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Sample records for generator tube rupture

  1. Steam generator tube rupture (SGTR) scenarios

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kweon, Young Chul

    2002-01-01

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

  3. Radioactivity transport following steam generator tube rupture

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-03-01

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

  4. CATHENA simulations of steam generator tube rupture

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  5. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

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

  6. 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.

  7. 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.

  8. Corrosion and Rupture of Steam Generator Tubings in PWRs

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-01

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

  9. Corrosion and Rupture of Steam Generator Tubings in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-15

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  11. Indian Point 2 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    Dayan, A.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Rodliffe, R.S.

    1983-08-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  14. 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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Kim, Gibeom; Heo, Gyunyoung; Kim, Hyeonmin

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

    Fontes, S.W.F.

    1989-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-10-01

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

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

    International Nuclear Information System (INIS)

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

    1986-10-01

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

  2. 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.

  3. 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)

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. 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)

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

    International Nuclear Information System (INIS)

    Depond, G.; Sureau, H.

    1984-01-01

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

  7. Steam generator tube rupture: studies to improve plant procedure

    International Nuclear Information System (INIS)

    Tellier, N.; Zilliox, C.

    1984-10-01

    These accidents have the particularities to lead to atmospheric radioactive release and to be able to be determinated with appropriate operator actions. These radioactive releases are function of several parameters of which sensitivity is analyzed. The major part of the calculations were performed by EDF with an home made code called ''AXEL''. The main conclusions are: - the optimization of the safety injection monitoring to minimize radioactive releases to atmosphere, while ensuring the cooling of the core; - the radioactive releases to atmosphere are very low in any case but much more important if the filling of the steam generator secondary side cannot be avoided

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-12-01

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

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

    International Nuclear Information System (INIS)

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

    2010-12-01

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

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

    International Nuclear Information System (INIS)

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

    1991-12-01

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

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

    International Nuclear Information System (INIS)

    1982-04-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    Science.gov (United States)

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

    1994-01-01

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

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

    International Nuclear Information System (INIS)

    Luu Nam Hai; Truong Cong Thang

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Frano, R. Lo; Forasassi, G.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    1998-03-01

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

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

    International Nuclear Information System (INIS)

    Liao, Y.; Guentay, S.

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-12-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

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

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

    International Nuclear Information System (INIS)

    Yang Hongrun; Hidaka, Akihide; Sugimoto, Jun

    1999-03-01

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

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

    International Nuclear Information System (INIS)

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

    1977-10-01

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

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

    International Nuclear Information System (INIS)

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

    1977-10-01

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

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

    International Nuclear Information System (INIS)

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

    1977-10-01

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

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

    International Nuclear Information System (INIS)

    Esparza, V.; Sackett, K.E.

    1977-10-01

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

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

    International Nuclear Information System (INIS)

    Petkov, M.; Petkov, G.

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Majumdar, S.; Kasza, K.

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-07-01

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

  16. Steam generator tube failures

    International Nuclear Information System (INIS)

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

    1996-04-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-10-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Hamada, H.; Kurihara, A.

    2003-05-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

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

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Parrish, K.R.

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    Parrish, K.R.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

    Wichner, R.P.

    1977-01-01

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

  10. Pressure tube rupture in a closed tank

    International Nuclear Information System (INIS)

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

    1985-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-01

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

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

    International Nuclear Information System (INIS)

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

    1987-03-01

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

  13. Steam generator tube extraction

    International Nuclear Information System (INIS)

    Delorme, H.

    1985-05-01

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

  14. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

    Schwab, P.R.

    1981-01-01

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

  15. CFD modeling of a boiler's tubes rupture

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  20. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    International Nuclear Information System (INIS)

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J.

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR

  1. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    Energy Technology Data Exchange (ETDEWEB)

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

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

    International Nuclear Information System (INIS)

    Komine, Ryuji; Wada, Yusaku

    1998-08-01

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

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

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

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

  4. Application of the visual system analyzer (ViSA): simulation of the steam generator tube rupture event at Ulchin unit 4

    International Nuclear Information System (INIS)

    Lee, S.W.; Kim, K.D.; Hwang, M.K.; Jeong, J.J.

    2004-01-01

    Korea Atomic Energy Research Institute (KAERI) has developed the Visual System Analyzer (ViSA) based on the best-estimate (B-E) codes, MARS and RETRAN-3D. The key features of ViSA are: (1) The use of the same input and the same level of accuracy as the original codes is guaranteed (2) Users can design their own plant mimic by a drag-and-drop from the provided indicators (3) The on-line interactive control enables users to simulate the operator's actions (4) The nodalization window is designed to display the transient temperature and void distributions. ViSA is composed of two parts; the B-E code with plant input and the Graphic User Interface (GUI) that includes the plant mimic and an interactive control function, etc. The calculation results of the B-E code are transferred to a user via the GUI and a user can apply the operator action to the B-E code using an interactive control function. Therefore, it is not necessary to prepare complex control input data to simulate the various manual operations which may occur during the plant transient. In this study, the Steam Generator Tube Rupture (SGTR) Accident, which occurred at Ulchin Unit 4 in April 2002, has been simulated using ViSA and the simulation results are compared with the measured plant data. The RETRAN-3D plant input data used in this simulation is a genetic input deck prepared for the simulation from a normal operation condition to a Small-Break LOCA. From the results of the SGTR simulation, we found that the GUI functions of ViSA and the input data for Ulchin Unit 4 have enough effectiveness and soundness. (author)

  5. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  6. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  7. 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

  8. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1992-03-01

    A survey of steam generator operating experience for 1986 has been carried out for 184 pressurized water and pressurized heavy-water reactors, and 1 water-cooled, graphite-moderated reactor. Tubes were plugged at 75 of the reactors (40.5%). In 1986, 3737 tubes were plugged (0.14% of those in service) and 3148 tubes were repaired by sleeving. A small number of reactors accounted for the bulk of the plugged tubes, a phenomenon consistent with previous years. For 1986, the available tubesheet sludge data for 38 reactors has been compiled into tabular form, and sludge/deposit data will be incorporated into all future surveys

  9. 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

  10. 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

  11. CFD modeling of a boiler's tubes rupture

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-12-15

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

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

    African Journals Online (AJOL)

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

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

    International Nuclear Information System (INIS)

    Dillon, R.L.

    1981-01-01

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

  14. 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

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

    International Nuclear Information System (INIS)

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

    1982-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-11-15

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  18. Double wall steam generator tubing

    International Nuclear Information System (INIS)

    Padden, T.R.; Uber, C.F.

    1983-01-01

    Double-walled steam generator tubing for the steam generators of a liquid metal cooled fast breeder reactor prevents sliding between the surfaces due to a mechanical interlock. Forces resulting from differential thermal expansion between the outer tube and the inner tube are insufficient in magnitude to cause shearing of base metal. The interlock is formed by jointly drawing the tubing, with the inside wall of the outer tube being already formed with grooves. The drawing causes the outer wall of the inner tube to form corrugations locking with the grooves. (author)

  19. Steam generator tube integrity program

    International Nuclear Information System (INIS)

    Dierks, D.R.; Shack, W.J.; Muscara, J.

    1996-01-01

    A new research program on steam generator tubing degradation is being sponsored by the U.S. Nuclear Regulatory Commission (NRC) at Argonne National Laboratory. This program is intended to support a performance-based steam generator tube integrity rule. Critical areas addressed by the program include evaluation of the processes used for the in-service inspection of steam generator tubes and recommendations for improving the reliability and accuracy of inspections; validation and improvement of correlations for evaluating integrity and leakage of degraded steam generator tubes, and validation and improvement of correlations and models for predicting degradation in steam generator tubes as aging occurs. The studies will focus on mill-annealed Alloy 600 tubing, however, tests will also be performed on replacement materials such as thermally-treated Alloy 600 or 690. An overview of the technical work planned for the program is given

  20. 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

  1. 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)

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

    International Nuclear Information System (INIS)

    Atsumo, H.

    1977-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  7. 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

  8. The Sealed Tube Neutron Generator

    International Nuclear Information System (INIS)

    Tunnell, L.N.; Beyerle, A.; Durkee, R.; Headley, G.; Hurley, P.

    1992-01-01

    A Sealed Tube Neutron Generator (STNG) has been designed and tested at Special Technologies Laboratories (STL) in Santa Barbara, California. Unlike similar tubes that have been used for years in other applications, e.g., by the oil well logging industry, the present device was designed primarily to be part of the Associated Particle Imaging (API) system. Consequently, the size and quality of the neutron spot produced by the STNG is of primary importance. Results from initial measurements indicate that performance goals are satisfied

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. Impulse generation by detonation tubes

    Science.gov (United States)

    Cooper, Marcia Ann

    Impulse generation with gaseous detonation requires conversion of chemical energy into mechanical energy. This conversion process is well understood in rocket engines where the high pressure combustion products expand through a nozzle generating high velocity exhaust gases. The propulsion community is now focusing on advanced concepts that utilize non-traditional forms of combustion like detonation. Such a device is called a pulse detonation engine in which laboratory tests have proven that thrust can be achieved through continuous cyclic operation. Because of poor performance of straight detonation tubes compared to conventional propulsion systems and the success of using nozzles on rocket engines, the effect of nozzles on detonation tubes is being investigated. Although previous studies of detonation tube nozzles have suggested substantial benefits, up to now there has been no systematic investigations over a range of operating conditions and nozzle configurations. As a result, no models predicting the impulse when nozzles are used exist. This lack of data has severely limited the development and evaluation of models and simulations of nozzles on pulse detonation engines. The first experimental investigation measuring impulse by gaseous detonation in plain tubes and tubes with nozzles operating in varying environment pressures is presented. Converging, diverging, and converging-diverging nozzles were tested to determine the effect of divergence angle, nozzle length, and volumetric fill fraction on impulse. The largest increases in specific impulse, 72% at an environment pressure of 100 kPa and 43% at an environment pressure of 1.4 kPa, were measured with the largest diverging nozzle tested that had a 12° half angle and was 0.6 m long. Two regimes of nozzle operation that depend on the environment pressure are responsible for these increases and were first observed from these data. To augment this experimental investigation, all data in the literature regarding

  11. Safety significance of steam generator tube degradation mechanisms

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-01

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

  12. 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

  13. PWR steam generator tubing sample library

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    In order to compile the tubing sample library, two approaches were employed: (a) tubing sample replication by either chemical or mechanical means, based on field tube data and metallography reports for tubes already destructively examined; and (b) acquisition of field tubes removed from operating or retired steam generators. In addition, a unique mercury modeling concept is in use to guide the selection of replica samples. A compendium was compiled that summarizes field observations and morphologies of steam generator tube degradation types based on available NDE, destructive examinations, and field reports. This compendium was used in selecting candidate degradation types that were manufactured for inclusion in the tube library

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  16. In service inspection for steam generator tubes

    International Nuclear Information System (INIS)

    Comby, R.; Eyrolles, Ph.

    1988-01-01

    In this paper the authors show the means putting in place for examination of steam generators tubes. These means (eddy current probes, ultrasonic testing) associated with a knowledge on degradation phenomena allow mapping controlled tubes and limiting undesirable obturations [fr

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

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

    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

  20. 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.

  1. Steam generator tubing NDE performance

    International Nuclear Information System (INIS)

    Henry, G.; Welty, C.S. Jr.

    1997-01-01

    Steam generator (SG) non-destructive examination (NDE) is a fundamental element in the broader SG in-service inspection (ISI) process, a cornerstone in the management of PWR steam generators. Based on objective performance measures (tube leak forced outages and SG-related capacity factor loss), ISI performance has shown a continually improving trend over the years. Performance of the NDE element is a function of the fundamental capability of the technique, and the ability of the analysis portion of the process in field implementation of the technique. The technology continues to improve in several areas, e.g. system sensitivity, data collection rates, probe/coil design, and data analysis software. With these improvements comes the attendant requirement for qualification of the technique on the damage form(s) to which it will be applied, and for training and qualification of the data analysis element of the ISI process on the field implementation of the technique. The introduction of data transfer via fiber optic line allows for remote data acquisition and analysis, thus improving the efficiency of analysis for a limited pool of data analysts. This paper provides an overview of the current status of SG NDE, and identifies several important issues to be addressed

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

    Science.gov (United States)

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

    2013-08-23

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

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

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  5. 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

  6. 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.

  7. A Flue Gas Tube for Thermoelectric Generator

    DEFF Research Database (Denmark)

    2013-01-01

    The invention relates to a flue gas tube (FGT) (1) for generation of thermoelectric power having thermoelectric elements (8) that are integrated in the tube. The FTG may be used in combined heat and power (CHP) system (13) to produce directly electricity from waste heat from, e.g. a biomass boiler...

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

    International Nuclear Information System (INIS)

    Klippel, H.T.

    1985-01-01

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

  9. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

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

  10. Tube sheet design for PFBR steam generator

    International Nuclear Information System (INIS)

    Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1991-01-01

    Top and bottom tube sheets of PFBR Steam Generators have been analysed with 3D and axisymmetric models using CASTEM Programs. Analysis indicates that the effects of piping reactions at the inlet/outlet nozzles on the primary stresses in the tube sheets are negligible and the asymmetricity of the deformation pattern introduced in the tube sheet by the presence of inlet/outlet and manhole nozzles is insignificant. The minimum tube sheet thicknesses for evaporator and reheater are 135 mm and 75 mm respectively. Further analysis has indicated the minimum fillet radius at the junction of tube sheet and dished end should be 20 mm. Simplified methodology has been developed to arrive at the number of thermal baffles required to protect the tube sheet against fatigue damage due to thermal transient. This method has been applied to PFBR steam generators to determine the required number of thermal baffles. For protecting the bottom tube sheet of evaporator against the thermal shock due to feed water and secondary pump trip, one thermal shield is found to be sufficient. Further analysis is required to decide upon the actual number to take care of the severe thermal transient, following the event of sudden dumping of water/steam, immediately after the sodium-water reaction. (author)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  12. CANDU steam generator tubing material service experience and allied development

    International Nuclear Information System (INIS)

    Hart, A.E.; Lesurf, J.E.

    1976-01-01

    This paper covers the following aspects for the tube materials in CANDU-PHW steam generators: inservice performance with respect to tube leaks and coolant activity attributable to boiler tube corrosion, selection of tube materials for use with non-boiling and boiling primary coolants, supporting development on corrosion, vibration, fretting wear, tube inspection, leak detection and plugging of defective tubes. (author)

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

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1996-01-01

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

  14. 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

  15. Neutron generator tube ion source control

    International Nuclear Information System (INIS)

    Bridges, J.R.

    1982-01-01

    A system is claimed for controlling the output of a neutron generator tube of the deuterium-tritium accelerator type and having an ion source to produce sharply defined pulses of neutrons for well logging use. It comprises: means for inputting a relatively low voltage input control pulse having a leading edge and a trailing edge; means, responsive to the input control pulse, for producing a relatively high voltage ion source voltage pulse after receipt of the input pulse; and means, responsive to the input control pulse, for quenching, after receipt of the input pulse, the ion source control pulse, thereby providing a sharply time defined neutron output from the generator tube

  16. Data analysis for steam generator tubing samples

    International Nuclear Information System (INIS)

    Dodd, C.V.

    1996-07-01

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

  17. Eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Neumaier, P.

    1981-01-01

    A rotating probe is described for improving the inspection of tubes and end plate in steam generators. The method allows a representation of the whole defect, consequently the observer is able to determine directly the type of defect, signal processing in-line or off-line is possible [fr

  18. Flow induced pulsations generated in corrugated tubes

    NARCIS (Netherlands)

    Belfroid, S.P.C.; Swindell, R.; Tummers, R.

    2008-01-01

    Corrugated tubes can produce a tonal noise when used for gas transport, for instance in the case of flexible risers. The whistling sound is generated by shear layer instability due to the boundary layer separation at each corrugation. This whistling is examined by investigating the frequency,

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  20. Fatigue cracking on a steam generator tube

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  1. Pulse tube coolers for Meteosat third generation

    International Nuclear Information System (INIS)

    Butterworth, James; Aigouy, Gérald; Chassaing, Clement; Debray, Benoît; Huguet, Alexandre

    2014-01-01

    Air Liquide's Large Pulse Tube Coolers (LPTC) will be used to cool the focal planes of the Infrared Sounder (IRS) and Flexible Combined Imager (FCI) instruments aboard the ESA/Eumetsat satellites Meteosat Third Generation (MTG). This cooler consists of an opposed piston linear compressor driving a pulse tube cold head and the associated drive electronics including temperature regulation and vibration cancellation algorithms. Preparations for flight qualification of the cooler are now underway. In this paper we present results of the optimization and qualification activities as well as an update on endurance testing

  2. Vertical steam generator with slab-type tube-plate with even tube bundle washing

    International Nuclear Information System (INIS)

    Manek, O.; Masek, V.; Motejl, V.; Quitta, R.

    1980-01-01

    A shielding plate supporting the tubes attached to the tube plate of a vertical steam generator is mounted above the tube plate. Tube sleeves are designed with a dimensional tolerance relative to the heat transfer tubes and the sleeve end and the tube plate end. A separate space is thus formed above the tube plate in which circulation or feed water is introduced to flow between the branch and the heat transfer tube. This provides intensive washing of heat transfer tubes at a critical point and prevents deposit formation, thus excluding heat transfer tube failures. (J.B.)

  3. 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)

  4. WWER-1000/320 steam generator collector rupture. Radiological consequences

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova, A; Sartmadzhiev, A; Balabanov, E [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    A model describing a hypothetical accident with direct release of primary coolant to the atmosphere is proposed. Cover lifting of the primary collector due to a rupture of the fixing bolts leads to a coolant release. The initial and boundary conditions of the accident scenario have been selected to provide for the most unfavorable conditions. The total release of primary coolant during the first 15 min of transient are estimated to 50.8 tons, of these 48.5 t with the initial activity in the primary coolant circuit. Without evacuation or sheltering, after 7 days of exposure, the expected dose at the boundary of the restricted zone is 0.0182 Sv for the whole body and 0.184 Sv for the thyroid gland. The effective equivalent dose on the site would be 0.0521 Sv. As a result of the analysis it is concluded that the steam generator collector rupture is not jeopardizing the core heat removal even with a minimum configuration of ECCS as the cooling is accomplished through the steam generators. The radiological consequences of the accident would be relatively small if an emergency procedure is applied at the 15-th minute of the transient. 1 ref.

  5. Method to detect steam generator tube leakage

    International Nuclear Information System (INIS)

    Watabe, Kiyomi

    1994-01-01

    It is important for plant operation to detect minor leakages from the steam generator tube at an early stage, thus, leakage detection has been performed using a condenser air ejector gas monitor and a steam generator blow down monitor, etc. In this study highly-sensitive main steam line monitors have been developed in order to identify leakages in the steam generator more quickly and accurately. The performance of the monitors was verified and the demonstration test at the actual plant was conducted for their intended application to the plants. (author)

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

    International Nuclear Information System (INIS)

    Stanic, D.

    1996-01-01

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

  7. 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.

  8. Integrity Analysis of Damaged Steam Generator Tubes

    International Nuclear Information System (INIS)

    Stanic, D.

    1998-01-01

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

  9. Verification of the computer code ATHLET in the framework of the external verification group ATHLET BETHSY test 9.3 - steam generator U-tube rupture with failure of the high pressure injection. Final report; Verifikation des ATHLET-Rechenprogramms im Rahmen der externen Verifikationsgruppe ATHLET BETHSY Test 9.3 - Heizrohrbruch mit Versagen der Hochdruck-Noteinspeisung. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E.; Schaefer, F. [Forschungszentrum Rossendorf e.V. (FZR) (Germany). Inst. fuer Sicherheitsforschung

    1998-08-01

    In the framework of the external validation of the thermalhydraulic code ATHLET MOD 1.1 CYCLE D, which is being developed by the GRS, post test analyses of two experiments were done, which were performed at the French integral test facility BETHSY. During test 9.3 the consequences of a steam generator U-tube rupture with failure of the high pressure injection and of the auxiliary feedwater supply were investigated. As accident management measures, the depressurization of the secondary sides, first of the two intact steam generators, then of the damaged steam generator and finally the primary depressurization by opening of the pressurizer valve were performed. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. The safety relevant statement could be reproduced. Deviations, which did not impose the general results, occurred concerning the break mass flow during the depressurization of the damaged steam generator and the description of the failure of the heat transfer to the damaged steam generator. Reasons are hardly to find, because these processes are highly complex. (orig.) [Deutsch] Im Rahmen der externen Validierung des von der Gesellschaft fuer Anlagen- und Reaktorsicherheit entwickelten Stoerfallcodes ATHLET, der in der Version Mod 1.1 Cycle D vorlag, wurden zwei Experimente nachgerechnet und analysiert, die an der franzoesischen Versuchsanlage BETHSY durchgefuehrt wurden. Im Test 9.3 werden die Konsequenzen untersucht, wenn bei einem Heizrohrbruch die Hochdruckeinspeisung sowie die Not-Speisewasserversorgung der Dampferzeuger versagen und nur die Druckspeicher sowie die Niederdruckeinspeisung zur Verfuegung stehen. Als Accident Management Massnahmen wurde die sekundaere Druckentlastung und schliesslich die primaere Entlastung ueber den Druckhalter untersucht. Die Analyse kommt zu dem Ergebnis, dass der Code ATHLET in der Lage ist, dieses komplexe Szenario recht gut zu beschreiben. Die

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  11. 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

  12. Study on antioxidant experiment on forged steel tube sheet and tube hole for steam generator

    International Nuclear Information System (INIS)

    Zong Hai; Wang Detai; Ding Yang

    2012-01-01

    Antioxidant experiment on forged steel tube sheet and tube hole for steam generator was studied and the influence of different simulated heat treatments on the antioxidant performance of tube sheet and tube hole was made. The influence of different antioxidant methods on the size of tube hole was drawn. Furthermore, the change of size and weight of 18MnD5 forged steel tube sheet on the condition of different simulated heat treatments was also studied. The analytical results have proved reference information for the use of 18MnD5 material and for key processes of processing tube hole and wearing and expanding U-style tube. (authors)

  13. Analysis of the Steam Generator Tubes Rupture Initiating Event

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  14. Rupture

    CERN Multimedia

    Association du personnel

    2006-01-01

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

  15. Hideout in steam generator tube deposits

    International Nuclear Information System (INIS)

    Balakrishnan, P.V.; Franklin, K.J.; Turner, C.W.

    1998-05-01

    Hideout in deposits on steam generator tubes was studied using tubes coated with magnetite. Hideout from sodium chloride solutions at 279 degrees C was followed using an on-line high-temperature conductivity probe, as well as by chemical analysis of solution samples from the autoclave in which the studies were done. Significant hideout was observed only at a heat flux greater than 200 kW/m 2 , corresponding to a temperature drop greater than 2 degrees C across the deposits. The concentration factor resulting from the hideout increased highly non-linearly with the heat flux (varying as high as the fourth power of the heat flux). The decrease in the apparent concentration factor with increasing deposit thickness suggested that the pores in the deposit were occupied by a mixture of steam and water, which is consistent with the conclusion from the thermal conductivity measurements on deposits in a separate study. Analyses of the deposits after the hideout tests showed no evidence of any hidden-out solute species, probably due to the concentrations being very near the detection limits and to their escape from the deposit as the tests were being ended. This study showed that hideout in deposits may concentrate solutes in the steam generator bulk water by a factor as high as 2 x 10 3 . Corrosion was evident under the deposit in some tests, with some chromium enrichment on the surface of the tube. Chromium enrichment usually indicates an acidic environment, but the mobility required of chromium to become incorporated into the thick magnetite deposit may indicate corrosion under an alkaline environment. An alkaline environment could result from preferential accumulation of sodium in the solution in the deposit during the hideout process. (author)

  16. ARTIST: a cooperative safety project to study fission product retention in a ruptured steam generator

    International Nuclear Information System (INIS)

    Guentay, S.; Dehbi, A.; Suckow, D.; Birchley, J.

