, J.H.; Brown, S.D.; Weeks, J.R.

1976-01-01

The objective of this project has been to evaluate the eddy-current method of inspecting steam generator tubing by conducting a series of laboratory experiments with conventional eddy-current equipment. The experiments have involved obtaining eddy-current measurements on samples of 7/8-inch OD Inconel-600 tubing provided by the Westinghouse Nuclear Energy Systems Division. A variety of machined defects and some chemically induced flaws, such as stress corrosion cracks were fabricated in the tubing. Statistical evaluation of the data was employed to estimate the error encountered in measuring corrosion defects of various depths. It appears that the eddy-current technique can provide a reasonable measure of defect depth under certain conditions. On the other hand, the evaluation indicates that it is difficult to determine the depth of certain types of flaws with reliability and precision. Furthermore, although some defects as shallow as 10 percent of the tube wall could be detected, it was not possible to detect other types of flaws that were less than 40 percent deep even when the tube supports were not near the defects. The difficulty in detecting small volume flaws is attributed to low signal-to-noise ratio. Noise is a result of unwanted signals from test variables, such as wobble and variations in tube properties. The error in measurement of certain types of larger defects is associatedin part with test variables and also with the effects that the geometry of the defect has on the eddy-current signal patterns. The distortions in signal patterns caused by gradual wastage type defects and the poor reproducibility of signal patterns obtained from notches that represent stress corrosion cracks are described. Some developments that will rectify these detection and depth measurement problems are discussed

Burst pressure and leak rate from fretted SG tubes

International Nuclear Information System (INIS)

Hwang, Seong Sik; Jung, Man Kyo; Kim, Hong Pyo; Kim, Joung Soo

2005-01-01

Steam generator(SG) tubes of a pressurized water reactor(PWR) have suffered from various types of corrosion, such as pitting, wastage and stress corrosion cracking (SCC) on both the primary and secondary side. Recently, fretting/wear degradation at the tube support region has been reported in some Korean nuclear power plants. In order to prevent the primary coolant from leaking to the secondary side, the tubes are repaired by a sleeving or plugging. It is important to establish the repair criteria to assure a reactor integrity and yet maintain the plugging ratio within the limits needed for an efficient operation. The objective of the burst test is to obtain a relationship between the burst/leak rate and the shape of the fretted flaws machined with an electro discharge machining (EDM)

Propagation of stress-corrosion cracks in unirradiated zircaloy

International Nuclear Information System (INIS)

Norring, K.; Haag, Y.; Wikstroem, C.

1982-01-01

Propagation of iodine-induced stress-corrosion cracks in Zircaloy was studied using pre-cracked and internally pressurized cladding tubes. These were recrystallized at different temperatures, to obtain grain sizes between 4 μm and 10 μm. No statistically significant difference in propagation rate due to the difference in grain size was observed. If the obtained data, with Ksub(I) values ranging from 4 to 11 MNmsup(-3/2), were log-log plotted (da/dt = CKsub(I)sup(N)), as usual, they fell within the scatter-band of data reported earlier. But from this plot it could also be seen that the Ksub(I) interval can be divided into two separate parts having different da/dt-Ksub(I) relations. The transition takes place at a Ksub(I) value of about 8 MNmsup(-3/2). The region with lower Ksub(I) values shows a substantially lower n value than the upper region (2.4 and 9.8 respectively), and earlier reported values (n = 7 to 10). This transition is in good agreement with a transition from an intergranular to a transgranular propagation mode of the stress-corrosion crack. (orig.)

Cracks propagation by stress corrosion cracking in conditions of Boiling Water Reactor (BWR); Propagacion de grietas por corrosion bajo esfuerzo en condiciones de reactor de agua hirviente (BWR)

Energy Technology Data Exchange (ETDEWEB)

Fuentes C, P

2003-07-01

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{sub 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

Quality assurance and quality control in fabrication of heat exchanger tubes

International Nuclear Information System (INIS)

Duennewald, A.

1980-01-01

Object of this report is the manufacture of heat exchanger tubes. A comprehensive manufacturing and test program has to be established to assure and prove and equal tube quality. This requires a functionally operating quality assurance system combined with a production exactly planned in advance. A specific continuous production line for heat exchanger tubes has been erected at the Hellenthal plant of the Mannesmannroehren-Werke. All production steps and heat treatments are generally controlled by a quality control department. Non-destructive testing of each tube produced in standard length is performed on several agregates in line using ultrasonic and/or eddy current technique. All tests are generally performed in the presence of quality inspectors or surveyors. For a lot of heat exchangers the straight tubes have to be hairpin bended. To avoid the risk of stress corrosion cracking, it is recommended to procreate defined compression stresses in the outside tube surface. Prior to releasing the tubes to shipment, the completeness of the documentation as to the manufacturing steps and inspection agreed upon is thoroughly checked. (RW)

Ultrasonic inspection experience of steam generator tubes at Ontario Hydro and the TRUSTIE inspection system

International Nuclear Information System (INIS)

Choi, E.I.; Jansen, D.

1998-01-01

Ontario Hydro have been using ultrasonic test (UT) technique to inspect steam generator (SG) tubes since 1993. The UT technique has higher sensitivity in detecting flaws in SG tubes and can characterize the flaws with higher accuracy. Although an outside contractor was used initially, Ontario Hydro has been using a self-developed system since 1995. The TRUSTIE system (Tiny Rotating UltraSonic Tube Inspection Equipment) was developed by Ontario Hydro Technologies specifically for 12.7 mm outside diameter (OD) tubes, and later expanded to larger tubes. To date TRUSTIE has been used in all of Ontario Hydro's nuclear generating stations inspecting for flaws such as pitting, denting, and cracks at top-of-tubesheet to the U-bend region. (author)

Fatigue life evaluation of a cracked pipeline pipe operating under variable; Estudo da vida em fadiga com variacao de pressao interna em tubos com defeitos

Energy Technology Data Exchange (ETDEWEB)

Betanzos, J.; Mattei, F. [Rio Grande do Sul Univ., Porto Alegre, RS (Brazil); Strohaecker, T. [Rio Grande do Sul Univ., Porto Alegre, RS (Brazil). Programa de Pos-graduacao em Engenharia de Minas, Metalurgia e Materiais; Marinho, R.; Hoppe, R.; Donato, G. [PETROBRAS, Rio de Janeiro, RJ (Brazil)

2005-07-01

Fatigue cracks in welded joints are commonly found in gas and oil pipelines and usually pumping is interrupted for repair. The strength of welded joints with defects, and consequently, the useful life of a pipeline, is an important information that could lead to great improvements in the reliability. To evaluate the working conditions determined by standards, specimens with 22'' diameter and 1/2'' wall thickness API 5L X60 pipes were used. Two semi-elliptical cracks were electro spark machined, with 50% and 25% of the tube wall thickness depths and a length/depth ratio of 1/20. Both defects were located in the HAZ of the tube longitudinal seam weld. Applying a cyclic internal pressure it was possible to propagate a fatigue crack until a through-thickness condition. The results show that the pressures to initiate crack propagation are too much higher than the pressures allowed by the standards for this kind of defect. (author)

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.

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

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