    2001-01-01

    Sequences such as a steam generator tube rupture (SGTR) with stuck-open relief valve represent a significant public risk by virtue of the open path for release of radioactivity. The release may be lessened by deposition of fission products on the steam generator (SG) tubes and other structures or by scrubbing in the secondary coolant. The absence of empirical data, the complexity of the geometry and controlling processes, however, make the retention difficult to quantify and credit for it is typically not taken in risk assessments. The ARTIST experimental program to be conducted at Paul Scherrer Institut, Switzerland, will simulate the flow and retention of aerosol-borne fission products in the SG secondary, and thus provide a unique database to support safety assessments and analytical models. The project, foreseen in seven phases, will study phenomena at the separate effect and integral levels, and also address accident management (AM) issues. The prescribed values of the controlling parameters (aerosol size, aerosol type, gas flow velocity, residence time, etc) cover the range expected in severe and design basis accident scenarios. (authors)

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

    International Nuclear Information System (INIS)

    Pitner, P.; Riffard, T.

    1993-09-01

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

  18. Repair technology for steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

    2001-02-01

    The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating.

  19. Repair technology for steam generator tubes

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

    2001-02-01

    The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating

  20. 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)

  1. Steam generator tubing development for commercial fast breeder reactors

    International Nuclear Information System (INIS)

    Sessions, C.E.; Uber, C.F.

    1981-01-01

    The development work to design, manufacture, and evaluate pre-stressed double-wall 2/one quarter/ Cr-1 Mo steel tubing for commercial fast breeder reactor steam generator application is discussed. The Westinghouse plan for qualifying tubing vendors to produce this tubing is described. The results achieved to date show that a long length pre-stressed double-wall tube is both feasible and commercially available. The evaluation included structural analysis and experimental measurement of the pre-stress within tubes, as well as dimensional, metallurgical, and interface wear tests of tube samples produced. This work is summarized and found to meet the steam generator design requirements. 10 refs

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

    Science.gov (United States)

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

    2006-09-01

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

  3. Neutron generator tube ion source control apparatus

    International Nuclear Information System (INIS)

    Bridges, J.R.

    1982-01-01

    A pulsed neutron well logging system includes a neutron generator tube of the deuterium-tritium accelerator type and an ion source control apparatus providing extremely sharply time-defined neutron pulses. A low voltage control pulse supplied to an input by timing circuits turns a power FET on via a buffer-driver whereby a 2000 volt pulse is produced in the secondary of a pulse transformer and applied to the ion source of the tube. A rapid fall in this ion source control pulse is ensured by a quenching circuit wherein a one-shot responds to the falling edge of the control pulse and produces a 3 microsecond delay to compensate for the propagation delay. A second one-shot is triggered by the falling edge of the output of the first one-shot and gives an 8 microsecond pulse to turn on the power FET which, via an isolation transformer turns on a series-connected transistor to ground the secondary of the pulse transformer and the ion source. (author)

  4. 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

  5. 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

  6. ARTIST: An International Project Investigating Aerosol Retention in a Ruptured Steam Generator

    International Nuclear Information System (INIS)

    Guentay, S.; Dehbi, A.; Suckow, D.; Birchley, J.

    2002-01-01

    Steam generator tube ruptures (SGTR) with a concurrent stuck open safety relief valve are counted among the risk dominant accident sequences because of the potential for radioactive products to bypass the containment. Owing to the absence of relevant empirical data and the complexity of the geometry and controlling processes, the aerosol removal in the steam generator (SG) tubes and in the secondary side is not well understood. Therefore, little or no credit is usually taken for aerosol retention due to natural processes in the various components of a SG. To help reduce the uncertainties associated with fission product release following an SGTR sequence, the Paul Scherrer Institut has initiated an international experimental project to be performed in the ARTIST (AeRosol Trapping In a Steam generaTor) facility in the time period from 2002 to 2007. The ARTIST test section is a scaled model of a real SG, and is comprised of a 264-tube bundle with a maximum height of 3.8 m, as well as one full-size droplet separator and one full-size steam dryer. The ARTIST facility is capable of producing soluble and insoluble aerosols and entrain them at sonic gas flow rates (up to 0.25 kg/s, thus matching comparable values predicted by the codes. In addition, aerosols can be generated at prototypical concentrations (up to 5 g/m 3 ) and sizes (0.2-5 mm AMMD). State of the art instrumentation is used (Low-pressure impactors, photometers, on-line particle sizer, online droplet sizer, etc.). The ARTIST project will simulate the flow and retention of aerosol-borne fission products in the SG, and provide a unique database to support safety assessments and analytical models. The project is foreseen in seven phases: 1) Aerosol retention in the tube under dry secondary side conditions, 2) Aerosol retention in the near field close to break under dry conditions, 3) Aerosol retention in the bundle far field under dry conditions, 4) Aerosol retention in the separator and dryer under dry

  7. 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

  8. Failure of fretted steam generator tubes under accident conditions

    International Nuclear Information System (INIS)

    Forrest, C.F.

    1996-10-01

    Tests were carried out with a bank of tubes in a water tunnel to determine the tolerance of flawed nuclear reactor steam generator tubes to accident conditions which would result in high cross-flow velocities. Fourteen specimen tubes were tested, each having one or two types of defect machined into the surface simulating fretting-wear type scars found in some operating steam generators. The tubes were tested at flow velocities sufficient to induce high fluid elastic-type vibrations. Seven of the tubes failed near the thinnest section of the defects during the one-hour tests, due to impacting and/or rubbing between the tube and the support. Strain gauges, displacement transducers, force gauges and an accelerometer were used on the target tube and/or the tube immediately downstream of it to measure their vibrational characteristics

  9. Status of steam generator tubing integrity at Jaslovske Bohunice NPP

    International Nuclear Information System (INIS)

    Cepcek, S.

    1997-01-01

    Steam generator represents one of the most important component of nuclear power plants. Especially, loss of tubing integrity of steam generators can lead to the primary coolant leak to secondary circuit and in worse cases to the unit shut down or to the PTS events occurrence. Therefore, to ensure the steam generator tubing integrity and the current knowledge about tube degradation propagation and development is of the highest importance. In this paper the present status of steam generator tubing integrity in operated NPP in Slovak Republic is presented

  10. Extending service life of steam generators by sleeving tubes

    International Nuclear Information System (INIS)

    Gutzwiller, J.E.

    1982-01-01

    Steam generator tubes that are failing due to IGA in the tubesheet crevice can be kept in service by using the basic sealable sleeve design developed by BandW. Variations of the present sleeve design could significantly reduce the number of tubes that must be plugged each year. Sleeving had the potential of keeping 28 percent more tubes in service during 1979. Lowering the overall rate at which tubes are removed from service by plugging will reduce the probability of having to derate the plant or replace the steam generator. Considering tomorrow's replacement power costs, sleeving to keep tubes in service is a practical and sound investment

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

    International Nuclear Information System (INIS)

    Dussarte, D.

    1991-11-01

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

  12. 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)

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

    Directory of Open Access Journals (Sweden)

    Konstantinos Blouhos

    2013-01-01

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

  14. Experimental study of tube/support impact forces in multi-span PWR steam generator tubes

    International Nuclear Information System (INIS)

    Axisa, F.; Desseaux, A.; Gibert, R.J.

    1984-12-01

    The vibro-impact response of a straight part of a steam generator tube is investigated experimentally and using numerical simulation with the aim to relate tube overall dynamics with excitation and tube-support clearance. Configuration studied here corresponds to the tube being excited in only one direction at its first resonance presenting an antinode of vibration at the impacted support. Tests show namely that midspan displacement of tube is almost proportional to excitation level and clearance. Impact forces averaged over a cycle of vibration are almost proportional to excitation and poorly dependent on clearance. Results of numerical simulation are in fairly good agreement with test results

  15. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

    Corrosion resistance is the major consideration in selecting tubing material for CANDU steam generators. Corrosion, and additional considerations, lead to the following steam generator tubing material recommendations: for CANDU-BPHWR's (boiling pressurized heavy water reactors) low-cobalt Incoloy-800; for CANDU-PHWR's (pressurized, non-boiling, heavy water reactors), low-cobalt Monel-400

  16. 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)

  17. Inspection and repair of steam generator tubing with a robot

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1985-01-01

    During inspection and repair of steam generator tubing, radiation exposure to personnel is an unrequested endowment. To combat this intrinsic handicap, a robot has been designed for deployment in all operations inside the steam generator water chamber. This measure drastically reduces entering time and also improves inspection capabilities with regard to the accuracy and reproduction of the desired tube address. The inherent flexibility of the robot allows for performing various inspection and repair techniques: eddy-current testing of tubing; ultrasonic testing of tubing; visual examination of tube ends; profilometry measurements; tube plugging; plug removal; tube extraction; sleeving of tubes; tube end repair; chemical cleaning; and thermal treatment. Plant experience has highlighted the following features of the robot: 1) short installation and demounting periods; 2) installation independent of manhole location; 3) installation possible from outside the steam generator; 4) only one relocation required to address all the tube positions; 5) fast and highly accurate positioning; 6) operational surveillance not required; and 7) drastic reduction of radiation exposure to personnel during repair work

  18. Multifrequency eddy current testing of helical tubes of steam generators

    International Nuclear Information System (INIS)

    Pigeon, M.; David, B.

    1983-06-01

    In the event of a water-sodium reaction in a steam-generator of a fast breeder reactor, it is necessary to test the tubes close to the leak to evaluate the damage. In SUPERPHENIX, the tubes are about 100m long and are coiled on a dead body. This report describes the equipment and the technic to test such tubes with multifrequency eddy current technics [fr

  19. Vibration and wear prediction for steam generator tubes: Final report

    International Nuclear Information System (INIS)

    Rao, M.S.M.; Gupta, G.D.; Eisinger, F.L.

    1988-06-01

    As part of the overall EPRI program to develop a mechanistic model for tube fretting and wear prediction, Foster Wheeler Development Corporation undertook the responsibility of developing analytical models to predict structural response and wear in a multispan tube. The project objective was to develop the analytical capability to simulate the time-dependent motion of a multispan steam generator tube in the presence of the clearance gaps at each tube baffle or support. The models developed were to simulate nonlinear tube-to-tube support interaction by determining the impact force, the sliding distance, and the resultant tube wear. Other objectives of the project included: validate the models by comparing the analytical results with the EPRI tests done at Combustion Engineering (C-E) on single multispan tubes; test the models for simulating the U-bend region of the steam generator tube, including the antivibration bars; and develop simplified methods to treat the nonlinear dynamic problem of a multispan tube so that computing costs could be minimized. 15 refs., 53 figs., 27 tabs

  20. Wastage of Steam Generator Tubes by Sodium-Water Reaction

    International Nuclear Information System (INIS)

    Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

    2010-01-01

    The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown in Figure 1. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. For this, multi-target wastage tests for modified 9Cr-1Mo steel tube bundle by intermediate leaks are being prepared

  1. Modeling and Simulation of U-tube Steam Generator

    Science.gov (United States)

    Zhang, Mingming; Fu, Zhongguang; Li, Jinyao; Wang, Mingfei

    2018-03-01

    The U-tube natural circulation steam generator was mainly researched with modeling and simulation in this article. The research is based on simuworks system simulation software platform. By analyzing the structural characteristics and the operating principle of U-tube steam generator, there are 14 control volumes in the model, including primary side, secondary side, down channel and steam plenum, etc. The model depends completely on conservation laws, and it is applied to make some simulation tests. The results show that the model is capable of simulating properly the dynamic response of U-tube steam generator.

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

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2000-01-01

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

  3. Vibration and wear characteristics of steam generator tubes

    International Nuclear Information System (INIS)

    Choi, Young Hwan

    2003-06-01

    This study investigates the fluid elastic instability characteristics of Steam Generator (SG) U-tubes with defect and the safety assessment of the potential for fretting-wear damages on Steam Generator (SG) U-tubes caused by foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions for determining the fluid elastic instability or fretting-wear parameters such as damping ratio, added mass and flow velocity are obtained from three-dimensional SG flow calculation using the ATHOS3 code. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed is the effect of the internal pressure on the vibration and fretting-wear characteristics of the tube

  4. Maintenance and plugging technology for CANDU steam generator tubing

    International Nuclear Information System (INIS)

    Prince, J.; Nicholson, A.; Hare, J.; McGoey, L.; Stafford, T.; Gowthorpe, P.

    2006-01-01

    In order to keep aging steam generators in service and to successfully manage the life of these critical components, the capability must exist to perform tube plugging and other complex maintenance activities in-situ. In the early days of CANDU steam generator operation, the only option was to perform these activities manually, which had inherent safety and quality risks. The challenge was to be able to perform these activities remotely thus eliminating some of the confined space and radiological exposure risks. The additional challenge was to develop equipment and techniques which would result in significantly improved quality, particularly for the completed plug welds which would be returned to service. Over the past fifteen years, this technology has matured and has produced remarkable results in field application. Some 14000 tube plugs have been successfully installed to date using automated plugging techniques. This paper presents an overview of the development of techniques available to utilities for steam generator tube plugging as well as some highlights of other steam generator tube maintenance activities such as primary side tube removal and tube end damage repair. Aspects covered in the paper include plug and procedure development, automated equipment and manipulators for tool deployment, process controls and personnel requirements. Recently, the steam generator tube plugging performed by OPG has been incorporated into a formal quality program under the requirements of ASME NCA 4000. An overview of the quality program will be presented and details of some of the important aspects of the quality program will be discussed. (author)

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

    International Nuclear Information System (INIS)

    Nagato, Kotaro; Satoh, Keisuke

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  7. Set-up for steam generator tube bundle washing after explosion expanding the tubes

    International Nuclear Information System (INIS)

    Osipov, S.I.; Kal'nin, A.Ya.; Mazanenko, M.F.

    1985-01-01

    Set-up for steam generator tube bundle washing after the explosion expanding of tubes is described. Washing is accomplished by distillate. Steam is added to distillate for heating, and compersed air for preventing hydraulic shock. The set-up is equiped by control equipment. Set-up performances are presented. Time for one steam generator washing constitutes 8-12 h. High economic efficiency is realized due to the set-up introduction

  8. Study of tritium permeation through Peach Bottom Steam Generator tubes

    International Nuclear Information System (INIS)

    Yang, L.; Baugh, W.A.; Baldwin, N.L.

    1977-06-01

    The report describes the equipment developed, samples tested, procedures used, and results obtained in the tritium permeation tests conducted on steam generator tubing samples which were removed from the Peach Bottom Unit No. 1 reactor

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Schaefer, L.; Kempe, H.

    1983-12-01

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

  12. Steam generator tube degradation at the Doel 4 plant influence on plant operation and safety

    International Nuclear Information System (INIS)

    Scheveneels, G.

    1997-01-01

    The steam generator tubes of Doel 4 are affected by a multitude of corrosion phenomena. Some of them have been very difficult to manage because of their extremely fast evolution, non linear evolution behavior or difficult detectability and/or measurability. The exceptional corrosion behavior of the steam generator tubes has had its drawbacks on plant operation and safety. Extensive inspection and repair campaigns have been necessary and have largely increased outage times and radiation exposure to personnel. Although considerable effort was invested by the utility to control corrosion problems, non anticipated phenomena and/or evolution have jeopardized plant safety. The extensive plugging and repairs performed on the steam generators have necessitated continual review of the design basis safety studies and the adaptation of the protection system setpoints. The large asymmetric plugging has further complicated these reviews. During the years many preventive and recently also defence measures have been implemented by the utility to manage corrosion and to decrease the probability and consequences of single or multiple tube rupture. The present state of the Doel 4 steam generators remains troublesome and further examinations are performed to evaluate if continued operation until June '96, when the steam generators will be replaced, is justified

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  14. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gorman, J.A.; Turner, A.P.L.

    1997-01-01

    This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

  15. Steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Gorman, J.A.; Harris, J.E.; Lowenstein, D.B.

    1995-07-01

    The objectives of this project were to characterize defect mechanisms which could affect the integrity of steam generator tubes, to review and critique state-of-the-art Canadian and international steam generator tube fitness-for-service criteria and guidelines, and to obtain recommendations for criteria that could be used to assess fitness-for service guidelines for steam generator tubes containing defects in Canadian power plant service. Degradation mechanisms, that could affect CANDU steam generator tubes in Canada, have been characterized. The design standards and safety criteria that apply to steam generator tubing in nuclear power plant service in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA have been reviewed and described. The fitness-for-service guidelines used for a variety of specific defect types in Canada and internationally have been evaluated and described in detail in order to highlight the considerations involved in developing such defect specific guidelines. Existing procedures for defect assessment and disposition have been identified, including inspection and examination practices. The approaches used in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA for fitness-for-service guidelines were compared and contrasted for a variety of defect mechanisms. The strengths and weaknesses of the various approaches have been assessed. The report presents recommendations on approaches that may be adopted in the development of fitness-for-service guidelines for use in the dispositioning of steam generator tubing defects in Canada. (author). 175 refs., 2 tabs., 28 figs

  16. 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

  17. WWER Steam Generators Tubing Performance and Aging Management

    International Nuclear Information System (INIS)

    Trunov, Nikolay B.; Davidenko, Stanislav E.; Grigoriev, Vladimir A.; Popadchuk, Valery S.; Brykov, Sergery I.; Karzov, Georgy P.

    2006-01-01

    At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned. (authors)

  18. Wolsong 3 and 4 steam generator tube inspection

    International Nuclear Information System (INIS)

    Jang, Kyoung Sik; Son, Tai Bong; Kwon, Dong Ki; Choi, Jin Hyuk

    2001-01-01

    During the pre-service inspection for Wolsong unit 3 and 4 in 1997/1998 respectively, 17 distorted roll transition indications (over expanded beyond tubesheet secondary face) were identified at the unit 4 (S/G B, D). Six(6) tubes out of these tubes were plugged in 1998. However the first periodic inspection identified additional 110 indications in 1999 and 2000. The additionally identified 110 indication call, not reported at the pre-service inspection, are; 2 not-finally-expanded-tubes and 108 distorted roll transition tubes. Design limit of each steam generator tube plugging is 6.4.%. Plugging was performed by the steam generator manufacturer under the warranty. When distorted roll transition indications were first identified on the unit 4 in 1998 the degree of over-expansion was measured using an inner dial-gage to make the disposition of nonconformance report. 2 Not-finally-expanded-tubes were plugged and 10 tubes out of 108 distorted roll transition tubes were also plugged as a preventive measure

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  20. Development of the double-wall-tube steam generator. Evaluation of inner tube leak detection system

    International Nuclear Information System (INIS)

    Teraoku, Takuji; Kisohara, Naoyuki

    1995-01-01

    A double-wall-tube steam generator (DWT-SG) is considered to have possibility of eliminating a secondary heat transport system to realize a reliable and simplified FBR plant. Thus, basic tests for inner/outer tube leak detection and prototypical leak tests by use of the 1MWt DWT-SG model have been performed to evaluate the feasibility of DWT-SG. Their results demonstrated that the inner leak detection system can definitely detect a steam leak from an inner tube flaw. Analyses of the inner tube leak and detection behavior obtained in the 1MWt DWT-SG test enabled to estimate the performance of the inner tube detection system of the commercial DWT-SG system. (author)

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  3. 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

  4. Steam generator tube integrity program: Phase II, Final report

    Energy Technology Data Exchange (ETDEWEB)

    Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

    1988-08-01

    The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted.

  5. Steam generator tube integrity program: Phase II, Final report

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

    1988-08-01

    The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  7. Development of weld plugging for steam generator tubes of FBR

    International Nuclear Information System (INIS)

    Shimoyama, T.; Matsuyama, T.; Matsumoto, O.; Nagura, Y.; Nakamura, H.; Tohguchi, Y.; Kurokawa, M.; Fukada, T.

    2002-01-01

    This study was undertaken to develop a method of weld plugging of the heat-exchanger tubes of steam generator of Prototype FBR 'MONJU' in case these tubes are damaged for some reason. We studied mainly the shape of plug, welding procedure and effect of postweld heat treatment (PWHT). Evaporator tube sheet, tube and plug are made of 2-1/4Cr-1Mo steel and usually preheating and PWHT will be required for welding of this steel. The results of this study is as follows. 1) Plug was designed to make butt joint welding with grooved tube sheet around the tube hole to satisfy the requirements of plug designing, stress analysis, and good weldability. 2) TIG welding process was selected and certified its good weldability and good performance. 3) PWHT can be done by using high frequency induction heating method locally and also designing the plug to weld joint with tube sheet which was grooved around the tube hole. 4) Mock up test was done and it was certified that this plugging procedure has good weldability and good performance ability for Non Destructive Inspection. (author)

  8. Dynamic characteristics of steam generator U-tubes with defect

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

    2005-01-01

    This study investigates the fluid elastic instability characteristics of steam generator (SG) U-tubes with defect and the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for the U-tubes either with axial or circumferential flaw with different sizes. Special emphases are on the effects of flaw orientation and size on the modal and instability characteristics of tubes, which are expressed in terms of the natural frequency, corresponding mode shape and stability ratio. Also, the wear rate of U-tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted, and discussed in this study is the effect of the flow velocity and vibration of the tube on the remaining life of the tube. In addition, addressed in this study is the effect of the internal pressure on the vibration and fretting-wear characteristics of the tube

  9. Effect of tube plugging in the thermalhydraulic performance of 'U' tube steam generators

    International Nuclear Information System (INIS)

    Braga, C.V.M.; Carajilescov, P.

    1981-05-01

    The thermalhydraulic performance of Angra II steam generator has been simulated using the model developed by Braga, C.V.M., 'Thermohydraulic model for steam generator of PWR power plants', in steady state, with plugging up to 40% of total number of tubes. (E.G.) [pt

  10. Steam generator tube integrity program. Phase I report

    International Nuclear Information System (INIS)

    Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

    1979-09-01

    The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators

  11. PROFIL-360 high resolution steam generator tube profilometry system

    International Nuclear Information System (INIS)

    Glass, S.W.

    1985-01-01

    A high-resolution profilometry system, PROFIL 360, has been developed to assess the condition of steam generator tubes and rapidly produce the data to evaluate the potential for developing in-service leaks. The probe has an electromechanical sensor in a rotating head. This technique has been demonstrated in the field, saving tubes that would have been plugged with the go-gauge criterion and indicating plugging other high-risk candidates that might otherwise not have been removed from service

  12. Profil-360 high resolution steam generator tube profilometry system

    International Nuclear Information System (INIS)

    Glass, S.W.

    1985-01-01

    A high-resolution profilometry system, PROFIL 360, has been developed to assess the condition of steam generator tubes and rapidly produce the data to evaluate the potential for developing in-service leaks. The probe has an electromechanical sensor in a rotating head. This technique has been demonstrated in the field, saving tubes that would have been plugged with the go-gauge criterion and indicating plugging other high-risk candidates that might otherwise not have been removed from service

  13. Performance demonstration requirements for eddy current steam generator tube inspection

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Heasler, P.G.; Anderson, C.M.

    1992-10-01

    This paper describes the methodology used for developing performance demonstration tests for steam generator tube eddy current (ET) inspection systems. The methodology is based on statistical design principles. Implementation of a performance demonstration test based on these design principles will help to ensure that field inspection systems have a high probability of detecting and correctly sizing tube degradation. The technical basis for the ET system performance thresholds is presented. Probability of detection and flaw sizing tests are described

  14. An advanced tube wear and fatigue workstation to predict flow induced vibrations of steam generator tubes

    International Nuclear Information System (INIS)

    Gay, N.; Baratte, C.; Flesch, B.

    1997-01-01

    Flow induced tube vibration damage is a major concern for designers and operators of nuclear power plant steam generators (SG). The operating flow-induced vibrational behaviour has to be estimated accurately to allow a precise evaluation of the new safety margins in order to optimize the maintenance policy. For this purpose, an industrial 'Tube Wear and Fatigue Workstation', called 'GEVIBUS Workstation' and based on an advanced methodology for predictive analysis of flow-induced vibration of tube bundles subject to cross-flow has been developed at Electricite de France. The GEVIBUS Workstation is an interactive processor linking modules as: thermalhydraulic computation, parametric finite element builder, interface between finite element model, thermalhydraulic code and vibratory response computations, refining modelling of fluid-elastic and random forces, linear and non-linear dynamic response and the coupled fluid-structure system, evaluation of tube damage due to fatigue and wear, graphical outputs. Two practical applications are also presented in the paper; the first simulation refers to an experimental set-up consisting of a straight tube bundle subject to water cross-flow, while the second one deals with an industrial configuration which has been observed in some operating steam generators i.e., top tube support plate degradation. In the first case the GEVIBUS predictions in terms of tube displacement time histories and phase planes have been found in very good agreement with experiment. In the second application the GEVIBUS computation showed that a tube with localized degradation is much more stable than a tube located in an extended degradation zone. Important conclusions are also drawn concerning maintenance. (author)

  15. Anatomy Education for the YouTube Generation

    Science.gov (United States)

    Barry, Denis S.; Marzouk, Fadi; Chulak-Oglu, Kyrylo; Bennett, Deirdre; Tierney, Paul; O'Keeffe, Gerard W.

    2016-01-01

    Anatomy remains a cornerstone of medical education despite challenges that have seen a significant reduction in contact hours over recent decades; however, the rise of the "YouTube Generation" or "Generation Connected" (Gen C), offers new possibilities for anatomy education. Gen C, which consists of 80% Millennials, actively…

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-25

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

  17. 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.

    2001-01-01

    Industry effects have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, 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 are not accurate enough to permit continued operation. 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

  18. Reliability of eddy current examination of steam generator tubes

    International Nuclear Information System (INIS)

    Birks, A.S.; Ferris, R.H.; Doctor, P.G.; Clark, R.A.; Spanner, G.E.

    1985-04-01

    A unique study of nondestructive examination reliability is underway at the Pacific Northwest Laboratory under US Nuclear Regulatory Commission sponsorship. Project participants include the Electric Power Research Institute and consortiums from France, Italy, and Japan. This study group has conducted a series of NDE examinations of tubes from a retired-from-service steam generator, using commercially available multifrequency eddy current equipment and ASME procedures. The examination results have been analyzed to identify factors contributing to variations in NDE inspection findings. The reliability of these examinations will then be validated by destructive analyses of the steam generator tubes. The program is expected to contribute to development of a model for steam generator inservice inspection sampling plans and inspection periods, as well as to improved regulatory guidelines for tube plugging

  19. Fatigue analysis of a PWR steam generator tube sheet

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  20. Decontamination of Steam Generator tube using Abrasive Blasting Technology

    International Nuclear Information System (INIS)

    Min, B. Y.; Kim, G. N.; Choi, W. K.; Lee, K. W.; Kim, D. H.; Kim, K. H.; Kim, B. T.

    2010-01-01

    As a part of a technology development of volume reduction and self disposal for large metal waste project, We at KAERI and our Sunkwang Atomic Energy Safety (KAES) subcontractor colleagues are demonstrating radioactively contaminated steam generator tube by abrasive blasting technology at Kori-1 NPP. A steam generator is a crucial component in a PWR (pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary waste-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tube, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be cause of tube leakage, more and more steam generators are replaced today. Only in Korea, already 2 of them are replaced and will be replaced in the near future. The retired 300 ton heavy Steam generator was stored at the storage waste building of Kori NPP site. The steam generator waste has a large volume, so that it is necessary to reduce its volume by decontamination. A waste reduction effect can be obtained through decontamination of the inner surface of a steam generator. Therefore, it is necessary to develop an optimum method for decontamination of the inner surface of bundle tubes. The dry abrasive blasting is a very interesting technology for the realization of three-dimensional microstructures in brittle materials like glass or silicon. Dry abrasive blasting is applicable to most surface materials except those that might be shattered by the abrasive. It is most effective on flat surface and because the abrasive is sprayed and can also applicable on 'hard to reach' areas such as inner tube ceilings or behind equipment. Abrasive decontamination techniques have been applied in several countries, including Belgium, the CIS, France, Germany, Japan, the UK and the USA

  1. 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

  2. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

  3. 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)

  4. Recent developments in plugging of steam generator tubes

    International Nuclear Information System (INIS)

    Buhay, S.; Abucay, R.C.

    1995-01-01

    Mechanical Plugging capability has been developed for Bruce Nuclear Generating Station (BNGS) steam generator (SG) tubes and Darlington Nuclear Generating Station (DNGS) SG tubes and tubesheet holes. The plug concept was a modified ABB/Combustion Engineering Inconel 690 plug with a nickel band, rolled into the tube or tubesheet hole from the primary side of the tubesheet. The qualification program included analytical justification of the plug body and experimental testing to verify the leak tightness of the rolled joint under conditions which meet or exceed all service or design requirements. Tools and procedures were developed and tested for manual and remote/robotic installation and removal of the mechanical plugs. Additionally, tools and procedures were developed to plug tubes/tubesheet holes at DNGS in the event the steam generator is recalled to service to act as a heat sink. A crew of Ontario Hydro personnel were trained and qualified for the installation of mechanical plugs for permanent and recall applications. During the DNGS Unit 4 spring 1995 outage, 6 tubes were plugged and the 'Recall Plugging Capability' was deployed and ready for use during a primary side SG tube removal. The mechanical plugs were installed manually with a typical 3 minute/plug in-bowl duration time with an average radiation dose of 12.5 mrem per plug. This compares favourably with manual plug welding during the same outage in the same SG bowl at approximately 15-30 minutes/plug in-bowl duration with an average radiation dose of 117 mrem/plug. (author)

  5. 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)

  6. 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

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

    International Nuclear Information System (INIS)

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

    1999-12-01

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

  8. Nuclear steam generator tube to tubesheet joint optimization

    International Nuclear Information System (INIS)

    McGregor, Rod

    1999-01-01

    Industry-wide problems with Stress Corrosion Cracking in the Nuclear Steam Generator tube-to-tubesheet joint have led to costly repairs, plugging, and replacement of entire vessels. To improve corrosion resistance, new and replacement Steam Generator developments typically employ the hydraulic tube expansion process (full depth) to minimize tensile residual stresses and cold work at the critical transition zone between the expanded and unexpanded tube. These variables have undergone detailed study using specialized X-ray diffraction and analytical techniques. Responding to increased demands from Nuclear Steam Generator operators and manufacturers to credit the leak-tightness and strength contributions of the hydraulic expansion, various experimental tasks with complimentary analytical modelling were applied to improve understanding and control of tube to hole contact pressure. With careful consideration to residual stress impact, design for strength/leak tightness optimization addresses: Experimentally determined minimum contact pressure levels necessary to preclude incipient leakage into the tube/hole interface. The degradation of contact pressure at surrounding expansions caused by the sequential expansion process. The transient and permanent contact pressure variation associated with tubesheet hole dilation during Steam Generator operation. An experimental/analytical simulation has been developed to reproduce cyclic Steam Generator operating strains on the tubesheet and expanded joint. Leak tightness and pullout tests were performed during and following simulated Steam Generator operating transients. The overall development has provided a comprehensive understanding of the fabrication and in-service mechanics of hydraulically expanded joints. Based on this, the hydraulic expansion process can be optimized with respect to critical residual stress/cold work and the strength/leakage barrier criteria. (author)

  9. Modeling and dynamic simulation of U-tube steam generator

    International Nuclear Information System (INIS)

    Cui Zhenghua; Jia Dounan; Chen Xuejun; Yu Erjun

    1992-01-01

    An accurate and simple dynamic mathematical model of U-tube steam generator is presented. It is solved by Adams method and Gear method respectively. The results of this model are in good agreements with that of Kerlin's model which has been validated by the tests. And the two calculating methods are compared

  10. GRUVAL for ET inspection of the steam generator tubes

    International Nuclear Information System (INIS)

    Garcia Bueno, A.; Francia, L.; Jimenez Garcia, J. J.; Garcia, R.; Castelinou, M.; Torrens, J.

    2013-01-01

    The steam generators of the nuclear power plants, PWR type are one of the most important components from the point of view of safety and plant availability. Thousands of tubes that form, approximately 1 mm of thickness, required to be inspected in accordance with codes and standards, to ensure the integrity of the component during the operation of the plant.

  11. Heat exchanger tubing materials for CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Taylor, G.F.

    1977-07-01

    The performance of steam generator tubing (nickel-chromium-iron alloy in NPD and nickel-copper alloy in Douglas Point and Pickering generating stations) has been outstanding and no corrosion-induced failures have occurred. The primary coolant will be allowed to boil in the 600 MW (electrical) CANDU-PHW reactors. An iron-nickel-chromium alloy has been selected for the steam generator tubing because it will result in lower radiation fields than the alloys used before. It is also more resistant than nickel-chromium-iron alloy to stress corrosion cracking in the high purity water of the primary circuit, an unlikely but conceivable hazard associated with higher operating temperatures. Austenitic alloy and ferritic-austenitic stainless steel tubing have been selected for the moderator coolers in CANDU reactors being designed and under construction. These materials will reduce the radiation fields around the moderator circuit while retaining the good resistance to corrosion in service water that has characterized the copper-nickel alloys now in use. Brass and bronze tubes in feedwater heaters and condensers have given satisfactory service but do, however, complicate corrosion control in the steam cycle and, to reduce the transport of corrosion products from the feedtrain to the steam generator, stainless steel is preferred for feedwater heaters and stainlss steel or titanium for condensers. (author)

  12. Probabilistic methodology for assessing steam generator tube inspection - Phase II: CANTIA - a probabilistic method for assessing steam generator tube inspections

    International Nuclear Information System (INIS)

    Harris, J.E.; Gorman, J.A.; Turner, A.P.L.

    1997-03-01

    The objectives of this project were to develop a computer-based method for probabilistic assessment of inspection strategies for steam generator tubes, and to document the source code and to provide a user's manual for it. The program CANTIA was created to fulfill this objective, and the documentation and verification of the code is provided in this volume. The user's manual for CANTIA is provided as a separate report. CANTIA uses Monte Carlo techniques to determine approximate probabilities of steam generator tube failures under accident conditions and primary-to-secondary leak rates under normal and accident conditions at future points in time. The program also determines approximate future flaw distributions and non-destructive examination results from the input data. The probabilities of failure and leak rates and the future flaw distributions can be influenced by performing inspections of the steam generator tubes at some future points in time, and removing defective tubes from the population. The effect of different inspection and maintenance strategies can therefore be determined as a direct effect on the probability of tube failure and primary-to-secondary leak rate

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

  14. Repairing a steam generator tube by inserting a sheath

    International Nuclear Information System (INIS)

    Gaudin, J.P.

    1986-01-01

    According to the invention, the mechanical deformation of the sheath, realized by expansion in its end part opposite to the expanded end within the tube plate, is situated along a limited height, and the parameters of the said mechanical deformation are calculated according to the welding parameters applied consecutively. Besides, the said welding parameters are determined according to the initial mechanical deformation to obtain stress relaxation more particularly in the singular zones of the mechanical deformation. The present invention applies to the repair of PWR steam generator tubes [fr

  15. Flow-induced decentering and tube support interaction for steam generator tubes: experiment and physical interpretation

    International Nuclear Information System (INIS)

    Gay, N.; Granger, S.

    1992-11-01

    Maintaining PWR components under reliable operating conditions requires a complex design to prevent various damaging processes including flow-induced vibration and wear mechanisms. To improve the prediction of tube/support interaction and wear in PWR components, EDF has undertaken a comprehensive program oriented to both experimental and computational studies. The present paper illustrates one aspect of this program, related to the determination of contact forces between steam generator tubes and anti-vibration bars (AVBs). The dynamic, nonlinear behavior of a U-tube excited by an air cross-flow is investigated on the CLAVECIN experiment. Interesting and rather unexpected results have been obtained, by varying clearances and flow velocities. The paper is focused on four main points: (i) the originality of the experiment with a force measurement device located in flow; (ii) the importance of a refined data processing for accurately measuring contact forces; (iii) the presentation of the unexpected phenomena revealed in the CLAVECIN experiment, i.e. a flow-induced decentering of the tube which changed the initial tube/AVB clearance, and the consequences on tube/support interaction; (iv) the influence of the actual tube/support clearance in flow on wear mechanisms. The work, presented in the second part of this paper, concentrates exclusively on the physical interpretation of the flow-induced decentering phenomenon and on the theoretical analysis of its consequences on dynamic tube/support interaction. We show that the flow-induced decentering phenomenon can be generated by an unstable quasi-static coupling between the flexible tube and the confined flow, in the vicinity of the support system. This phenomenon is not specific to the CLAVECIN tests and it can be expected every time that a movable obstacle is subjected to confined flow. Moreover, in single-sided impacting conditions, the theoretical analysis confirms the linear relation, found in the CLAVECIN tests

  16. Analysis of steam generator tube sections removed from Gentilly-2 nuclear generating station

    International Nuclear Information System (INIS)

    Semmler, J.; Lockley, A.J.; Doyon, D.

    2010-01-01

    In order to meet the requirements of CSA Standards CAN/CSA N285.4-94, which states, 'A section of one tube in a deposit region shall be removed from one steam generator for metallurgical examination', Gentilly-2 has been removing steam generator tube sections on a regular basis for analysis at Chalk River Laboratories. In 2009 April, sections from the hot leg and the cold leg of a steam generator tube were removed for detailed metallographic examination and characterization. The hot leg tube section covered the area from within the tube sheet up to below the second support plate, and the cold leg tube section covered the area from within the tube sheet to below the third preheater support plate. After a general visual and photographic examination, the area above the tube sheet on the hot leg side where the sludge pile is highest was examined in detail. Visual and macro-photography of the two tube sections within the tube sheet were also examined. Additional metallographic and surface examinations of both tube inner diameter and tube outer diameter, and surface roughness measurements of tube inner diameter were also completed. The surface activities (μCi/cm 2 ) of cold leg and hot leg specimens were measured before and after electrolytic descaling, and major and minor radionuclides were identified; a comparison of the surface activities for hot leg with the values for the cold leg were made. The results from the initial γ-spectroscopy measurements, and the measurements after the descaling of the specimens were used to estimate decontamination factors for each specimen and for each radionuclide. The tube specimens had thin outer diameter oxides; all four steam generators were chemically cleaned in 2005. All specimens had inner diameter deposits; the inner diameter deposits on the cold leg were heavier than those on the hot leg as expected. Primary side oxide loadings of specimens were used to estimate the total oxide inventory in 2009. The oxide

  17. The development and application of overheating failure model of FBR steam generator tubes. 3

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi; Wada, Yusaku; Miyakawa, Akira; Okabe, Ayao; Nakai, Ryodai; Hiroi, Hiroshi

    2002-03-01

    The model has been developed for the assessment of the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). The model has been applied to the Monju SG studies. Major results obtained in the studies are as follows: 1. To evaluate the structural integrity of tube material, the strength standard for 2. 25Cr-1Mo steel was established taking account of time dependent effect based on the high temperature (700-1200degC) creep data. This standard has been validated with the tube rupture simulation test data. 2. The conditions for overheating by the high temperature reaction were determined by use of the SWAT-3 experimental data. The realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed as the cosine-shaped temperature profile. 3. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the model. 4. The model has been validated with experimental data obtained by SWAT-3 and LLTR. The results were satisfactory with conservatism. The PFR superheater leak event in 1987 was studied, and the cause of event and the effectiveness of the improvement after the leak event could be identified by the analysis. 5. The model has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% water flow operating conditions when an initial leak is detected by the cover gas pressure detection system. (author)

  18. 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.)

  19. Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

    2016-12-15

    Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

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

    Science.gov (United States)

    Shingledecker, John

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

  1. Analysis of the State of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Bergunker, Olga

    2008-01-01

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

  2. 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)

  3. 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)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  5. 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)

  6. Stress analysis of steam generator row-1 tubes

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  7. Process and device for locating a defective tube, particularly in the tube bundle of a steam generator

    International Nuclear Information System (INIS)

    Denis, Jean.

    1977-01-01

    A process is described for locating a defective tube, particularly in the tube bundle of a steam generator of the reversed U tube kind with the ends connected to a tube plate, marking with the bottom of the generator casing a space separated into two adjacent collectors, respectively for the inlet and outlet of a primary fluid flowing inside the tubes of the bundle, these being externally washed by a secondary vaporizing fluid. In this process a television camera that can be inserted into the casing is used. This process consists in transmitting to a display system outside the generator an image of the tube plate in each collector by means of a directional television camera and then to place over this image a luminous marker to locate the end or the faulty tube [fr

  8. Probe for detection of denting in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Gerardin, J.P.; Germain, J.L.; Nio, J.C.

    1994-07-01

    In certain types of PWR steam generator, oxide deposits can lead to embedding, and subsequently to deformation of a tube (the phenomenon of ''denting''). Such embedding changes the vibratory behavior of the tubes and can result in fatigue cracking. This type of cracking can also be worsened in the event of improper assembly of the anti-vibration spacer bars supporting the U-bends. To prevent such incidents and provide for effective preventive condition-directed maintenance of its PWR steam generators, EDF has undertaken the study and development of a probe to detect this type of phenomenon. The studies began in 1990 and led to the building of an initial prototype probe. The principle behind the probe consists in inducing vibration in the U-bend and determining the main resonance modes of the tube. Measurements of frequency and amplitude and calculation of damping enable characterization of the mechanical behavior of the U-bend. The most important parameter is damping, for which the value must be sufficiently high to ensure that the tube is not subjected to major vibratory amplitudes during operation. Numerous tests have been performed with the first prototype version of the probe, on a mock-up in the test area and on one of the demounted steam generators on the Dampierre site. These different tests have enabled validation of the operating principle, fine-tuning the process, pinpointing certain mechanical problems in the probe design, and obtaining the first indications as to the real vibratory behavior of U-bends on a steam generator. On the basis of these preliminary tests, the specifications were drawn up for an industrial version of the probe. Following a call for bids and the choice of a manufacturer, work began on fabrication of a new probe model in 1993. This version was delivered at the end of 1993 and testing began in 1994. (authors). 5 figs., 2 tabs

  9. Detection of steam generator tube leaks in pressurized water reactors

    International Nuclear Information System (INIS)

    Roach, W.H.

    1984-11-01

    This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

  10. Stress relief treatment of Alloy 600 steam generator tubing

    International Nuclear Information System (INIS)

    Rooyen, D. van; Cragnolino, C.

    1994-01-01

    The intergranular stress corrosion cracking (IGSCC) of Alloy 600 tubing in the primary side of operating steam generators is the subject of this investigation. The objective of the program was to examine the feasibility of heat treatment to alleviate the IGSCC problem. In addition to this, tests were also performed to examine the IGSCC susceptibility of nuclear grade Alloy 600 tubing obtained from various sources. Examination of temperature-time combinations that may hold potential for improved IGSCC resistance of the transition regions of tubes expanded into tube sheet holes was done. The combinations fall in two categories. One is of short duration and relatively high temperature, where induction is the best method of heating because the treatment only lasts from some tens of seconds to a few minutes. The other is carried out in a lower temperature range and lasts for several hours. This latter combination of temperatures and times is considered for the so-called global heat treatment of entire tube sheet. To assess the effect of these treatments, reverse U-bend testing in high purity deaerated water containing an overpressure of hydrogen was employed and several heats of Alloy 600 were compared in tests at 365 degrees C, which is well above actual operating temperatures of steam generators, but provides an accelerated test procedure. Results of furnace heating in the range of 550-610 degrees C indicated improvement in IGSCC resistance, with best performance after a heat treatment at 610 degrees C for nine hours. In addition to stress relief, carbide precipitation can also occur, and their relative contributions to the improvement is discussed

  11. Steam Generator tube integrity -- US Nuclear Regulatory Commission perspective

    International Nuclear Information System (INIS)

    Murphy, E.L.; Sullivan, E.J.

    1997-01-01

    In the US, the current regulatory framework was developed in the 1970s when general wall thinning was the dominant degradation mechanism; and, as a result of changes in the forms of degradation being observed and improvements in inspection and tube repair technology, the regulatory framework needs to be updated. Operating experience indicates that the current U.S. requirements should be more stringent in some areas, while in other areas they are overly conservative. To date, this situation has been dealt with on a plant-specific basis in the US. However, the NRC staff is now developing a proposed steam generator rule as a generic framework for ensuring that the steam generator tubes are capable of performing their intended safety functions. This paper discusses the current U.S. regulatory framework for assuring steam generator (SG) tube integrity, the need to update this regulatory framework, the objectives of the new proposed rule, the US Nuclear Regulatory Commission (NRC) regulatory guide (RG) that will accompany the rule, how risk considerations affect the development of the new rule, and some outstanding issues relating to the rule that the NRC is still dealing with

  12. Anatomy education for the YouTube generation.

    Science.gov (United States)

    Barry, Denis S; Marzouk, Fadi; Chulak-Oglu, Kyrylo; Bennett, Deirdre; Tierney, Paul; O'Keeffe, Gerard W

    2016-01-01

    Anatomy remains a cornerstone of medical education despite challenges that have seen a significant reduction in contact hours over recent decades; however, the rise of the "YouTube Generation" or "Generation Connected" (Gen C), offers new possibilities for anatomy education. Gen C, which consists of 80% Millennials, actively interact with social media and integrate it into their education experience. Most are willing to merge their online presence with their degree programs by engaging with course materials and sharing their knowledge freely using these platforms. This integration of social media into undergraduate learning, and the attitudes and mindset of Gen C, who routinely creates and publishes blogs, podcasts, and videos online, has changed traditional learning approaches and the student/teacher relationship. To gauge this, second year undergraduate medical and radiation therapy students (n = 73) were surveyed regarding their use of online social media in relation to anatomy learning. The vast majority of students had employed web-based platforms to source information with 78% using YouTube as their primary source of anatomy-related video clips. These findings suggest that the academic anatomy community may find value in the integration of social media into blended learning approaches in anatomy programs. This will ensure continued connection with the YouTube generation of students while also allowing for academic and ethical oversight regarding the use of online video clips whose provenance may not otherwise be known. © 2015 American Association of Anatomists.

  13. Process of corrosion protection for a steam generator tube and device to apply it

    International Nuclear Information System (INIS)

    Malagola, P.; Vassal, J.M.

    1985-01-01

    The steam generator tube is fixed by crimping in a tube plate; a metallic layer compatible with the tube material is electrodeposited on the inner side of the tube after its mounting in the tube plate, on both side of the plate face in contact with the water to be steamed, along a length approximately longer than the transition zone between the crimped part of the tube and the part which is not crimped. The external side of the tube can be also covered by a metallic layer before its mounting through the tube plate. The metallic layer can be nickel. The invention applies, more particularly, to PWR steam generators [fr

  14. Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

    International Nuclear Information System (INIS)

    Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

    1976-01-01

    Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

  15. Repair technique for steam generator tubes using electroforming

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Jeong, Hyun Kyu; Seo, Moon Hong

    2001-07-01

    Pickering B CANDU Unit 5 had experienced leakage at sleeve/tube joint due to severe and local pitting in 1992{delta}1993. One year later, OHT developed electrosleeving techniques for steam generator tube repair which was applied at Pickering B CANDU Unit 5, Oconee Unit 1 and Callaway in 1994, 1995 and 1999 respectively. In the results of electrosleeved tube test, electrosleeve materials were stronger than mother tubes in mechanical properties and corrosion resistance under design criteria. Two analytical models were originally developed for estimating the failure temperature under severe accident transients. Electrosleeve, a structural layer of fine grained nickel is electroformed onto the strike by circulating an aqueous solution of Ni sulfate or sulfamate with NiCO3. The patents published by FTI said that the electrolyte for electroforming the structural layer contains a pinning agent to inhibit growth of metal grains in the electroformed layer. The pinning agent contains phosphoric, phosphorous acid, molybdenum. In localization of electrosleeving, there are some problems like as 1)low plating rate, 2)high residual stress, 3)alloy composition, 4)low material properties at high temperature. Ni-Fe plating exhibit anomalous codeposition; that is less noble metal, Fe, deposits preferentially to the more noble metal, Ni. Ductility decrease and residual stress increase with increase of Fe content in plate layer. Addition of particle size of 10{delta}400{mu}m makes residual stress compressive in plate layer. Composite plating show excellent high temperature properties.

  16. Repair technique for steam generator tubes using electroforming

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Jeong, Hyun Kyu; Seo, Moon Hong

    2001-07-01

    Pickering B CANDU Unit 5 had experienced leakage at sleeve/tube joint due to severe and local pitting in 1992δ1993. One year later, OHT developed electrosleeving techniques for steam generator tube repair which was applied at Pickering B CANDU Unit 5, Oconee Unit 1 and Callaway in 1994, 1995 and 1999 respectively. In the results of electrosleeved tube test, electrosleeve materials were stronger than mother tubes in mechanical properties and corrosion resistance under design criteria. Two analytical models were originally developed for estimating the failure temperature under severe accident transients. Electrosleeve, a structural layer of fine grained nickel is electroformed onto the strike by circulating an aqueous solution of Ni sulfate or sulfamate with NiCO3. The patents published by FTI said that the electrolyte for electroforming the structural layer contains a pinning agent to inhibit growth of metal grains in the electroformed layer. The pinning agent contains phosphoric, phosphorous acid, molybdenum. In localization of electrosleeving, there are some problems like as 1)low plating rate, 2)high residual stress, 3)alloy composition, 4)low material properties at high temperature. Ni-Fe plating exhibit anomalous codeposition; that is less noble metal, Fe, deposits preferentially to the more noble metal, Ni. Ductility decrease and residual stress increase with increase of Fe content in plate layer. Addition of particle size of 10δ400μm makes residual stress compressive in plate layer. Composite plating show excellent high temperature properties

  17. 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

  18. Tube tightness survey during Phenix steam generator operation

    International Nuclear Information System (INIS)

    Cambillard, E.

    1976-01-01

    Phenix steam generators are once-through vessels with single-wall heat-exchange tubes. This design means that any leakage of water into the sodium must be detected as quickly as possible so that the installation can be shut down before extensive damage occurs. The detection of water leaks in Phenix steam generators is based on measurement of the concentration in the sodium, of hydrogen produced by the sodium-water reaction. Since the various modules--evaporators, superheaters, and reheaters--have no free sodium surfaces, detection of hydrogen in argon is not used in Phenix steam generators. The measurement systems employ a probe made of nickel tubes 0.3 mm thick. Hydrogen in the sodium diffuses into a chamber kept under vacuum by an ion pump. The hydrogen pressure in the chamber is measured by a quadrupole mass spectrometer. The nine measurement systems (three per steam generator) are calibrated by injecting hydrogen into the sodium of the secondary circuits. The data-processing computer calculates the hydrogen concentration in the sodium from the spectrometer signals and the probe temperatures, which are not regulated in Phenix; it generates instructions that enable the operator to act if a leak appears. So far, no leaks have been detected. These systems also make it possible to determine rates of hydrogen diffusion caused by corrosion of the steel walls on the water side

  19. Viewing device of a steam generator tube-plate

    International Nuclear Information System (INIS)

    Denis, J.; Poirier, D.

    1984-01-01

    The invention proposes a device to observe the tubular plate of a steam generator including rows of parallel tubes situated in a shell provided with at least one entrance situated face to the interval between two adjacent rows. The device comprises a boom of which transversal dimension is less important than the interval; the boom can be inserted by the entrance; it contains a rigid endoscope terminated in an eyepiece and an optical fibre lighguide in the same vertical plane for illumination of the far end. The respective rotary angled mirrors are driven simultaneously by drums connected to a rack-and-pinion mechanism which is operated by a plunger held by a spring against a rocking lever driven by a motor and cam. As the mirrors rotate, the illuminated zone overlaps the field of view of the endoscope. The tube plate area in the shadow of the endoscope mirror (20) is illuminated separately by an ailiary fibre with a fixed terminal mirror. The invention enables the observation of the tube plate on both sides of the boom. It can be used in the case of the inspection of the steam generator of a pressurized water reactor [fr

  20. Process Technology Development of Ni Electroplating in Steam Generator Tube

    International Nuclear Information System (INIS)

    Kim, Joung Soo; Kim, H. P.; Lim, Y. S.; Kim, S. S.; Hwang, S. S.; Yi, Y. S.; Kim, D. J.; Jeong, M. K.

    2009-11-01

    Operating nuclear power steam generator tubing material, Alloy 600, having superior resistance to corrosion has many experiences of damage by various corrosion mechanisms during long term operation period. In this research project, a new Ni electroplating technology to be applied to repair the damaged steam generator tubes has been developed. In this technology development, the optimum conditions for variables affecting the Ni electroplating process, optimum process conditions for maximum adhesion forces at interface between were established. The various mechanical properties (RT and HT tensile, fatigue, creep, burst, etc.) and corrosion properties (general corrosion, pitting, crevice corrosion, stress corrosion cracking, boric acid corrosion, doped steam) of the Ni plated layers made at the established optimum conditions have been evaluated and confirmed to satisfy the specifications. In addition, a new ECT probe developed at KAERI enable to detect defects from magnetic materials was confirmed to be used for Ni electroplated Alloy 600 tubes at the field. For the application of this developed technology to operating plants, a mock-up electroplating system has been designed and manufactured, and set up at Doosan Heavy Industry Co. and also its performance test has been done. At same time, the anode probe has been modified and improved to be used with the established mock-up system without any problem

  1. Duplex-tube sodium-indication steam generator

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1984-01-01

    The steam generator with duplex tubes and sodium indication is connected to the main sodium input and output via the inlet and outlet chambers and has indication spaces connected to the interspaces of the duplex tubes. The first indication space is linked with the auxiliary inlet pipe to the inlet chamber and the second indication space is connected with the auxiliary pipe to the outlet chamber. Mounted to the auxiliary inlet pipe is at least one closure, i.e., a valve or electromagnetic stop. Mounted on the auxiliary outlet pipe is an indication sensor, e.g., a sodium flow sensor. At least one indication space is provided with an alarm sensor, e.g., a thermocouple, a pressure gauge and one sensor to monitor the hydrogen content of sodium. (J.P.)

  2. 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)

  3. LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

    This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

  4. Transduced for determining if steam generator tubes are locked in at support plate

    International Nuclear Information System (INIS)

    Hayes, J.K.

    1984-01-01

    A nuclear steam generator is described which includes a vessel, means to introduce vaporizable fluid into the bottom portion of the vessel, an outlet near the top through which vapor is discharged, a horizontal tube sheet extending across the vessel, a plurality of U-shaped tubes, having each end secured to and extending through the tube sheet, means for introducing heating fluid to one end of each of the U-shaped tubes, means for removing heating fluid from the other end of each of the U-shaped tubes, tube support means positioned within the vessel for preventing tube vibration, the tube support means including horizontally positioned means closely surrounding, but slightly spaced from each tube, means through which access can be had to the vessel interior beneath the tube sheet when the steam generator is not in operation, and testing means for determining whether or not a tube is locked into a tube support means including a longitudinal member, with a first end located inside the tube to be tested, and a second end located outside of the tube, means for securing the first end of the member to the inside of the tube, means for heating a length of the longitudinal member, and an equal length of the tube, to an elevated temperature, and means for indicating movement of the second end of the longitudinal member away from the tube end, which would indicate that the tube is locked into the support means

  5. Development of a sealed-accelerator-tube neutron generator

    Science.gov (United States)

    Verbeke; Leung; Vujic

    2000-10-01

    Sealed-accelerator-tube neutron generators are being developed in Lawrence Berkeley National Laboratory (LBNL) for applications ranging from neutron radiography to boron neutron capture therapy and neutron activation analysis. The new generation of high-output neutron generators is based on the D-T fusion reaction, producing 14.1-MeV neutrons. The main components of the neutron tube--the ion source, the accelerator and the target--are all housed in a sealed metal container without external pumping. Thick-target neutron yield computations are performed in this paper to estimate the neutron yield of titanium and scandium targets. With an average deuteron beam current of 1 A and an energy of 120 keV, a time-averaged neutron production of approximately 10(14) n/s can be estimated for a tritiated target, for both pulsed and cw operations. In mixed deuteron/triton beam operation, a beam current of 2 A at 150 keV is required for the same neutron output. Recent experimental results on ion sources and accelerator columns are presented and discussed.

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

    Directory of Open Access Journals (Sweden)

    HEOK-SOON LIM

    2014-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-15

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  9. Overview of steam generator tube-inspection technology

    International Nuclear Information System (INIS)

    Obrutsky, L.; Renaud, J.; Lakhan, R.

    2014-01-01

    Degradation of steam generator (SG) tubing due to both mechanical and corrosion modes has resulted in extensive repairs and replacement of SGs around the world. The variety of degradation modes challenges the integrity of SG tubing and, therefore, the stations' reliability. Inspection and monitoring aimed at timely detection and characterization of the degradation is a key element for ensuring tube integrity. Up to the early-70's, the in-service inspection of SG tubing was carried out using single-frequency eddy current testing (ET) bobbin coils, which were adequate for the detection of volumetric degradation. By the mid-80's, additional modes of degradation such as pitting, intergranular attack, and axial and circumferential inside or outside diameter stress corrosion cracking had to be addressed. The need for timely, fast detection and characterization of these diverse modes of degradation motivated the development in the 90's of inspection systems based on advanced probe technology coupled with versatile instruments operated by fast computers and remote communication systems. SG inspection systems have progressed in the new millennium to a much higher level of automation, efficiency and reliability. Also, the role of Non Destructive Evaluation (NDE) has evolved from simple detection tools to diagnostic tools that provide input into integrity assessment decisions, fitness-far-service and operational assessments. This new role was motivated by tighter regulatory requirements to assure the safety of the public and the environment, better SG life management strategies and often self-imposed regulations. It led to the development of advanced probe technologies, more reliable and versatile instruments and robotics, better training and qualification of personnel and better data management and analysis systems. This paper provides a brief historical perspective regarding the evolution of SG inspections and analyzes the motivations behind that evolution. It presents an

  10. Overview of steam generator tube-inspection technology

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.; Renaud, J.; Lakhan, R., E-mail: obrutskl@aecl.ca, E-mail: renaudj@aecl.ca, E-mail: lakhanr@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-03-15

    Degradation of steam generator (SG) tubing due to both mechanical and corrosion modes has resulted in extensive repairs and replacement of SGs around the world. The variety of degradation modes challenges the integrity of SG tubing and, therefore, the stations' reliability. Inspection and monitoring aimed at timely detection and characterization of the degradation is a key element for ensuring tube integrity. Up to the early-70's, the in-service inspection of SG tubing was carried out using single-frequency eddy current testing (ET) bobbin coils, which were adequate for the detection of volumetric degradation. By the mid-80's, additional modes of degradation such as pitting, intergranular attack, and axial and circumferential inside or outside diameter stress corrosion cracking had to be addressed. The need for timely, fast detection and characterization of these diverse modes of degradation motivated the development in the 90's of inspection systems based on advanced probe technology coupled with versatile instruments operated by fast computers and remote communication systems. SG inspection systems have progressed in the new millennium to a much higher level of automation, efficiency and reliability. Also, the role of Non Destructive Evaluation (NDE) has evolved from simple detection tools to diagnostic tools that provide input into integrity assessment decisions, fitness-far-service and operational assessments. This new role was motivated by tighter regulatory requirements to assure the safety of the public and the environment, better SG life management strategies and often self-imposed regulations. It led to the development of advanced probe technologies, more reliable and versatile instruments and robotics, better training and qualification of personnel and better data management and analysis systems. This paper provides a brief historical perspective regarding the evolution of SG inspections and analyzes the motivations behind that

  11. Entropy Generation of Shell and Double Concentric Tubes Heat Exchanger

    Directory of Open Access Journals (Sweden)

    basma abbas abdulmajeed

    2016-06-01

    Full Text Available Entropy generation was studied for new type of heat exchanger (shell and double concentric tubes heat exchanger. Parameters of hot oil flow rate, temperature of inlet hot oil and pressure drop were investigated with the concept of entropy generation. The results showed that the value of entropy generation increased with increasing the flow rate of hot oil and when cold water flow rate was doubled from 20 to 40 l/min, these values were larger. On the other hand, entropy generation increased with increasing the hot oil inlet temperature at a certain flow rate of hot oil. Furthermore, at a certain hot oil inlet temperature, the entropy generation increased with the pressure drop at different hot oil inlet flow rates. Finally, in order to keep up with modern technology, infrared thermography camera was used in order to measure the temperatures. The entropy generation was determined with lower values when infrared thermography camera was used to measure the temperatures, compared with the values obtained by using thermocouples.

  12. A compact nanosecond pulse generator for DBD tube characterization

    Science.gov (United States)

    Rai, S. K.; Dhakar, A. K.; Pal, U. N.

    2018-03-01

    High voltage pulses of very short duration and fast rise time are required for generating uniform and diffuse plasma under various operating conditions. Dielectric Barrier Discharge (DBD) has been generated by high voltage pulses of short duration and fast rise time to produce diffuse plasma in the discharge gap. The high voltage pulse power generators have been chosen according to the requirement for the DBD applications. In this paper, a compact solid-state unipolar pulse generator has been constructed for characterization of DBD plasma. This pulsar is designed to provide repetitive pulses of 315 ns pulse width, pulse amplitude up to 5 kV, and frequency variation up to 10 kHz. The amplitude of the output pulse depends on the dc input voltage. The output frequency has been varied by changing the trigger pulse frequency. The pulsar is capable of generating pulses of positive or negative polarity by changing the polarity of pulse transformer's secondary. Uniform and stable homogeneous dielectric barrier discharge plasma has been produced successfully in a xenon DBD tube at 400-mbar pressure using the developed high voltage pulse generator.

  13. Tests and analysis on steam generator tube failure propagation

    International Nuclear Information System (INIS)

    Tanabe, Hiromi

    1990-01-01

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

  14. Process for installing tubes in a steam generator

    International Nuclear Information System (INIS)

    Boula, G.; George, A.

    1988-01-01

    This process consists essentially to introduce the tubes by planar layers, to place antivibration bars above the layer and tensioning the bars with forces perpendicular to the layer, to check the play between the bars and the tubes and to replace the tubes beyond tolerance by other tubes [fr

  15. Wear on Plugged Tube due to the Foreign Objects on the Secondary Side of Steam Generator

    International Nuclear Information System (INIS)

    Kim, Hyung Nam; Cho, Nam Cheoul; Nam, Min Woo

    2013-01-01

    In this paper, the changes of the tube frequency and amplitude are introduced before and after plugging. The amplitude of the bottom span for the steam generator tube is not much changed after tube plugging. Moreover, the contact force between the plugged tube and the foreign object is the same as that of intact tube and the foreign object. However, the frequencies of plugged tubes are about 9∼12% higher than those of intact tubes. That means the wear due to the foreign object would be accelerated after the tube plugging. Therefore, the tube stabilizer should be installed when the tube is plugged due to the foreign object wear. The tube wall of steam generator is a pressure boundary between the coolant of the primary system and the feedwater of the secondary system. It is very important to insure the structural integrity of the tubes because the radioactive coolant is flow into the feedwater due to the pressure difference as the result of tube failure. The degradations of steam generator tubes are corrosion, wear, fatigue and foreign object wear, etc. The foreign object wear is one of mechanical degradation due to materials flew into the secondary side of steam generator. The steam generator tubes, estimated not to insure structural integrity from the results of the nondestructive evaluation such as eddy current test and visual inspection, are excluded from the service with plugging. However, the tube wear is still being progressed after the plugging because the relative motion between the tube and structure is still existed due to the secondary side flow in the steam generator. If the tube is completely cut because of the degradation, the tube can be a stress or of failure of tubes around the plugged tube. The contact force between the structure and tube is lowered as the wear is progressed. However, the contact force between the foreign object and tube is not changed as the wear is progressed. Therefore, the structural integrity of tubes around the foreign

  16. Framatome recent developments and application on site in NDE of steam generator tubes

    International Nuclear Information System (INIS)

    Bieth, M.

    1986-01-01

    The increasing needs concerning the follow up and expertise of PWR steam generator (SG) tubing have led Framatome to develop a quick on-site intervention mobile unit, which could implement any current technique and equipment. Besides, Framatome has developed several non destructrive examination methods to solve the specific problems encountered in service on the SG tubes: profilometry of the SG tubes by eddy current. Inside and above the tube sheet, eddy current inspection of tube sleeving by ultrasonic testing and eddy current

  17. A study on integrity of LMFBR secondary cooling system to hypothetical tube failure propagation in the steam generator

    International Nuclear Information System (INIS)

    Yoshihisa Shindo; Kazuo Haga

    2005-01-01

    Full text of publication follows: A fundamental safety issue of liquid-metal-cooled fast breeder reactor (LMFBR) is to maintain the integrity of the secondary cooling system components against violent chemical sodium-water reaction caused by the water leak from the heat transfer tube of steam generators (SG). The produced sodium-water reaction jet would attack more severely surrounding tubes and would cause other tube failures (tube failure propagation), if it was assumed that the water leak was not detected by function-less detectors and proper operating actions to mitigate the tube failure propagation, such as isolations of the SG from the secondary cooling system and turbine water/steam system, and blowing water and steam inside tubes in the SG, were not taken. This study has been made focusing on the affection of large-scale water leak enlarged due to SG tube failure propagation to the structural integrity of the secondary cooling system because the generated pressure pulse caused by a large-scale sodium-water reaction might break heat transfer tubes of the intermediate heat exchanger (IHX). The present work has been made as one part of the study of probabilistic safety assessment (PSA) of LMFBR, because if the heat-transfer tubes of IHX were failed, the reactor core may be affected by the pressure pulse and/or by the sodium-water reaction products transported through the primary cooling system. As tools for PSA of the water leak incident of SG, we have developed QUARK-LP Version 4 code that mainly analyzes the high temperature rupture phenomena and estimates the number of failed tubes during the middle-scale water leak. The pressure pulse behavior generated by sodium-water reaction in the failure SG and the pressure propagation in the secondary cooling system are calculated by using the SWAAM-2 code developed by ANL. Furthermore, the quasi-steady state high pressure and temperature of the secondary cooling system in a long term is estimated by using the SWAAM

  18. Define optimal conditions for steam generator tube integrity and an extended steam generator service life

    International Nuclear Information System (INIS)

    Lu, Y.C.

    2007-01-01

    Steam generator (SG) tubing materials are susceptible to corrosion degradation in certain electrochemical corrosion potential regions in the presence of some aggressive ions. Because of the hideout of impurities, the local chemistry conditions in areas under sludge and inside SG crevices may be very aggressive with high concentrations of chlorides and other impurities. These areas are the locations where SG tubing materials are susceptible to degradation such as pitting, crevice corrosion, intergranular attack (IGA) and stress corrosion cracking (SCC). The corrosion susceptibility of each SG alloy is different and is a function of the electrochemical corrosion potential (ECP) and chemical environment. Electrochemical corrosion behaviors of major SG tube alloys were studied under some plausible aggressive crevice chemistry conditions. The possible hazardous conditions leading to SG tube degradation and the conditions, which can minimize SG tube degradation have been determined. Optimal operating conditions in the form of a 'Recommended ECP/pH zone' for minimizing corrosion degradation have been defined for all major SG tube materials, including Alloys 600, 800, 690 and 400, under CANDU SG operating and startup conditions. SCC tests and accelerated corrosion tests were carried out to verify and revise the recommended ECP/pH zones. This information is being incorporated into ChemAND, a system health monitor for plant chemistry management developed by AECL, which alloys utilities to evaluate the status of the SG alloys and to minimize SG material degradation by appropriate SG water chemistry management. (author)

  19. Pressure loss characteristics of LSTF steam generator heat-transfer tubes. Pressure loss increase due to tube internal instruments

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro

    1994-11-01

    The steam generator of the Large-Scale Test Facility (LSTF) includes 141 heat-transfer U-tubes with different lengths. Six U-tubes among them are furnished with 15 or 17 probe-type instruments (conduction probe with a thermocouple; CPT) protuberant into the primary side of the U-tubes. Other 135 U-tubes are not instrumented. This results in different hydraulic conditions between the instrumented and non-instrumented U-tubes with the same length. A series of pressure loss characteristics tests was conducted at a test apparatus simulating both types of U-tube. The following pressure loss coefficient (K CPT ) was reduced as a function of Reynolds number (Re) from these tests under single-phase water flow conditions. K CPT =0.16 5600≤Re≤52820, K CPT =60.66xRe -0.688 2420≤Re≤5600, K CPT =2.664x10 6 Re -2.06 1371≤Re≤2420. The maximum uncertainty is 22%. By using these results, the total pressure loss coefficients of full length U-tubes were estimated. It is clarified that the total pressure loss of the shortest instrumented U-tube is equivalent to that of the middle-length non-instrumented U-tube and also that a middle-length instrumented U-tube is equivalent to the longest non-instrumented U-tube. Concludingly. it is important to take account of the CPT pressure loss mentioned above in estimation of fluid behavior at the non-instrumented U-tubes either by using the LSTF experiment data from the CPT-installed U-tubes or by using any analytical codes. (author)

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

    International Nuclear Information System (INIS)

    Cheng Weili; Finnie, I.

    1996-01-01

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

  1. Sleeve type repair of degraded nuclear steam generator tubes

    International Nuclear Information System (INIS)

    Ayres, P.S.; Stark, L.E.; Feldstein, J.G.; Fu, T.

    1986-01-01

    A sealable sleeve is described for insertion into the repair of a degraded tube which consists of: a hollow core inner member of the same material as the degraded tube; a thinner outer member of substantially pure nickel and resistant to corrosive attack, the outer member being metallurgically bonded with the inner member; an expanded portion of the sleeve at one end for positioning in the tube within a tube sheet; a multiplicity of grooves formed in and adjacent to the other end of the sleeve which extends into the free-standing portion of the tube beyond the tube sheet, and a noble metal braze material contained in the grooves

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  3. Rotating sensor technology for the inspection of steam generator tubing

    International Nuclear Information System (INIS)

    Glass, S.W.; Richards, T.A.

    1986-01-01

    A high-resolution profilometry system, has been developed to assess the dimensional condition of steam generator tubes and rapidly produce the data to evaluate the potential for developing in-service leaks. The probe has an electromechanical sensor in a rotating head. This technique has been demonstrated in the field at four U.S. plants and one plant owned by Electricite de France. The Indian Point-2 plant of Consolidated Edison has twice used this technology to save tubes that would have been plugged with the go-gauge criterion and identifying other high-risk candidates for plugging that might otherwise not have been removed from service. As an extension of the PROFIL-360 technology, a rotating eddy current system (EDDY-360) has also been developed. The system provides improved sensitivity, resolution, and characterization of small-volume flaws and complete circumferential coverage as compared to conventional (bobbin and 8 x 1) eddy current techniques. Enhanced eddy current data processing provides on-line data analysis and real-time imaging of detected flaws. (author)

  4. Direct solar steam generation inside evacuated tube absorber

    Directory of Open Access Journals (Sweden)

    Khaled M. Bataineh

    2016-12-01

    Full Text Available Direct steam generation by solar radiation falling on absorber tube is studied in this paper. A system of single pipe covered by glass material in which the subcooled undergoes heating and evaporation process is analyzed. Mathematical equations are derived based on energy, momentum and mass balances for system components. A Matlab code is built to simulate the flow of water inside the absorber tube and determine properties of water along the pipe. Widely accepted empirical correlations and mathematical models of turbulent flow, pressure drop for single and multiphase flow, and heat transfer are used in the simulation. The influences of major parameters on the system performance are investigated. The pressure profiles obtained by present numerical solution for each operation condition (3 and 10 MPa matches very well experimental data from the DISS system of Plataforma Solar de Almería. Furthermore, results obtained by simulation model for pressure profiles are closer to the experimental data than those predicted by already existed other numerical model.

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

    Science.gov (United States)

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

    2013-05-01

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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1964-01-01

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

  8. Generation of Submicron Bubbles using Venturi Tube Method

    Science.gov (United States)

    Wiraputra, I. G. P. A. E.; Edikresnha, D.; Munir, M. M.; Khairurrijal

    2016-08-01

    In this experiment, submicron bubbles that have diameters less than 1 millimeter were generated by mixing water and gas by hydrodynamic cavitation method. The water was forced to pass through a venturi tube in which the speed of the water will increase in the narrow section, the throat, of the venturi. When the speed of water increased, the pressure would drop at the throat of the venturi causing the outside air to be absorbed via the gas inlet. The gas was then trapped inside the water producing bubbles. The effects of several physical parameters on the characteristics of the bubbles will be discussed thoroughly in this paper. It was found that larger amount of gas pressure during compression will increase the production rate of bubbles and increase the density of bubble within water.

  9. Internal oxidation as a mechanism for steam generator tube degradation

    Energy Technology Data Exchange (ETDEWEB)

    Gendron, T.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Scott, P.M. [Framatome, Paris (France); Bruemmer, S.M. [Pacific Northwest National Laboratory, Richland, WA (United States); Thomas, L.E. [Washington State Univ., School of Mechanical and Materials Engineering, Pullman, WA (United States)

    1999-12-01

    Internal oxidation has been proposed as a plausible mechanism for intergranular stress-corrosion cracking (IGSCC) of alloy 600 steam generator tubing. This theory can reconcile the main thermodynamic and kinetic characteristics of the observed cracking in hydrogenated primary water. Although secondary-side IG attack or IGSCC is commonly attributed to the presence of strong, caustic or acidic solutions, more recent evidence suggests that this degradation takes place in a near neutral environment, possibly dry polluted steam. As a result, internal oxidation is also a feasible mechanism for secondary side degradation. The present paper reviews experimental work conducted in an attempt to determine the validity of this mechanism. The consequences for the expected behaviour of alloys 690 and 800 replacement materials are also described. (author)

  10. Internal oxidation as a mechanism for steam generator tube degradation

    Energy Technology Data Exchange (ETDEWEB)

    Gendron, T.S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Scott, P.M. [Framatome, Paris (France); Bruemmer, S.M. [Pacific Northwest National Lab., Richland, Washington (United States); Thomas, L.E. [Washington State Univ., School of Mechanical and Materials Engineering, Pullman, WA (United States)

    1998-07-01

    Internal oxidation has been proposed as a plausible mechanism for intergranular stress corrosion cracking (IGSCC) of alloy 600 steam generator tubing. This theory can reconcile the main thermodynamic and kinetic characteristics of the observed cracking in hydrogenated primary water. Although secondary side IG attack or IGSCC is commonly attributed to the presence of strong caustic or acidic solutions, more recent evidence suggests that this degradation takes place in a near-neutral environment, possibly dry polluted steam. As a result, internal oxidation is also a feasible mechanism for secondary side degradation. The present paper reviews experimental work carried out in an attempt to determine the validity of this mechanism. The consequences for the expected behaviour of alloys 690 and 800 replacement materials are also described. (author)

  11. Internal oxidation as a mechanism for steam generator tube degradation

    International Nuclear Information System (INIS)

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

    1998-01-01

    Internal oxidation has been proposed as a plausible mechanism for intergranular stress corrosion cracking (IGSCC) of alloy 600 steam generator tubing. This theory can reconcile the main thermodynamic and kinetic characteristics of the observed cracking in hydrogenated primary water. Although secondary side IG attack or IGSCC is commonly attributed to the presence of strong caustic or acidic solutions, more recent evidence suggests that this degradation takes place in a near-neutral environment, possibly dry polluted steam. As a result, internal oxidation is also a feasible mechanism for secondary side degradation. The present paper reviews experimental work carried out in an attempt to determine the validity of this mechanism. The consequences for the expected behaviour of alloys 690 and 800 replacement materials are also described. (author)

  12. Intelligent monitoring of YAG laser welding on steam generator tubes

    International Nuclear Information System (INIS)

    Hosaka, Shigetaka; Nagura, Yasumi; Ishide, Takashi; Nagashima, Tadashi; Akaba, Takashi

    1992-01-01

    The 'KASHIKOKI' intelligent device for monitoring the YAG laser welding of steam generator tubes is described in this paper. The 'KASHIKOKI', it monitors the series of six channels, for example, the reflected laser beam and the welding speed, etc. It learns the normal criteria and the anomalous criteria of welding, and discriminates between normal and anomalous welding using the learned criteria, and distinguishes the anomaly into several types. As the results of evaluation test, the degree of correspondence between this device and an expert is about 90%. This paper describes the new methods the multi-variate analysis model for discriminating between normal and anomalous welding, and a neural network model for distinguishing the types of anomaly. (author)

  13. Experimental results of the consequences of sodium water reactions at the bottom tube plate region of straight tube steam generators

    International Nuclear Information System (INIS)

    Ruloff, G.

    1990-01-01

    Experience with sodium water reactions has shown, that the course of such a steam generator accident depends strongly on its place in the steam generator. For the EFR steam generators we have to differentiate between: weld region at the upper tube plate (gas space); bundle region; weld region at the bottom tube plate. This paper describes results of a running tests program simulating the bottom tube plate area. One main part of these tests is the investigation of the influence of wastage protection shrouds between the tubes in the weld region to avoid a fast leak propagation and to give time for leak detection and mastering of the accidents. (author). 10 figs, 2 tabs

  14. Fixing device for a tube bundle especially for steam generator

    International Nuclear Information System (INIS)

    Fournier, Y.

    1983-01-01

    The helical tubes in concentric layers are maintained by a device comprising longitudinal rods with concave cylindrical slots to hold the tubes of the same cylindrical layer. The rods are radially disposed for every cylindrical layers. The tubes are fixed on the rods by fixation elements between two successive rods on the same radius. The tube is maintained in the slot isostatically by three points [fr

  15. Implementation status of performance demonstration program for steam generator tubing analysts in Korea

    International Nuclear Information System (INIS)

    Cho, Chan Hee; Lee, Hee Jong; Yoo, Hyun Ju; Nam, Min Woo; Hong, Sung Yull

    2013-01-01

    Some essential components in nuclear power plants are periodically inspected using non destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%

  16. Implementation status of performance demonstration program for steam generator tubing analysts in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chan Hee; Lee, Hee Jong; Yoo, Hyun Ju; Nam, Min Woo [KHNP Central Research Institute, Daejeon (Korea, Republic of); Hong, Sung Yull [Yeungnam Univ., Gyeongsan (Korea, Republic of)

    2013-02-15

    Some essential components in nuclear power plants are periodically inspected using non destructive examinations, for example ultrasonic, eddy current and radiographic examinations, in order to determine their integrity. These components include nuclear power plant items such as vessels, containments, piping systems, pumps, valves, tubes and core support structure. Steam generator tubes have an important safety role because they constitute one of the primary barriers between the radioactive and non radioactive sides of the nuclear power plant. There is potential that if a tube bursts while a plant is operating, radioactivity from the primary coolant system could escape directly to the atmosphere. Therefore, in service inspections are critical in maintaining steam generator tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due tube integrity. In general, the eddy current testing is widely used for the inspection of steam generator tubes due to its high inspection speed and flaw detectability on non magnetic tubes. However, it is not easy to analyze correctly eddy current signals because they are influenced by many factors. Therefore, the performance of eddy current data analysts for steam generator tubing should be demonstrated comprehensively. In Korea, the performance of steam generator tubing analysts has been demonstrated using the Qualified Data Analyst program. This paper describes the performance demonstration program for steam generator tubing analysts and its implementation results in Korea. The pass rate of domestic analysts for this program was 71.4%.

  17. Development of safety evaluation technique of steam generator tubes for the next generation

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyuk Sang; Kim, I. S.; Ann, Se Jin; Lee, S. J.; Seo, M. S.; Lee, Y. H.; Kim, J. H.; Hong, J. G. [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2000-02-15

    Subject 1 - a technique for predicting the SCC susceptibility of steam generator tube material based on the repassivation kinetics was developed and the effects of Pb in the repassivation rate and SCC susceptibility rate of tube material was investigated with this technique. An alloy with a higher slope value of log i(t) vs. q(t) plot based on the current transient curve obtained by scratch test and a lower slope value log i(t) vs. l/q(t) plot (cBV) is repassivated faster with a more protective passive film and it can be predicted that it will show higher resistance to SCC. With PbO addition in all solution studied (pH 4, pH 10, Cl- containing pH 4), alloy 690TT showed decreased repassivation rate. So it can be predict that PbO addition lower the resistance of SCC of steam generator tune material. Subject 2 - SG wear testing of tube and support materials has been conducted at various load and sliding amplitude in air environment. The results showed effect of normal load and sliding amplitude on SG tube wear damage. It was also shown that, for predominantly sliding motion, the SG wear coefficient of work-rate model is lower for Inconel 690TT compared with inconel 600MA. SG tube wear data show that, for work-rates ranging from 4 to 25mW, average tube wear coefficient of 43.76{approx}54.05 X 10{sup 15} Pa{sup -1} for Inconel 600MA and 26.88{approx}33.94 X 10{sup -15} Pa{sup 1} for Inconel 690TT against 405 and 409 stainless steels.

  18. Calculation of reverse flow in inverted U-Tubes of steam generator during natural circulation

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Jinggong; Liu Ruolei; Qin Shiwei; Huang Yanping

    2010-01-01

    The mechanism of reverse flow in inverted U-tubes of steam generators of pressurized water reactors during natural circulation is analyzed by using the full range characteristic curve of parallel U-tubes. A lumped-distributed model to calculate the reverse flow occurred in inverted U-tubes of real steam generators with a large number of U-tubes during natural circulation is developed. The model has the advantages of quick calculation and high accuracy for the analysis of reverse flow in inverted U-tubes of real steam generators with natural circulation. This model has been used to calculate the normal and reverse flows occurred in inverted U-tubes of a steam generator with natural circulation. The comparison of calculated results indicates a well agreement with that predicted by the model in which normal or reverse flow in each individual U-tube is analyzed, which verifies the reliability of the model developed in this paper. (authors)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  20. Globe Rupture

    Directory of Open Access Journals (Sweden)

    Reid Honda

    2017-07-01

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

  1. 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

  2. Measurement by eddy currents of tube-antivibratory bar gap steam generators of PWR power plants

    International Nuclear Information System (INIS)

    Savin, E.; Bieth, M.; Floze, J.C.

    1990-01-01

    In steam generators tubes are maintained by AVB to limit vibrations amplitude induced by secondary fluid flow. After some years wear sometimes occurs. For gap measurement between tubes and AVB Framatome developed a method based on eddy current and using a probe rotating inside the tube [fr

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

    International Nuclear Information System (INIS)

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

    1978-03-01

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

  4. Deposition of magnetite particles onto alloy-800 steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Basset, M.; Arbeau, N.; McInerney, J.; Lister, D.H. [Univ. of New Brunswick, Dept. of Chemical Engineering, Fredericton, NB (Canada)

    1998-07-01

    Fouling is a particularly serious problem in the power generating industry. Deposits modify the thermalhydraulic characteristics of heat transfer surfaces by changing the resistance to heat transfer and the resistance to fluid flow, and, if thick enough, can harbour aggressive chemicals. Deposits are also implicated in the increase of radiation fields around working areas in the primary heat transfer systems of nuclear power plants. In order to understand the preliminary steps of the formation of corrosion product deposits on the outsides of steam generator tubes, a laboratory program has investigated the deposition of magnetite particles from suspension in water onto Alloy-800 surfaces under various conditions of flow, chemistry and boiling heat transfer. A recirculating loop made of stainless steel operating at less than 400kPa pressure, with a nominal coolant temperature of 90 degrees C, was equipped with a vertical glass column which housed a 2.5E-01m-long Alloy-800 boiler tube capable of generating a heat flux of 240kW/m{sup 2} . A concentration of suspended magnetite of 5.0E-03kg/m{sup 3} was maintained in the recirculating coolant, which was maintained at a pH of 7.5. The magnetite was synthesized with a sol-gel process, which was developed to produce reproducibly monodispersed, colloidal (<1{mu}m) and nearly spherical particles. A radiotracing method was used to characterize the deposit evolution with time and to quantify the removal of magnetite particles. The results from a series of deposition experiments are presented here. The deposition process is described in terms of a two-step mechanism: the transport step, involving the transport from the bulk of the liquid to the vicinity of the surface, followed by the attachment step, involving the attachment of the particle onto the surface. Under non-boiling heat transfer conditions, diffusion seems to be the dominant factor ruling deposition with a small contribution from thermophoresis; removal was

  5. Deposition of magnetite particles onto alloy-800 steam generator tubes

    International Nuclear Information System (INIS)

    Basset, M.; Arbeau, N.; McInerney, J.; Lister, D.H.

    1998-01-01

    Fouling is a particularly serious problem in the power generating industry. Deposits modify the thermalhydraulic characteristics of heat transfer surfaces by changing the resistance to heat transfer and the resistance to fluid flow, and, if thick enough, can harbour aggressive chemicals. Deposits are also implicated in the increase of radiation fields around working areas in the primary heat transfer systems of nuclear power plants. In order to understand the preliminary steps of the formation of corrosion product deposits on the outsides of steam generator tubes, a laboratory program has investigated the deposition of magnetite particles from suspension in water onto Alloy-800 surfaces under various conditions of flow, chemistry and boiling heat transfer. A recirculating loop made of stainless steel operating at less than 400kPa pressure, with a nominal coolant temperature of 90 degrees C, was equipped with a vertical glass column which housed a 2.5E-01m-long Alloy-800 boiler tube capable of generating a heat flux of 240kW/m 2 . A concentration of suspended magnetite of 5.0E-03kg/m 3 was maintained in the recirculating coolant, which was maintained at a pH of 7.5. The magnetite was synthesized with a sol-gel process, which was developed to produce reproducibly monodispersed, colloidal (<1μm) and nearly spherical particles. A radiotracing method was used to characterize the deposit evolution with time and to quantify the removal of magnetite particles. The results from a series of deposition experiments are presented here. The deposition process is described in terms of a two-step mechanism: the transport step, involving the transport from the bulk of the liquid to the vicinity of the surface, followed by the attachment step, involving the attachment of the particle onto the surface. Under non-boiling heat transfer conditions, diffusion seems to be the dominant factor ruling deposition with a small contribution from thermophoresis; removal was considered

  6. Improved eddy-current inspection for steam generator tubing

    International Nuclear Information System (INIS)

    Dodd, C.V.; Pate, J.R.; Allen, J.D. Jr.

    1989-01-01

    Computer programs have been written to allow the analysis of different types of eddy-current probes and their performance under different steam generator test conditions. The probe types include the differential bobbin probe, the absolute bobbin probe, the pancake probe and the reflection probe. The generator test conditions include tube supports, copper deposits, magnetite deposits, denting, wastage, pitting, cracking and IGA. These studies are based mostly on computed values, with the limited number of test specimens available used to verify the computed results. The instrument readings were computed for a complete matrix of the different test conditions, and then the test conditions determined as a function of the readings by a least-squares technique. A comparison was made of the errors in fit and instrument drift for the different probe types. The computations of the change in instrument reading due to the defects have led to an ''inversion'' technique in which the defect properties can be computed from the instrument readings. This has been done both experimentally and analytically for each of these probe types. 3 refs., 13 figs., 1 tab

  7. Heat transfer enhancement for fin-tube heat exchanger using vortex generators

    International Nuclear Information System (INIS)

    Yoo, Seong Yeon; Park, Dong Seong; Chung, Min Ho; Lee, Sang Yun

    2002-01-01

    Vortex generators are fabricated on the fin surface of a fin-tube heat exchanger to augment the convective heat transfer. In addition to horseshoe vortices formed naturally around the tube of the fin-tube heat exchanger, longitudinal vortices are artificially created on the fin surface by vortex generators. The purpose of this study is to investigate the local heat transfer phenomena in the fin-tube heat exchangers with and without vortex generators, and to evaluate the effect of vortices on the heat transfer enhancement. Naphthalene sublimation technique is employed to measure local mass transfer coefficients, then analogy equation between heat and mass transfer is used to calculate heat transfer coefficients. Experiments are performed for the model of fin-circular tube heat exchangers with and without vortex generators, and of fin-flat tube heat exchangers with and without vortex generators. Average heat transfer coefficients of fin-flat tube heat exchanger without vortex generator are much lower than those of fin-circular tube heat exchanger. On the other hand, fin-flat tube heat exchanger with vortex generators has much higher heat transfer value than conventional fin-circular tube heat exchanger. At the same time, pressure losses for four types of heat exchanger is measured and compared

  8. 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

  9. 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

  10. New eddy current testing system in steam generator tubes

    International Nuclear Information System (INIS)

    Nagata, Yasuyuki; Shimone, Junri; Maeda, Kotaro; Harada, Yutaka; Murakami, Ryuji

    2003-01-01

    Eddy Current Testing (ECT) technique is applied to PWR Steam Generator (SG) tubes during refueling outage inspection. Bobbin type probe is generally used for the inspection, but it cannot provide sufficient detectability for very tight cracks. NEL, R/D Tech (Canada) and AECL (Canada) jointly developed new ECT probe. The probe is transmit and receive type array probe. It has axial defect detection mode and circumferential defect detection mode. Its inspection speed is as fast as conventional bobbin probe, and detectability performance is same or better than performance of rotating pancake coil (RPC) probe, which is used for detailed inspection. As a tester of the new probe, we have developed new ECT tester TC7700. It is able to transmit composite wave from the 4 independent oscillators (simultaneous excitation system) compared with the conventional inspection system. As regarding for field inspection system, we have integrated Zetec (USA) probe positioner, probe pusher and R/D Tech TC7700. We are also developing new analysis software. The software has automated data screening function, which can deal with large amount of data compared with conventional bobbin probe. We plan to perform a large-scale field trial using the probe and new system in near future. (author)

  11. Eddy current inspection of steam generator tubing plugs

    International Nuclear Information System (INIS)

    Cullen, W.K.

    1990-01-01

    In response to the issues raised regarding the integrity of nuclear steam generator tubing plugs manufactured from certain heats of Inconel 600, Westinghouse engineers have developed, qualified and implemented an eddy current inspection system for the in-place assessment of these plugs. The heart of the system is a robotic and effector which delivers an eddy current sensor through the reduced diameter of the plug expander and actuates the coil to physical contact with the expanded inside bore of the plug. Once deployed, the eddy current sensor is rotated along a helical path to produce a detailed scan of the plug surface above the final position of the expander. This testing produces an isometric display of degradation due to primary water stress corrosion cracking, on the inside surface of the plug. To date, successful inspections have been conducted at two nuclear units with two different robotic delivery systems. While designed specifically for mechanical plugs with a bottle bore cavity, the inspection system can also be used for expanded straight bore plugs. Details of the inspection system along with a discussion of qualification activities and actual field results are presented in this paper

  12. Possibilities of the metallurgical base in the manufacture of tubes for nuclear power plant steam generators

    International Nuclear Information System (INIS)

    Prnka, T.; Walder, V.; Dolenek, J.

    Current possibilities are briefly summarized of metallurgy in the manufacture of high-quality tubes for nuclear power plant steam generators, mainly for fast reactor power plants. Discussed are steel making possibilities, semi-finished product and tube forming with special regard to 2.25Cr1MoNiNb steel problems, heat treatment, finishing, and testing. Necessary equipment and technology for the production of steam generator tubes are less common in the existing practice and are demanding on investment; their introduction, however, is inevitable for securing quality production of steam generator tubes. (Kr)

  13. 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)

  14. 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)

  15. 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)

  16. Numerical and Experimental Study on a Model Draft Tube with Vortex Generators

    Directory of Open Access Journals (Sweden)

    Tian Xiaoqing

    2013-01-01

    Full Text Available A model water turbine draft tube containing vortex generators (VG was studied. Numerical simulations were performed to investigate 55 design variations of the vortex generators in a draft tube. After analyzing the shapes of streamlines and velocity distributions in the tube and comparing static pressure recovery coefficients (SPRC in different design variations, an optimum vortex generator layout, which can raise SPRC of the draft tube by 4.8 percent, was found. To verify the effectiveness of the vortex generator application, a series of experiments were carried out. The results show that by choosing optimal vortex generator parameters, such as the installation type, installation position, blade-to-blade distance, and blade inclination angle, the draft tube equipped vortex generators can effectively raise their SPRC andworking stability.

  17. Dryout in sodium-heated helically-coiled steam generator tubes

    International Nuclear Information System (INIS)

    Tomita, Y.; Kosugi, T.; Kubota, J.; Nakajima, K.; Tsuchiya, T.

    1984-01-01

    Experimental research on the dryout phenomenon in sodium heated, helically coiled steam generator tubes was carried out. The fluctuation of the tube wall temperature caused by dryout was measured with thermocouples installed in the center of the tube wall. Empirical correlations of dryout quality were developed as functions of critical heat flux, water mass velocity and saturation pressure. These correlations confirmed that the design criterion of the MONJU steam generator was reasonable. (author)

  18. 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)

  19. On the constitutive criteria for the fault: influence of size and tensile cracks generation during rupture

    International Nuclear Information System (INIS)

    Riera, J.D.; Miguel, L.F.F.; Gudiel, L.A.D.

    2005-01-01

    In recent studies the authors consider the possibility of occurrence of tensile crack generation near the free surface during faulting. The main assumption is that the dynamic tensile stress created during the shear rupture process would be dominant over the background stress near the free-surface. The proposed model was able to simulate the mechanism of cracks developed as a flower like-structure surrounding the shear fault in the vicinity of the free surface and explain some aspects of the fracture zones found after the 2000 Tottori earthquake. The applicability of various constitutive laws for the fault employed in numerical analysis of the near source motion, such as the slip-weakening, velocity-weakening or rate- and state-dependent friction laws, were also recently discussed by the authors in connection with results of laboratory experiments on friction. In this paper, the most important conclusions of the studies outlined above are summarized by way of introduction to the numerical modeling of the region adjacent to the fault surface, which includes material nonhomogeneities as well as the possibility of generation of new tensile cracks. In the presence of fracture, both factors exert important influence on the macro constitutive laws for the fault, which relate the static (mean normal and shear stresses) with kinetic (displacements) variables, as well as with their time derivatives. By means of simulation, it is shown that the coefficients in the equation that relate the relevant variables depend on size, that is, they vary with the contact area over which stresses, displacements and velocities are averaged. Any variability in Young's modulus or mass density of the surround rock may however be neglected, the only significant random fields being those that describe friction at the interface and the specific fracture energy. (authors)

  20. 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.

  1. Turbulence induced Fretting-wear characteristics of steam generator helical tubes

    International Nuclear Information System (INIS)

    Jhung, Myung Jo; Jo, Jong Chull; Kim, Hho Jung; Yune, Young Gill; Yu, Seon Oh

    2005-01-01

    This study addresses safety assessment of the potential for fretting-wear damages on steam generator helical tubes due to turbulence-induced vibration in operating nuclear power plants. To get the natural frequency, corresponding mode shape and participation factor, modal analyses are performed for helical type tubes with various conditions. Special emphases are put on the effects of coil diameter and the number of turns on the modal and fretting wear characteristics of tubes. Also, investigated are the effects of external pressure on the tube modal characteristics as well as the effects of turbulence induced vibration on the fretting-wear characteristics of tubes

  2. Process to repair a steam generator tube by inserting a tubular sleeve and the associated sleeve

    International Nuclear Information System (INIS)

    Gaudin, J.P.

    1986-01-01

    The tubular sleeve is introduced in the tube and is mechanically expanded inside the tube plate, and is diametrally expanded at its upper part within the tube and outside the tube plate. Tightness is ensured by brazing the end part of the sleeve within the tube. The end part of the sleeve is brazed by melting of brazing metal previously applied to the outer surface of the sleeve of its end region. The invention applies more particularly to steam generators of pressurized water nuclear reactors [fr

  3. 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.

  4. Fretting-wear characteristics of steam generator tubes contacting with foreign object

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

    2003-01-01

    Fretting-wear characteristics of steam generator tubes contacting with foreign object has been investigated in this study. The operating steam generator shell-side flow field conditions are obtained from three-dimensional steam generator flow calculation using a well-validated steam generator thermal-hydraulic analysis computer code. Modal analyses are performed for the finite element modelings of tubes to get the natural frequency, corresponding mode shape and participation factor. The wear rate of a steam generator tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted. In addition, the effects of internal pressure and flow velocity on the remaining life of the tube are discussed in this paper

  5. Use of virtual steam generator cassette for tube spatial design and SGC assembling procedure

    International Nuclear Information System (INIS)

    Kim, Y. W.; Kim, J. I.; Ji, S. K.

    2003-01-01

    A method of determining spatial arrangement of tube connection and assembling procedure of once-through helical steam generator cassette utilizing three dimensional virtual steam generator cassette has been developed on the basis of recent 3-D modelling technology. One ends of the steam generator tubes are connected to the module feed water header and the other sides are connected to the module steam header. Due to the complex geometry of tube arrangement, it is very difficult to connect the tubes to the module headers without the help of a physical engineering mock up. A comparative study has been performed at each design step for the tube arrangement and heat transfer area. Heat transfer area computed from thermal sizing was 4% less than that of measured. Heat transfer area calculated from the virtual steam generator cassette mock up has only 0.2% difference with that of measured. Assembling procedure of the steam generator cassette also, can be developed in the design stage

  6. Analysis of reverse flow in inverted U-tubes of steam generator under natural circulation condition

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Ruolei; Liu Jinggong; Qin Shiwei

    2008-01-01

    In this paper, we report on the analysis of reverse flow in inverted U-tubes of a steam generator under natural circulation condition. The mechanism of reverse flow in inverted U-tubes of the steam generator with natural circulation is graphically analyzed by using the full-range characteristic curve of parallel U-tubes. The mathematical model and numerical calculation method for analyzing the reverse flow in inverted U-tubes of the steam generator with natural circulation have been developed. The reverse flow in an inverted U-tube steam generator of a simulated pressurized water reactor with natural circulation in analyzed. Through the calculation, the mass flow rates of normal and reverse flows in individual U-tubes are obtained. The predicted sharp drop of the fluid temperature in the inlet plenum of the steam generator due to reverse flow agrees very well with the experimental data. This indicates that the developed mathematical model and solution method can be used to correctly predict the reverse flow in the inverted U-tubes of the steam generator with natural circulation. The obtained results also show that in the analysis of natural circulation flow in the primary circuit, the reverse flow in the inverted U-tubes of the steam generator must be taken into account. (author)

  7. 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

  8. Characteristics of U-tube assembly design for CANDU 6 type steam generators

    International Nuclear Information System (INIS)

    Park, Jun Su; Jeong, Seung Ha

    1996-06-01

    Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy water reactors, e.g. Wolsong 2, 3, and 4 NPP, a high cost of heavy water will be imposed additionally. During the plant operation, steam generator tubes can potentially be subject to adverse environmental conditions which will cause damages to U-tube assembly. Types of the damage depend upon the combined effects of design factors, materials and chemical environment of steam generator, and they are the pure water stress corrosion cracking, intergranular attack, pitting, wastage, denting, fretting and fatigue, etc. In this report, a comprehensive review of major design factors of recirculating steam generators has been performed against the potential tube damages. Then the design characteristics of CANDU-type Wolsong steam generator were investigated in detail, including tube material, thermalhydraulic aspects, tube-to-tubesheet joint, tube supports, water chemistry and sludge management. 9 tabs., 18 figs., 38 refs. (Author) .new

  9. Development of a helical-coil double wall tube steam generator for 4S reactor

    International Nuclear Information System (INIS)

    Kitajima, Yuko; Maruyama, Shigeki; Jimbo, Noboru; Hino, Takehisa; Sato, Katsuhiko

    2011-01-01

    The 4S, Super-Safe Small and Simple, is a small-sized sodium-cooled fast reactor. A fast reactor usually uses sodium as a coolant to transfer heat from core to turbine/generator system. The heat of the intermediate heat transport system and that of the water stream systems are exchanged by the steam generator (SG) tubes. If the tube failure occurs, a sodium/water reaction could be occurred. To prevent the reaction and enhance safety, a helical-coil-type double wall tube with wire mesh interlayer and continuous monitoring systems of tube failure are applied to the SG of the 4S. The development and general features of this type double wall tube were described in Ref. 1) and Ref. 2). Those paper summarized following results; The tubes studied in these references were straight type. To establish this SG, development of manufacturing method of helical-coil-type double wall tube and validation of the tube failure monitoring system are needed. In this study, three demonstration tests have been performed; welding test of the double wall tube to manufacture the tubes with 70-80m length, assembling test of the helical-coil tube, and confirmation test of the tube processing system using the fabricated helical-coil tubes. As a result, following technologies have been successfully established. (1) Development of the welding techniques for manufacturing of the helical-coil-type double wall tube with wire mesh interlayer. (2) The confirmation test for manufacturing the helical coil tube of the SG. (author)

  10. Assessment on Event Classification of One Steam Generator Tube Rupture in EU-APR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji Hwan; Kim, Yong Soo [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Fukushima Daiichi nuclear power plant accident showed the vulnerability of coping strategy to beyond design natural disaster such as beyond design earthquake and tsunami. In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. Recently, the stress tests for WOLSONG Unit 1 and KORI Unit 1 have been performed and their assessment results have been reviewed by Korean regulatory body. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings. The several additional items for safety improvement has been drawn to enhance the coping capability for loss of safety functions under beyond design natural disaster in addition to post Fukushima action items.

  11. Assessment on Event Classification of One Steam Generator Tube Rupture in EU-APR

    International Nuclear Information System (INIS)

    Kim, Ji Hwan; Kim, Yong Soo

    2016-01-01

    The Fukushima Daiichi nuclear power plant accident showed the vulnerability of coping strategy to beyond design natural disaster such as beyond design earthquake and tsunami. In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. Recently, the stress tests for WOLSONG Unit 1 and KORI Unit 1 have been performed and their assessment results have been reviewed by Korean regulatory body. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings. The several additional items for safety improvement has been drawn to enhance the coping capability for loss of safety functions under beyond design natural disaster in addition to post Fukushima action items

  12. NRC integrated program for the resolution of Unresolved Safety Issues A-3, A-4 and A-5 regarding steam generator tube integrity: Final report

    International Nuclear Information System (INIS)

    1988-09-01

    This report presents the results of the NRC integrated program for the resolution of Unresolved Safety Issues (USIs) A-3, A-4, and A-5 regarding steam generator tube integrity. A generic risk assessment is provided and indicates that risk from steam generator tube rupture (SGTR) events is not a significant contributor to total risk at a given site, nor to the total risk to which the general public is routinely exposed. This finding is considered to be indicative of the effectiveness of licensee programs and regulatory requirements for ensuring steam generator tube integrity in accordance with 10 CFR 50, Appendices A and B. This report also identifies a number of staff-recommended actions that the staff finds can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes and in mitigating the consequences of an SGTR. As part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of pressurized water reactors (PWRs) to upgrade their programs, as necessary, to meet the intent of the staff-recommended actions; however, such actions do not constitute NRC requirements. In addition, this report describes a number of ongoing staff actions and studies involving steam generator issues which are being pursued to provide added assurance that risk from SGTR events will continue to be small. 146 refs., 5 figs., 11 tabs

  13. 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.

  14. Steam-generator tube failures: world experience in water-cooled nuclear power reactors during 1972

    International Nuclear Information System (INIS)

    Stevens-Guille, P.D.

    1975-01-01

    During 1972, approximately one in three operating reactors with steam generators incurred tube failures, predominantly near the tube sheet and in the bend region. Various forms of corrosion were the most frequent cause of failure. Eddy-current inspection was the preferred method for locating and investigating the cause of failure. Extensive use was made of both mechanical and explosive plugs for repair. As a class, steam generators with Monel 400 tubes had the lowest failure rates, and those with Inconel 600 tubes had the highest. (U.S.)

  15. Evaluation on mechanical and corrosion properties of steam generator tubing materials

    Energy Technology Data Exchange (ETDEWEB)

    Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-06-15

    Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials.

  16. Evaluation on mechanical and corrosion properties of steam generator tubing materials

    International Nuclear Information System (INIS)

    Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung

    1998-06-01

    Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials

  17. Scale Thickness Measurement of Steam Generator Tubing Using Eddy Current Signal of Bobbin Coil

    International Nuclear Information System (INIS)

    Kim, Chang Soo; Um, Ki Soo; Kim, Jae Dong

    2012-01-01

    Steam generator is one of the major components of nuclear power plant and steam generator tubes are the pressure boundary between primary and secondary side, which makes them critical for nuclear safety. As the operating time of nuclear power plant increases, not only damage mechanisms but also scaled deposits on steam generator tubes are known to be problematic causing tube support flow hole blockage and heat fouling. The ability to assess the extent and location of scaled deposits on tubes became essential for management and maintenance of steam generator and eddy current bobbin data can be utilized to measure thickness of scale on tubes. In this paper, tube reference standards with various thickness of scaled deposit has been set up to provide information about the overall deposit condition of steam generator tubes, providing essential tool for steam generator management and maintenance to predict and prevent future damages. Also, methodology to automatically measure scale thickness on tubes has been developed and applied to field data to estimate overall scale amount.

  18. Simulation of a loss of coolant accident with rupture in the steam generator hot collector

    International Nuclear Information System (INIS)

    1991-03-01

    The Central Research Institute for Physics of the Hungarian Academy of Sciences designed and constructed the PMK-NVH test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary made the PMK-NVH facility available to the IAEA. The IAEA, having identified the need for experimental data due to the difficulties of building integral test facilities and the high costs of these experiments, has accepted the offer of the Hungarian Academy of Sciences and has organized three standard problem exercises. In these exercises, experimental data from the simulation of loss of coolant accidents were compared with analytical predictions of the behaviour of the facility, calculated with computer codes. The third standard problem exercise involved a test, in which the rupture was simulated to occur at the top of the hot collector of the steam generator, therefore creating a leak from primary to secondary side. Both hydroaccumulators and high pressure injection were allowed to actuate as prescribed in the actual plant. Eighteen organizations from 15 Member States took part in the exercise presenting pre-test and some post-test analyses which were discussed in a final meeting in Vienna in August, 1990. This document presents a complete overview of the third standard problem exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many interrelated steps; therefore, no general conclusion or optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation. 42 refs, figs and tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    International Nuclear Information System (INIS)

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

    1979-05-01

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

  1. Development of the visual inspection system for the top of the tube sheet in steam generators

    International Nuclear Information System (INIS)

    Kim, Gyung Sub; Choi, Sang Hoon; Kim, Ki Chul

    2008-01-01

    Steam Generators at Nuclear Power plants have a important function to isolate Radioactivity between the primary side radioactive fluid running through tubes and the secondary side with non-radioactive fluid through out of a tube bundle, in addition to a function of steam generation. Therefore, To obtain integrity of Steam Generator is really important for safety in the nuclear power plant. At the same time, sludge and foreign objects in steam generators are known as major sources causing the damage of SG tubes. But there is no way to prevent those coming to steam generators until now. Therefore, a periodic inspection and removal of those in steam generators is the only way for those Generally, Most of the Nuclear Power Plants have been inspecting visually every outage for the top of the tube sheet in which sludge and foreign objects lead to the buildup to know how these are

  2. 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

  3. Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

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

    1980-01-01

    The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

  4. 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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  6. Steam generator tube failures: world experience in water-cooled nuclear power reactors in 1975

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-11-01

    Steam generator tube failures were reported in 22 out of 62 water-cooled nuclear power plants surveyed in 1975. This was less than in 1974, and the number of the tubes affected was noticeably less. This report summarizes these failures, most of which were due to corrosion. Secondary-water chemistry control, procedures for inspection and repair, tube materials, and failure rates are discussed. (author)

  7. Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

    International Nuclear Information System (INIS)

    Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

    2017-01-01

    Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

  8. 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

  9. Automated analysis technique developed for detection of ODSCC on the tubes of OPR1000 steam generator

    International Nuclear Information System (INIS)

    Kim, In Chul; Nam, Min Woo

    2013-01-01

    A steam generator (SG) tube is an important component of a nuclear power plant (NPP). It works as a pressure boundary between the primary and secondary systems. The integrity of a SG tube can be assessed by an eddy current test every outage. The eddy current technique(adopting a bobbin probe) is currently the main technique used to assess the integrity of the tubing of a steam generator. An eddy current signal analyst for steam generator tubes continuously analyzes data over a given period of time. However, there are possibilities that the analyst conducting the test may get tired and cause mistakes, such as: missing indications or not being able to separate a true defect signal from one that is more complicated. This error could lead to confusion and an improper interpretation of the signal analysis. In order to avoid these possibilities, many countries of opted for automated analyses. Axial ODSCC (outside diameter stress corrosion cracking) defects on the tubes of OPR1000 steam generators have been found on the tube that are in contract with tube support plates. In this study, automated analysis software called CDS (computer data screening) made by Zetec was used. This paper will discuss the results of introducing an automated analysis system for an axial ODSCC on the tubes of an OPR1000 steam generator.

  10. Failures of fine tubes of steam generators and the essential defects

    International Nuclear Information System (INIS)

    Kawano, Shinji; Ebisawa, Toru; Sato, Susumu.

    1976-01-01

    Light water reactors were sold to Japan as their economy and safety have been established, but the average availability of 11 reactors in Japan during 7 year operation is only 53%, and it is being proved that there are questions in the safety and economy. In this report, the failures of fine tubes of steam generators are discussed from the standpoint of the corrosion of materials. First, the functions and construction of the fine tubes of steam generators in PWRs are explained. The failures of the fine tubes of steam generators became frequent since the beginning of 1970s as large capacity nuclear power stations have started the operation. When the fine tubes are pierced with holes during operation and the radioactivity in primary coolant leaks into secondary coolant, it is detected with radioactivity monitors. In order to find out the broken tubes, eddy current flaw detectors are used, and the tubes on which flaws were detected we plugged by explosion welding. In these works, many manual operations are included, and the radiation exposure of workers and the difficulties in the operations are the problems. The cases of the tube failures in Japan and foreign countries, the causes and the countermeasures are described. Chemical corrosion, thermal stress cycle, shaving off due to eddy flow, and stress corrosion are the probable causes. The safety of steam generators is essentially in extremely poor state. The seriousness of the tube failures in steam generators is emphasized. (Kako, I.)

  11. Development of expanded type plugging technique for leaky tubes of steam generators of Indian PHWRs

    International Nuclear Information System (INIS)

    Das, Nirupam; Samuel, K.A.; Joemon, V.; Rupani, B.B.

    2006-01-01

    Steam generators are very important component of Nuclear Power Plant (NPP), as they are part of Primary Heat Transport (PHT) system of Pressurised Heavy Water Reactors (PHWRs). A nuclear power plant of 220 MWe capacity has four mushroom type steam generators, each consisting of 1830 U-tubes (16 mm outside diameter and 1 mm wall thickness) made of Incoloy-800 material. The tubes of 'tube and shell type steam generator' act as the pressure boundary of PHT System. Any structural failure of these tubes may lead to release of radioactivity along with plant outage and significant economic loss. Hence, it is necessary to plug the leaky tubes for continued and safe operation of a steam generator. An expanded type plugging technique has been developed at Reactor Engineering Division to plug the leaky tubes. This plugging technique is selected because of low residual stress imparted in the adjacent 'tube to tube-sheet' joints. This plug meets the various codal requirements of steam generator. A number of qualification trials have been carried out with such plugs in the mock up facility. The expanded plugs meet the design requirements for pull out strength and leak-tightness. This paper describes the design concept of the plug, developmental aspects and qualification of the plugging technique. (author)

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  14. Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

    International Nuclear Information System (INIS)

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

    1978-02-01

    A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author)

  15. Improvement of hydro-turbine draft tube efficiency using vortex generator

    Directory of Open Access Journals (Sweden)

    Xiaoqing Tian

    2015-07-01

    Full Text Available Computational fluid dynamics simulation was employed in a hydraulic turbine (from inlet tube to draft tube. The calculated turbine efficiencies were compared with measured results, and the relative error is 1.12%. In order to improve the efficiency of the hydraulic turbine, 15 kinds of vortex generators were installed at the vortex development section of the draft tube, and all of them were simulated using the same method. Based on the turbine efficiencies, distribution of streamlines, velocities, and pressures in the draft tube, an optimal draft tube was found, which can increase the efficiency of this hydraulic turbine more than 1.5%. The efficiency of turbine with the optimal draft tube, draft tube with four pairs of middle-sized vortex generator, and draft tube without vortex generator under different heads of turbine (5–14 m was calculated, and it was verified that these two kinds of draft tubes can increase the efficiency of this turbine in every situation.

  16. 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.

  17. 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

  18. New Media: Engaging and Educating the YouTube Generation

    Science.gov (United States)

    Vedantham, Anu; Hassen, Marjorie

    2011-01-01

    Today's undergraduates are clearly comfortable as consumers of technology and new media--purchasing ring tones for their cell phones and tunes for their iPods, text-messaging from handheld devices, scanning and tinkering with photos, keeping up with their Facebook friends and watching viral YouTube videos, sometimes all simultaneously. We share…

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  20. 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

  1. Current Status on the Development of a Double Wall Tube Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Ho Yun; Choi, Byoung Hae; Kim, Jong Man; Kim, Byung Ho

    2007-12-15

    A fast reactor, which uses sodium as a coolant, has a lot of merits as a next generation nuclear reactor. However, the possibility of a sodium-water reaction occurrence hinders the commercialization of this reactor. As one way to improve the reliability of a steam generator, a double-wall tube steam generator is being developed in GEN-4 program. In this report, the current state of the technical developments for a double-wall tube steam generator are reviewed and a future plan for the development of a double-wall tube steam generator is established. The current focuses of this research are an improvement of the heat transfer capability for a double-wall tube and the development of a proper leak detection method for the failure of a double-wall tube during a reactor operation. The ideal goal is an on-line leak detection of a double wall tube to prevent the sodium-water reaction. However, such a method is not developed as yet. An alternative method is being used to improve the reliability of a steam generator by performing a non-destructive test of a double wall tube during the refueling period of a reactor. In this method a straight double wall tube is employed to perform this test easily, but has a difficulty regarding an absorption of a thermal expansion of the used materials. If an on-line leak detection method is developed, the demerits of a straight double-wall tube are avoided by using a helical type double-wall tube, and the probability of a sodium-water reaction can be reduced to a level less than the design-based accident.

  2. Independent tube verification and dynamic tracking in et inspection of nuclear steam generator

    International Nuclear Information System (INIS)

    Xiongzi, Li; Zhongxue, Gan; Lance, Fitzgibbons

    2001-01-01

    The full text follows. In the examination of pressure boundary tubes in steam generators of commercial pressurized water nuclear power plants (PWR's), it is critical to know exactly which particular tube is being accessed. There are no definitive landmarks or markings on the individual tubes. Today this is done manually, it is tedious, and interrupts the normal inspection work, and is difficult due to the presence of water on the tube surface, plug ends instead of tube openings in the field of view, and varying lighting quality. In order to eliminate the human error and increase the efficiency of operation, there is a need to identify tube position during the inspection process, independent of robot encoder position and motion. A process based on a Cognex MVS-8200 system and its application function package has been developed to independently identify tube locations. ABB Combustion Engineering Nuclear Power's Outage Services group, USPPL in collaboration with ABB Power Plant Laboratories' Advanced Computers and Controls department has developed a new vision-based Independent Tube Verification system (GENESIS-ITVS-TM ). The system employ's a model-based tube-shape detection algorithm and dynamic tracking methodology to detect the true tool position and its offsets from identified tube location. GENESIS-ITVS-TM is an automatic Independent Tube Verification System (ITVS). Independent tube verification is a tube validation technique using computer vision, and not using any robot position parameters. This process independently counts the tubes in the horizontal and vertical axes of the plane of the steam generator tube sheet as the work tool is moved. Thus it knows the true position in the steam generator, given a known starting point. This is analogous to the operator's method of counting tubes for verification, but it is automated. GENESIS-ITVS-TM works independent of the robot position, velocity, or acceleration. The tube position information is solely obtained from

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1964-07-01

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

  4. 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

  5. Evaluation of EDTA based chemical formulations for the cleaning of monel-400 tubed steam generators

    International Nuclear Information System (INIS)

    Velmurugan, S.; Rufus, A.L.; Sathyaseelan, V.S.; Kumar, P.S.; Veena, S.N.; Srinivasan, M.P.; Narasimhan, S.V.

    1998-01-01

    The Steam Generator (SG) is an important component in any nuclear power plant which contributes significantly for the over all performance of the reactor. The failure of SG tubes occurs mainly by corrosion under accelerated conditions caused by fouling. There is continuous ingress of the corrosion products and ionic impurities from the condenser and feed train of the secondary heat transfer system. The corrosion products accumulate in the stagnant areas near the tube sheet, over the tube support plates and in the tube to tube support plate crevices. These accumulated deposits help to concentrate the aggressive impurities and induce a variety of corrosion processes affecting the structural materials and finally leading to failure of the SG tube. Scale forming impurities can deposit over the tube surfaces and result in reduction of heat transfer efficiency and over heating of the surfaces. Every effort is being made to control the transport of impurities to the steam generator. Increased blow down, installation of condensate polishers and use of all volatile amines have helped to reduce the corrosion product and ionic impurities input into the steam generators of PHWRs. Despite these efforts, failures of SG tubes in PHWRs have been reported. Hence, attempts are being made to develop chemical formulations to clean the deposits accumulated in the steam generators. The EPRI-SGOG chemical cleaning process has been tried with good success in steam generators of different designs including the steam generators of PHWRs. This paper discusses the work on the evaluation of EDTA based chemical cleaning formulations for monel-400 tubed steam generators of PHWRs. (author)

  6. 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

  7. Reliability of double-wall-tube steam generator for FBR considering water leak accident frequency

    International Nuclear Information System (INIS)

    Ueda, Nobuyuki; Kinoshita, Izumi; Nishi, Yoshihisa

    2000-01-01

    For early realization, a fast breeder reactor (FBR) is required to reduce construction cost. A reactor concept in which the intermediate heat transport system is eliminated by introducing a double-wall-tube steam generator is one convincing approach. The reliability of the double-wall-tube SG in a water leak accident (sodium-water reaction accident) due to tube failure is strongly related to the mitigating system design. The safety design of the double-wall-tube SG approach is investigated to limit the accident occurrence below 10 -7 (1/ry. A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to effectively prevent adhesion of the double-wall-tube. The reliability of the tube-to-tube plate was evaluated at 10 -10 (l/hr) for an inner tube and 10 -9 (l/hr) for an outer with reference to the failure experience of previous SGs. The failure must be detected within 30 to 60 minutes. (author)

  8. Steam generator tubes integrity: In-service-inspection

    International Nuclear Information System (INIS)

    Comby, R.J.

    1997-01-01

    The author's approach to tube integrity is in terms of looking for flaws in tubes. The basis for this approach is that no simple rules can be fixed to adopt a universal inspection methodology because of various concepts related to experience, leak acceptance, leak before break approach, etc. Flaw specific management is probably the most reliable approach as a compromise between safety, availability and economic issues. In that case, NDE capabilities have to be in accordance with information required by structural integrity demonstration. The author discusses the types of probes which can be used to search for flaws in addition to the types of flaws which are being sought, with examples of specific analysis experiences. The author also discusses the issue of a reporting level as it relates to avoiding false calls, classifying faults, and allowing for automation in analysis

  9. A State of the Art Report on Wear Damage of Steam Generator Tubes

    International Nuclear Information System (INIS)

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

    2004-10-01

    The recent status on wear damage of steam generator tubes caused by flow-induced vibration was investigated, and the criteria for structural integrity evaluation of the wear-damaged tubes were reviewed. It was surveyed how the wear damage of tubes could be affected by main parameters, such as, materials properties and their combination, impact load and vibration amplitude/frequency, contact areas and diametral clearance between the tube and tube support plate, wear test duration, and test temperature. Finally, corrosive wear, which means the combined action of corrosion and wear simultaneously, was also surveyed in this report. There has been only a few works concerned on the wear damage of steam generator tubes in Korea, compared with the leading foreign research institutes. Especially, the experience related to the wear characteristics of Alloy 690, which has become a replacement material for Alloy 600 as steam generator tubes, is far from satisfactory. Systematic studies, therefore, concerned with structural integrity of tubes as well as improvement of were resistance of Alloy 690 in the PWR environment are needed

  10. 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)

  11. Laminar fluid flow and heat transfer in a fin-tube heat exchanger with vortex generators

    Energy Technology Data Exchange (ETDEWEB)

    Yanagihara, J.I.; Rodriques, R. Jr. [Polytechnic School of Univ. of Sao Paolo, Sao Paolo (Brazil). Dept. of Mechanical Engineering

    1996-12-31

    Development of heat transfer enhancement techniques for fin-tube heat exchangers has great importance in industry. In recent years, heat transfer augmentation by vortex generators has been considered for use in plate fin-tube heat exchangers. The present work describes a numerical investigation about the influence of delta winglet pairs of vortex generators on the flow structure and heat transfer of a plate fin-tube channel. The Navier-Stokes and Energy equations are solved by the finite volume method using a boundary-fitted coordinate system. The influence of vortex generators parameters such as position, angle of attack and aspect ratio were investigated. Local and global influences of vortex generators in heat transfer and flow losses were analyzed by comparison with a model using smooth fin. The results indicate great advantages of this type of geometry for application in plate fin-tube heat exchangers, in terms of large heat transfer enhancement and small pressure loss penalty. (author)

  12. Laminar fluid flow and heat transfer in a fin-tube heat exchanger with vortex generators

    Energy Technology Data Exchange (ETDEWEB)

    Yanagihara, J I; Rodriques, R Jr [Polytechnic School of Univ. of Sao Paolo, Sao Paolo (Brazil). Dept. of Mechanical Engineering

    1997-12-31

    Development of heat transfer enhancement techniques for fin-tube heat exchangers has great importance in industry. In recent years, heat transfer augmentation by vortex generators has been considered for use in plate fin-tube heat exchangers. The present work describes a numerical investigation about the influence of delta winglet pairs of vortex generators on the flow structure and heat transfer of a plate fin-tube channel. The Navier-Stokes and Energy equations are solved by the finite volume method using a boundary-fitted coordinate system. The influence of vortex generators parameters such as position, angle of attack and aspect ratio were investigated. Local and global influences of vortex generators in heat transfer and flow losses were analyzed by comparison with a model using smooth fin. The results indicate great advantages of this type of geometry for application in plate fin-tube heat exchangers, in terms of large heat transfer enhancement and small pressure loss penalty. (author)

  13. Post-failure metallurgical investigation of KNK steam generator tube damage

    Energy Technology Data Exchange (ETDEWEB)

    Lorenz, H; Herberg, G

    1975-07-01

    In September 1973 the sodium-cooled reactor KNK was shut down due to a steam generator tube damage. Failure location and results of the metallurgical examination of the damage are described. The cause of the damage is discussed. (author)

  14. Steam generator tube integrity requirements and operating experience in the United States

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    2009-01-01

    Steam generator tube integrity is important to the safe operation of pressurized-water reactors. For ensuring tube integrity, the U.S. Nuclear Regulatory Commission uses a regulatory framework that is largely performance based. This performance-based framework is supplemented with some prescriptive requirements. The framework recognizes that there are three combinations of tube materials and heat treatments currently used in the United States and that the operating experience depends, in part, on the type of material used. This paper summarizes the regulatory framework for ensuring steam generator tube integrity, it highlights the current status of steam generators, and it highlights some of the steam generator issues and challenges that exist in the United States. (author)

  15. A survey on the corrosion susceptibility of Alloy 800 CANDU steam generator tubing materials

    International Nuclear Information System (INIS)

    Lu, Y.C.; Dupuis, M.; Burns, D.

    2008-01-01

    To provide support for a proactive steam generator (SG) aging management strategy, a survey on the corrosion susceptibility of the archived Alloy 800 tubing from CANDU SGs under plausible crevice chemistry conditions was conducted to assess the potential material degradation issues in CANDU SGs. Archived Alloy 800 samples were collected from four CANDU utilities. High-temperature electrochemical analysis was carried out to assess the corrosion susceptibility of the archived SG tubing under simulated CANDU crevice chemistry conditions at both 150 o C and 300 o C. The potentiodynamic polarization results obtained from the archived CANDU SG tubes were compared to the data from ex-service tubes removed from Darlington Nuclear Generating Station (DNGS) SGs and a reference nuclear grade Alloy 800 tubing. It was found that the removed Darlington SG tubes, with signs of in-service degradation, were more susceptible to pitting corrosion than the reference nuclear grade Alloy 800 tubing. At 150 o C, under the same neutral crevice chemistry conditions, the potentiodynamic polarization curve of the ex-service Darlington SG tubing has an active peak, which is a sign of propensity to crevice/underdeposit corrosion. This active peak was not observed in any of the potentiodynamic polarization curves of all archived Alloy 800 CANDU SG tubing indicating that archived CANDU SG tubes are less susceptible to the underdeposit corrosion under SG startup conditions. The corrosion behaviour of the archived Alloy 800 tubes from CANDU SG was similar to that of the reference nuclear grade Alloy 800 tubing. The results of this survey suggest that the Alloy 800 tubing materials used in the existing CANDU utilities (other than ex-service DNGS tubing) will continue to have reliable performance under specified CANDU operating conditions. Ex-service SG tubing from DNGS, although showing lower than average corrosion resistance, still has a wide acceptable operating margin and the in

  16. 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

  17. 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)

  18. Strategy for assessment of WWER steam generator tube integrity. Report prepared within the framework of the coordinated research project on verification of WWER steam generator tube integrity

    International Nuclear Information System (INIS)

    2007-12-01

    Steam generator heat exchanger tube degradations happen in WWER Nuclear Power Plant (NPP). The situation varies from country to country and from NPP to NPP. More severe degradation is observed in WWER-1000 NPPs than in case of WWER-440s. The reasons for these differences could be, among others, differences in heat exchanger tube material (chemical composition, microstructure, residual stresses), in thermal and mechanical loadings, as well as differences in water chemistry. However, WWER steam generators had not been designed for eddy current testing which is the usual testing method in steam generators of western PWRs. Moreover, their supplier provided neither adequate methodology and criteria nor equipment for planning and implementing In-Service Inspection (ISI). Consequently, WWER steam generator ISI infrastructure was established with delay. Even today, there are still big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment (plugging criteria for defective tubes vary from 40 to 90% wall thickness degradation). Recognizing this situation, the WWER operating countries expressed their need for a joint effort to develop methodology to establish reasonable commonly accepted integrity assessment criteria for the heat exchanger tubes. The IAEA's programme related to steam generator life management is embedded into the systematic activity of its Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP). Under the advice of the TWG-LMNPP, an IAEA coordinated research project (CRP) on Verification of WWER Steam Generator Tube Integrity was launched in 2001. It was completed in 2005. Thirteen organizations involved in in-service inspection of steam generators in WWER operating countries participated: Croatia, Czech Republic, Finland, France, Hungary, Russian Federation, Slovakia, Spain, Ukraine, and the USA. The overall objective was to

  19. Round robin tests of the PISC III programme on defective steam generators tubes

    International Nuclear Information System (INIS)

    Birac, C.; Herkenrath, H.; Crutzen, S.; Miyake, Y.; Maciga, G.

    1991-11-01

    The PISC III actions are intended to extend the results and methodologies of the previous PISC exercises, i.e. the assessment of the capabilities of the various examination techniques when used on real or realistic flaws in real components under real conditions of inspection. Being aware of the industrial problems that the degradation of steam generator tubes can create, the PISC III management board decided to include in the PISC III programme a special action on steam generator tubes testing (SGT). (author)

  20. A Comparison of the Predicted Tube Plugging Rate for Alloy 600HTMA Steam Generator

    Energy Technology Data Exchange (ETDEWEB)

    Boo, Myung Hwan; Kang, Yong Seok [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2010-10-15

    To manage components that are used in long term operations such as steam generation, it is important to know the tube plugging rate, which can cause the performance degradation. The life of components can be predicted by the method using determinism and probability theory. With a method using probability theory, damage prediction of tube is possible. In this study, damage prediction for steam generation (SG) tube is performed using Weibull distribution and predicted plugging rate (life) is compared with the simple sum plugging number and case by case (failure cause) plugging number

  1. Determination of thermal characteristics of combustion products of fire-tube heat generator with flow turbulator

    OpenAIRE

    Lukjanov Alexander V.; Ostapenko Dmitry V.; Basist Dmitry V.

    2014-01-01

    Boiler construction is one of the major industries of any state. The aim is to determine the effect of the turbulator on the intensity of heat transfer in the convective part of the fire-tube heat generator of domestic production. The improvement of convective heating surfaces is one of the ways to increase the energy efficiency of the fire-tube heat generator. Since model of the process of heat transfer of gas flow in the convective tubes is multifactorial and does not have clear analytical ...

  2. New Media: Engaging and Educating the YouTube Generation

    Directory of Open Access Journals (Sweden)

    Anu Vedantham

    2011-12-01

    Full Text Available Today's undergraduates are clearly comfortable as consumers of technology and new media—purchasing ring tones for their cell phones and tunes for their iPods, text-messaging from handheld devices, scanning and tinkering with photos, keeping up with their Facebook friends and watching viral YouTube videos, sometimes all simultaneously. We share examples of classroom assignments integrated with library support services that engage today's undergraduates with academic materials in a variety of course contexts. We discuss how specific arrangements of library learning spaces and the alignment of space and staffing can help undergraduate students succeed with new media projects for class assignments.

  3. Monitoring on corrosion behavior of steam generator tubings

    International Nuclear Information System (INIS)

    Takamatsu, H.; Isobe, S.; Sato, M.; Arioka, K.; Tsuruta, T.

    1988-01-01

    The importance of chemistry in high temperature aqueous solutions is widely recognized for understanding corrosion phenomena in PWR SG crevice environments. Potential and pH are two important parameters, among other environmental factors affecting localized corrosion processes, such as IGA and/or SCC in SG crevices. In this article, we discuss the potential-pH-IGA/SCC diagram of Alloy 600 as a basis for evaluating the corrosion behavior of SG tubings, and two examples of monitoring, corrosion potential monitoring in the bulk secondary water and pH monitoring in simulated SG crevices. (author)

  4. Leak behavior of steam generator tube-to-tubesheet joints under creep condition: Experimental study

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Majumdar, Saurin; Kasza, Ken E.; Shack, William J.

    2013-01-01

    To address concerns regarding excessive leakage from throughwall cracks in steam generator tube-to-tubesheet joints under severe accident conditions, leak rate testing was conducted using tube-to-collar joint specimens. The tube interior and the interface between tube and collar (crevice) were pressurized independently using nitrogen gas. The leak rate through the crevice was almost zero when the specimens were pressurized at ∼500 °C; this low leak rate is attributed to thermal mismatch effects preventing much leakage. The near zero leak rate was maintained until the onset of large leakage at higher temperatures. The leak rate behavior after the onset of the large leakage was not much affected by the crevice length or heat-to-heat variation of Alloy 600 tubes. This suggests that once the crevice gap opens, the creep rate of the low alloy steel collar becomes dominant. Specimens with different tube diameters behaved essentially the same way. To simulate a flawed steam generator tube in the tubesheet, the crevice region was pressurized through a hole in the tube. This simulation resulted in essentially the same behavior as those specimens whose tubes and crevices were pressurized independently. Oxidation of low alloy steel collars in air tests can increase the flow resistance, and thus tests using nitrogen gas would provide more conservative leak rate data. Highlights: ► Leak rates were measured by using tube-to-collar joint specimens under creep condition. ► Leak rate through the joint interface was almost zero at ∼500 °C due to thermal mismatch. ► The near zero leak rate was maintained until the onset of large leakage at ∼680 °C. ► The leak behavior after the onset of the large leakage was not affected by hydraulic expansion length or tube heats.

  5. Ultrasonic imaging of tube/support structure of power plant steam generators

    International Nuclear Information System (INIS)

    Saniie, J.; Nagle, D.T.

    1987-01-01

    The corrosion and erosion of steam generator tubing in nuclear power plants can present problems of both safety and economics. In steam generators, the inconel tubes are fit loosely through holes drilled in carbon steel support plates. Corrosion is of particular concern with such tube/support plate structures. Non-protective magnetite can build up on the inner surface of the support plate holes, and allowed to continue unchecked, will fill the gap, eventually denting and fracturing the tube walls. Therefore, periodic nondestructive inspection can be valuable in characterizing corrosion and can be used in evaluating the effectiveness of chemical treatments used to control or reduce corrosion. Presently, they are investigating the feasibility and practicality of using ultrasound in routing testing for gap measurement, for evaluating the corrosion and assessing the degree of denting. The tube/support structure can be modeled as a multilayer, reverberant target, which when tested with ultrasound results in two sets of reverberating echoes [1]. One set corresponds to the tube wall and the other to the support plate. These echoes must be decomposed and identified in order to evaluate the tube/support structure. This report presents experimental results along with a discussion of various measurements and processing techniques for decomposing and interpreting tube/support echoes at different stages of corrosion

  6. Residual stresses associated with the hydraulic expansion of steam generator tubing into tubesheets

    International Nuclear Information System (INIS)

    Middlebrooks, W.B.; Harrod, D.L.; Gold, R.E.

    1991-01-01

    Westinghouse has used three different processes for the full depth expansion of tubes into the tube sheets of recirculating nuclear steam generators: mechanical rolling, explosive expansion and hydraulic expansion. Each process aims at expanding tubes tightly to tube sheets, leaving the smallest possible secondary side crevice depth, and minimizing the residual stress in the expanded tubes, all for the purpose of mitigating the effect of corrosion phenomena. The hydraulic expansion process was qualified and has been implemented since 1978, and more than 1.1 million tube ends have been hydraulically expanded into production units. In this paper, the results of the recent analytical studies related to the residual stress in the expanded tubes are summarized. The method of hydraulic expansion is explained, and some important parameters are given. Finite element method, theoretical incremental analysis, tube sheet yielding and residual stress, contact pressure, sensitivity analysis and temperature effect in the central region of tube sheets, and the residual stress in the transition zone are described. (K.I.)

  7. 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.

  8. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    Langford, O.M.; Peelman, H.E.

    1980-01-01

    A gas filled neutron tube in a nuclear well logging tool has a target an ion source voltage and a replenisher connected to ground. A negative high voltage is applied to the target by a power supply also providing a target current corresponding to the neutron output of the neutron generator tube. A constant current source provides a constant current. A network receiving the target current and the constant current provides a portion of the constant current as a replenisher current which is applied to the replenisher in a neutron generating tube. The network controls the magnitude of the replenisher current in accordance with the target current so as to control the neutron output of the neutron generating tube. (auth)

  9. 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

  10. Evaluation and field validation of Eddy-Current array probes for steam generator tube inspection

    International Nuclear Information System (INIS)

    Dodd, C.V.; Pate, J.R.

    1996-07-01

    The objective of the Improved Eddy-Current ISI for Steam Generator Tubing 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 describes the design of specialized high-speed 16-coil eddy-current array probes. Both pancake and reflection coils are considered. Test results from inspections using the probes in working steam generators are given. Computer programs developed for probe calculations are also supplied

  11. 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

  12. Removal of portions of tubes from steam generator of nuclear reactor

    International Nuclear Information System (INIS)

    Wilkins, R.L.; Williams, C.F.

    1983-01-01

    After the tube portion to be removed is severed from the remainder of the U-tube and its weld to the header is machined off, the internal surface of the portion is engaged internally by an ID gripper and pulled out of the header. Then the external surface is engaged by an OD gripper and pulled further out of the header. The first tube length is pulled out as far as the space under the header permits and is then cut off. Successive lengths are likewise pulled out and cut off. The apparatus for accomplishing this object includes a base secured to the header by expanded mandrel mechanisms. A carriage is suspended from the base on screws which are driven by a motor to move the carriage away from and towards the base. An OD gripper assembly is suspended from the carriage and is movable by fluoroactuated piston rods away from and towards the carriage. An ID gripper assembly extends through the OD gripper assembly. The gripper of the ID assembly is actuable to engage the internal surface of the tube portion. With its gripper so engaged the ID assembly is engaged by the gripper of the OD assembly and the engaged tube portion is pulled out of the header by the OD assembly. The ID gripper is then disengaged and the OD gripper is engaged with the tube portion in the same way that it engages the ID assembly and the tube portion is pulled out further. The apparatus also includes a tube cutter having an abrasive wheel. The wheel cuts the lengths of the tube portion at an angle so that for examination and testing the tube lengths can be matched and the orientation of any defect with respect to the plate in the steam generator which separates the inlet and outlet ends of the tubes and the U-tube supports can be identified

  13. 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)

  14. Wear behavior of steam generator tubes in nuclear power plant operating condition

    International Nuclear Information System (INIS)

    Kim, In-Sup; Hong, Jin-Ki; Kim, Hyung-Nam; Jang, Ki-Sang

    2003-01-01

    Reciprocating sliding wear tests were performed on steam generator tubes materials at steam generator operating temperature. The material surfaces react with oxygen to form oxides. The oxide properties such as formation rate and mechanical properties are varied with the test temperature and alloy composition. So, it is important to investigate the wear properties of each steam generator tube materials in steam generator operating condition. The tests results indicated that the wear coefficient in work rate model of alloy 690 was faster than that of alloy 800. From the scanning electron microscopy observation, the wear scars were similar each other and worn surfaces were covered with oxide layers. It seemed that the oxide layers were formed by wear debris sintering or cold welding and these layer properties affected the wear rate of steam generator tube materials. (author)

  15. Steam generator tube performance: world experience with water-cooled nuclear power reactors during 1979

    International Nuclear Information System (INIS)

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

    1981-01-01

    The performance of steam generator tubes in water-cooled nuclear power reactors is reviewed for 1979. Tube failures occurred at 38 of the 93 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The defect rate, although higher than that in 1978, was still lower than the rates of the two previous years. Methods being employed to detect defects include the increased use of multifrequency eddy-current testing and a trend to full-length inspection of all tubes. To reduce the incidence of tube failure by corrosion, plant operators are turning to full-flow condensate demineralization and more leak-resistant condenser tubes. 10 tables

  16. Stress identification in steam generator tubes from profile measurements

    International Nuclear Information System (INIS)

    Andrieux, S.; Voldoire, F.

    1993-01-01

    An identification method devoted to the determination of stresses in tubes, by means of profile measurements, provided by on site non-destructive evaluations, is presented here. From the only available data (the radial displacement w on the inner wall), the computation of the strains, and consequently the stresses in the elastoplastic range, is made within the framework of the shell theory. For this purpose, we need to determine the associated curvature w'': this step is an ill-posed problem, because of the lack of continuity with respect to the discrete data. This difficulty is overridden by means of an appropriate regularization procedure. The predictive ability of the method has been tested by comparison with direct simulations; we present an industrial application. (author)

  17. SCC of Alloy 600 in PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pascali, R.; Buzzanca, G.; Quaglia, G.M.; Ronchetti, C.

    1986-01-01

    The studies reported in this paper concern the evaluation of Alloy 600 and 690 behaviour in chemical agressive conditions simulating the concentration film on heat exchanging tube. The corrosion tests have been performed to evidence the influence of metallurgical conditions and different heats. Various devices for reproducing dead areas and steam blanketing have been designed and tested, such as, umbrellas, rings, thin deposits, etc. A system to reproduce the S.G. areas with thick deposits has been designed successively and set up in a previous series of tests, in boiling water at 56 kg/cm/sup 2/, 270 0 C and heat flux 45 W/cm/sup 2/. Caustic SCC tests have been carried out in adiabatic conditions also using small autoclaves

  18. 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)

  19. Evaluation of tube to collector connection by hydraulic expansion method in PGV-1000 steam generators

    International Nuclear Information System (INIS)

    Dashti, H.G.; Hashemi, B.; Jahromi, S.A.

    2011-01-01

    Research highlights: → The produced residual stresses in the collector body due to hydraulic expansion method have been compared with explosive method. → The residual stresses were obtained using two methods of FEM and strain gauging tests. → The effect of clearance between tube and collector on the residual stresses was investigated. → The contact stresses between the tube and collector interface were modeled and the required connection strength between tube and collector is estimated based on ASME rules and compared with FE results. - Abstract: Investigations on steam generators failure due to cracking in collector ligaments at perforated parts determined that connection process of the tubes to collector could be one of the main breakdown causes. The stability and strength of tube to collector joint is dependent to the geometry of tube and collector, the joining process and the operational conditions. In this research hydraulic expansion method has been considered as connection method of tube to collector. The Finite Element Method (FEM) was used to simulate the hydraulic expansion process and determine stress condition of the joints. The contact stresses between the tube and collector interface were modeled using contact elements of ANSYS program. Furthermore, the effect of clearance between tube and collector on the residual stresses around of joints was investigated. Some specimens from collector and tube materials were tested at various temperatures and their results were used at rate-independent multi-linear Mises plasticity model for FE analysis. Required connection strength between tube and collector is estimated based on ASME rules and compared with FE results. The results show that the residual tensile stresses could be greatly increased by decreasing of initial clearance. The highest value of residual stresses was observed around of collector holes nevertheless it was considerably lesser than obtained residual stresses in explosive method. The

  20. 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

  1. 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)

  2. A heat transfer study for vertical straight-tube steam generators heated by liquid metal

    International Nuclear Information System (INIS)

    Valette, M.

    1984-04-01

    A single-tube mockup of a vertical straight-tube steam generator heated by sodium-potassium alloy NaK was submitted to thermal and hydraulic testing in conditions representative of fast breeder reactor operation. The mockup consisted of a 10mm I.D. ferritic steel heat exchange tube centered inside a cylindrical stainless steel shell. The complete assembly was 20.9 meters long. Water flowed upward inside the exchange tube, and NaK flowed downward in the annular gap between the tube and the shell. The steam outlet pressure ranged from 90 to 195 bars, while the liquid metal temperature at the mockup inlet was between 480 and 580 0 C. The water flowrate in the tube ranged from 153 to 2460 kg.m -2 .s -1 . During the tests the fluid inlet and outlet temperatures, flowrate and pressures were measured, as was the NaK temperature profile over the full length of the device. The test results were subsequently compared with heat exchange and pressure drop values calculated using the standard formulas for straight-tube heat exchangers. The heat exchange coefficients predicted by these correlations in the boiling zone were found to be largely overestimated, while the calculated pressure drop values proved satisfactory. A set of modified correlations is proposed to account for the observed phenomena, and for use in designing commercial units, provided the sodium flow in the tube bundle is adequately distributed

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  4. A calculating method of tube-to-tubesheet joints design for steam generator

    International Nuclear Information System (INIS)

    Zhang Fuyuan

    1993-01-01

    A theoretical calculating method of the hydraulically expanded tube-to-tubesheet joints design is described. As a mathematical model, the total expanded process of the joints is divided in four stages. with the elastic and plastic theories, the stress, strain and displacement of the tube or tube and tubesheet are analysed by stages, then expansion pressure, deformation, residual stress and push-out force are evaluated. The method may be used to design the steam generators and steel tubular heat exchangers. The paper points out that the hydraulic-expansion plus local roller expansion (hybrid expansion) is better than the only hydraulic-expansion for the tube-to-tubesheet joints of the nuclear steam generators

  5. Steady-state heat transfer in an inverted U-tube steam generator

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1987-01-01

    Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during stead-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The Semiscale (MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

  6. 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)

  7. 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)

  8. Study on thermal and mechanical properties of U-tube materials for steam generator

    International Nuclear Information System (INIS)

    Rheu, Woo Suk; Kang, Young Hwan; Park, Jong Man; Joo, Ki Nam; Kim, Sung Soo; Maeng, Wan Young; Park, Se Jin

    1993-01-01

    Most of domestic nuclear plants have used I600 TT material for steam generator U-tube, and piled up the field experience. I600 HTMA and I690 TT, however, are recommended for an alternative of U-tube by ABB-CE since YK-3 and 4. Field experience of I600 HTMA and I690 TT have not compiled in the country, so it is concerned to select the future materials for U-tube. Thus, database on the thermal and mechanical properties of U-tube materials is very necessary for design documentations. In this study, the thermal, mechanical and metallugical properties were tested and evaluated to establish the database for steam generator U-tube. In addition, thermal conductivity of I600 and I690 was measured and compared statistically, providing a basic document for applying I690 to U-tube. The results will be used to improve the manufacturing process in order to increase the integrity of U-tube. (Author)

  9. How to operate safely steam generators with multiple tube through-wall defects

    International Nuclear Information System (INIS)

    Hernalsteen, P.

    1993-01-01

    For a Nuclear Power Plant (NPP) of the Pressurized Water Reactor (PWR) type, the Steam Generator (SG) tube bundle represents the major but also the thinnest part of the primary pressure boundary. To the extent that no tube material has yet been identified to be immune to corrosion, defects may initiate in service and easily propagate through wall. While not a desirable feature, a Through Wall Deep (TWD) defect does not necessarily pose a threat to either the structural integrity or leaktightness and this paper shows how SG can (and indeed, do) operate safely and reliably while having many tubes affected by deep and even TWD defects

  10. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

    International Nuclear Information System (INIS)

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

    1980-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

  11. Process and device for removing sludge deposited on the tube plate of a steam generator

    International Nuclear Information System (INIS)

    Charamathieu, A.; Dessales, J.; Lebouc, B.

    1983-01-01

    To remove the sludges on the tubular plate, one lance, at least, is moved radially from the center of the tubular plate between two rows of tubes, in a parallel direction to the tubular plate and near this one. Two high pressure jets are moved from the extremity of the lance and, in fixed and symmetrical directions about the direction of the rows. The two jets are interrupted when passing in front of the heat exchange tubes of the generator; the cleaning liquid is simultaneously carried off from the periphery of the group of tubes [fr

  12. Stability of single-phase natural circulation with inverted U-tube steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, J.

    1988-08-01

    For natural circulation it is shown that parallel flow in the tubes of an inverted U-tube stream generator can be, at certain power levels, unstable. A mathematical model, based on one-dimensional Oberbeck-Boussinesq equations, shows that stability can be attained if in some tubes the water flows backward, opposite to the normal flow direction. The results are compared to measurements obtained from the natural circulation test A2-77A in the LOBI-MOD2 integral system test facility.

  13. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    1980-01-01

    A specification is given for an energizing and regulating circuit for a gas filled neutron generator tube consisting of a target, an ion source and a replenisher, the circuit consisting of a power supply to provide a negative high voltage to the target and a target current corresponding to the neutron output of the tube, a constant current source, and control means connected to the power supply and to the constant current source, the control means being responsive to the target current to provide a portion of the constant current to the replenisher substantially to regulate the neutron output of the tube. (author)

  14. Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

    International Nuclear Information System (INIS)

    Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

    2006-01-01

    It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

  15. 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

  16. Mode Selection for Axial Flaw Detection in Steam Generator Tube Using Ultrasonic Guided Wave

    International Nuclear Information System (INIS)

    Yoon, Byung Sik; Yang, Seung Han; Guon, Ki Il; Kim, Yong Sik

    2009-01-01

    The eddy current testing method is mainly used to inspect steam generator tube during in-service inspection period. But the general problem of assessing the structural integrity of the steam generator tube using eddy current inspection is rather complex due to the presence of noise and interference signal under various conditions. However, ultrasonic testing as a nondestructive testing tool has become quite popular and effective for the flaw detection and material characterization. Currently, ultrasonic guided wave is emerging technique in power industry because of its various merits. But most of previous studies are focused on detection of circumferential oriented flaws. In this study, the steam generator tube of nuclear power plant was selected to detect axially oriented flaws and investigate guided wave mode identification. The longitudinal wave mode is generated using piezoelectric transducer frequency from 0.5 MHz, 1.0 MHz, 2.25MHz and 5MHz. Dispersion based STFT algorithm is used as mode identification tool

  17. Optimization of Peripheral Finned-Tube Evaporators Using Entropy Generation Minimization

    OpenAIRE

    Pussoli, Bruno; Barbosa Jr., Jader; da Silva, Luciana; Kaviany, Massoud

    2012-01-01

    The peripheral finned-tube (PFT) is a new geometry for enhanced air-side heat transfer under moisture condensate blockage (evaporators). It consists of individual hexagonal (peripheral) fin arrangements with radial fins whose bases are attached to the tubes and tips are interconnected with the peripheral fins. In this paper, experimentally validated semi-empirical models for the air-side heat transfer and pressure drop are combined with the entropy generation minimization theory to determine ...

  18. Steam generator tube performance: experience with water-cooled nuclear power reactors during 1977

    International Nuclear Information System (INIS)

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

    1979-02-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1977. Failures were reported in 34 of the 79 reactors surveyed. Causes of these failures and inspection and repair procedures designed to deal with them are presented. Although corrosion remained the leading cause of tube failures, specific mechanisms have been identified and methods of dealing with them developed. These methods are being applied and should lead to a reduction of corrosion failures in future. (author)

  19. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

    Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

  20. Improved in-service inspection program for management of degradation in steam generator tubing

    International Nuclear Information System (INIS)

    Kurtz, R.; Heasler, P.; Muscara, J.

    1992-01-01

    This paper presents an overview of significant results from NRC-sponsored research on steam generator tube integrity and inspection. Burst test results are described along with empirical models to relate flaw geometry and size to tube burst pressure. Results of round robin examinations of a retired-from-service steam generator to determine eddy current inspection reliability are presented. An evaluation and comparison of various sampling plans for in-service inspection of steam generators is discussed. Finally, performance demonstration qualification efforts for eddy current inspection systems are described

  1. Technical basis for the CANDU steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Kozluk, M.J.; Scarth, D.A.; Graham, D.B.

    2002-01-01

    Active degradation mechanisms in steam generators and preheaters in Canadian CANDU T M generating stations are managed through Steam Generator Programs that incorporate tube inspection, maintenance (cleaning), fitness-for-service assessment, and preventative plugging as part of the overall steam generator management strategy. Steam generator and preheater tubes are inspected in accordance with the CSA Standard CAN/CSA-N285.4-94[l]. When a detected flaw indication does not satisfy the criteria of acceptance by examination, CSA-N285.4-94 permits a fitness-for-service assessment to determine acceptability. In 1999 Ontario Power Generation issued, for trial use, fitness-for-service guidelines for steam generator and preheater tubes in CANDU nuclear power plants. The main objectives of the Fitness-for-Service Guidelines are to provide reasonable assurance that tube structural integrity is maintained, and to provide reasonable assurance that there are adequate margins between estimated accumulated dose and applicable site dose limits. The Fitness-for-Service Guidelines are intended to provide industry-standard acceptance criteria and evaluation procedures for assessing the condition of steam generator and preheater tubes in terms of tube structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. This paper describes the technical basis for the minimum required safety factors specified in Table IC-1 of the Fitness-for-Service Guidelines and for the flaw models used to develop the flaw stability requirements in the nonmandatory, Appendix C of the Fitness-for-Service Guidelines. (author)

  2. Diagnostic of corrosion defects in steam generator tubes using advanced signal processing from Eddy current testing

    International Nuclear Information System (INIS)

    Formigoni, Andre L.; Lopez, Luiz A.N.M.; Ting, Daniel K.S.

    2009-01-01

    Recently, the Brazilian Angra I PWR nuclear power plant went into a programmed shutdown for substitution of its Steam Generator (SG) which life was shortened due to stress corrosion in its tubes. The total cost of investment were around R$724 million. The signals generated during an Eddy-current Testing (ECT) inspection in SG tubes of nuclear plant allows for the localization and dimensioning of defects in the tubes. The defects related with corrosion generate complex signals that are difficult to analyze and are the most common cause in SG replacement in nuclear power plants around the world. The objective of this paper is the development of a methodology that allows for the characterization of corrosion signals by ECT inspections applied in the heat exchangers tubes of SG of a nuclear power plant. In this present work, the aim is to investigate distributed type defects by inducing controlled corrosion in sample tubes of different materials The ECT signals obtained from these samples tubes with corrosion implanted, will be analyzed using Zetec ECT equipment, the MIZ-17ET and its probes. The data acquisition will use a NI PC A/D CARD 700 card and the LabVIEW program. Subsequently, we will apply mathematical tools for signal processing like time windowed Fast Fourier transforms and Wavelets transforms, in MATLAB platform, which will allow effectiveness to remove the noises and to extract representative characteristics for the defect being analyzed. Previously obtained results as well as the proposal for the future work will be presented. (author)

  3. 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

  4. 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

  5. Computing broadband accelerograms using kinematic rupture modeling; Generation d'accelerogrammes synthetiques large-bande par modelisation cinematique de la rupture sismique

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz Paredes, J.A

    2007-05-15

    In order to make the broadband kinematic rupture modeling more realistic with respect to dynamic modeling, physical constraints are added to the rupture parameters. To improve the slip velocity function (SVF) modeling, an evolution of the k{sup -2} source model is proposed, which consists to decompose the slip as a sum of sub-events by band of k. This model yields to SVF close to the solution proposed by Kostrov for a crack, while preserving the spectral characteristics of the radiated wave field, i.e. a w{sup 2} model with spectral amplitudes at high frequency scaled to the coefficient of directivity C{sub d}. To better control the directivity effects, a composite source description is combined with a scaling law defining the extent of the nucleation area for each sub-event. The resulting model allows to reduce the apparent coefficient of directivity to a fraction of C{sub d}, as well as to reproduce the standard deviation of the new empirical attenuation relationships proposed for Japan. To make source models more realistic, a variable rupture velocity in agreement with the physics of the rupture must be considered. The followed approach that is based on an analytical relation between the fracture energy, the slip and the rupture velocity, leads to higher values of the peak ground acceleration in the vicinity of the fault. Finally, to better account for the interaction of the wave field with the geological medium, a semi-empirical methodology is developed combining a composite source model with empirical Green functions, and is applied to the Yamaguchi, M{sub w} 5.9 earthquake. The modeled synthetics reproduce satisfactorily well the observed main characteristics of ground motions. (author)

  6. Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

    International Nuclear Information System (INIS)

    1980-03-01

    The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

  7. 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)

  8. High temperature technological heat exchangers and steam generators with helical coil assembly tube bundle

    International Nuclear Information System (INIS)

    Korotaev, O.J.; Mizonov, N.V.; Nikolaevsky, V.B.; Nazarov, E.K.

    1990-01-01

    Analysis of thermal hydraulics characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the basis of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given. The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the following types: straight tubes, field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conjunction with forced and natural mixing in connecting lines. Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers and steam generators within the wide range of operating conditions without additional expenses on the research and development work

  9. Determination of thermal characteristics of combustion products of fire-tube heat generator with flow turbulator

    Directory of Open Access Journals (Sweden)

    Lukjanov Alexander V.

    2014-12-01

    Full Text Available Boiler construction is one of the major industries of any state. The aim is to determine the effect of the turbulator on the intensity of heat transfer in the convective part of the fire-tube heat generator of domestic production. The improvement of convective heating surfaces is one of the ways to increase the energy efficiency of the fire-tube heat generator. Since model of the process of heat transfer of gas flow in the convective tubes is multifactorial and does not have clear analytical solution at present, the study of process above is carried out using the experimental method. The results of applying the flow turbulator as a broken tape in the fire-tube heat generator of KV-GM type are presented. On their basis it can be concluded about increasing of heat transfer in convective part of the unit. The use of efficient, reliable, easy to manufacture, relatively inexpensive turbulator in domestic fire-tube heat generators will allow to increase their energy conversion efficiency and reduce fuel consumption, which will have a positive economic effect.

  10. Simulating Porous Magnetite Layer Deposited on Alloy 690TT Steam Generator Tubes.

    Science.gov (United States)

    Jeon, Soon-Hyeok; Son, Yeong-Ho; Choi, Won-Ik; Song, Geun Dong; Hur, Do Haeng

    2018-01-02

    In nuclear power plants, the main corrosion product that is deposited on the outside of steam generator tubes is porous magnetite. The objective of this study was to simulate porous magnetite that is deposited on thermally treated (TT) Alloy 690 steam generator tubes. A magnetite layer was electrodeposited on an Alloy 690TT substrate in an Fe(III)-triethanolamine solution. After electrodeposition, the dense magnetite layer was immersed to simulate porous magnetite deposits in alkaline solution for 50 days at room temperature. The dense morphology of the magnetite layer was changed to a porous structure by reductive dissolution reaction. The simulated porous magnetite layer was compared with flakes of steam generator tubes, which were collected from the secondary water system of a real nuclear power plant during sludge lancing. Possible nuclear research applications using simulated porous magnetite specimens are also proposed.

  11. 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

  12. Predicting tube repair at French nuclear steam generators using statistical modeling

    Energy Technology Data Exchange (ETDEWEB)

    Mathon, C., E-mail: cedric.mathon@edf.fr [EDF Generation, Basic Design Department (SEPTEN), 69628 Villeurbanne (France); Chaudhary, A. [EDF Generation, Basic Design Department (SEPTEN), 69628 Villeurbanne (France); Gay, N.; Pitner, P. [EDF Generation, Nuclear Operation Division (UNIE), Saint-Denis (France)

    2014-04-01

    Electricité de France (EDF) currently operates a total of 58 Nuclear Pressurized Water Reactors (PWR) which are composed of 34 units of 900 MWe, 20 units of 1300 MWe and 4 units of 1450 MWe. This report provides an overall status of SG tube bundles on the 1300 MWe units. These units are 4 loop reactors using the AREVA 68/19 type SG model which are equipped either with Alloy 600 thermally treated (TT) tubes or Alloy 690 TT tubes. As of 2011, the effective full power years of operation (EFPY) ranges from 13 to 20 and during this time, the main degradation mechanisms observed on SG tubes are primary water stress corrosion cracking (PWSCC) and wear at anti-vibration bars (AVB) level. Statistical models have been developed for each type of degradation in order to predict the growth rate and number of affected tubes. Additional plugging is also performed to prevent other degradations such as tube wear due to foreign objects or high-cycle flow-induced fatigue. The contribution of these degradation mechanisms on the rate of tube plugging is described. The results from the statistical models are then used in predicting the long-term life of the steam generators and therefore providing a useful tool toward their effective life management and possible replacement.

  13. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1987-01-01

    This patent describes a pressurized water reactor steam generator having spaced rows of heat transfer tubes through which primary coolant from the reactor flows, the tubes being of a U-shaped design, with the U-bend portions of the U-shaped tubes stabilized by antivibration bars. The improvement described here comprises expandable antivibration bars for stabilizing the U-bend portions of the U-shaped tubes, the expandable bars having a pair of adjustable rods, formed from a pair of rod sections affixed to a connector, one rod section of each of the pair of rod sections having a plurality of protrusions. Each of the protrusions has slidable surfaces thereon. The other rod section of each of the pair of rod sections has indentations, each of the indentations having slidable surfaces thereon complementary to the sliding surfaces of the protrusions, such that the rods are expandable from a first cross-sectional width less than the spacing between two adjacent rows of the tubes, to a second cross-sectional width greater than the first cross-sectional width. The expanded rods are adapted to contact tubes of the two adjacent rows of the tubes

  14. An Investigation of Tube Wastage Caused by Small Leaks of Water in Sodium-Heated Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

    Muenchow, H. O.; Chase, W. L. [Atomic Power Development Associates, Inc., Detroit, MI (United States)

    1967-06-15

    Examination of the results from a sodium-water reaction that occurred in a steam generator in the Enrico Fermi Plant revealed a generalized corrosion in the area of the reaction, producing failure of some of the adjacent tubes. The questions raised by this experience clearly indicated that the problem of small leaks causing general tube wastage should be investigated. An experimental programme was designed and divided into two categories: (1) basic studies to establish the mechanisms of wastage, including wasted surface temperature measurements, chemical attack and hot erosion; (2) studies to establish the effect of design and operating parameters on wastage occurring in a test rig that simulates the dynamics of a commercial steam generator. A small test rig supporting a 4-ft-high reaction vessel containing 5 gal of sodium was constructed. Wasted surface temperatures measured in this rig using thermocouples as the target for the sodium - water reaction have indicated temperatures in excess of 2200 Degree-Sign F. Corrosion studies, using the reaction products NaOH and Na{sub 2}O, have shown severe attack on 24 Cr-1 Mo steel tabs in 1 h under an argon atmosphere from 2200 Degree-Sign F to 2400 Degree-Sign F. Evidence indicates that corrosion is (1) minimal in the gas space above the reaction products, (2) extremely severe at the gas-liquid interface, and (3) less severe in the molten reaction products. A 101-ft-high support structure was constructed to house a test rig designed to simulate the conditions in a commercial sodium-heated steam generator. The test rig circulates 5700 lb of sodium at 400 gal/min through a 16-ft-high reaction vessel. Tests conducted in this rig have shown that the circulating water coil, which served as the target tube during the sodium-water reaction, was wasted to the point of rupturing, causing a large secondary reaction and additional damage to surrounding material. The temperature recorder went off scale at 2200 Degree-Sign F

  15. Review on steam generator tube failure propagation study in Japan

    International Nuclear Information System (INIS)

    Tanabe, Hiromi; Wachi, Eiji

    1990-01-01

    Major objectives of a failure propagation study in Japan are to understand the leak development behavior, to develop an analytical method based on the knowledge obtained through the experiments, and finally to apply it to actual LMFBR steam generators to evaluate the conservatism of design basis leak (DBL) of the plants. Therefore, various kinds of experimental study in relevant fields were conducted using SWAT test facilities and the LEAP code was developed. (author). 2 figs

  16. Steady-state heat transfer in an inverted U-tube steam generator

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1986-01-01

    Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during steady-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K hot-leg fluid temperatures, 6.2 MPa secondary pressure). The MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

  17. Residual stresses associated with the hydraulic expansion of steam generator tubing into tubesheets

    International Nuclear Information System (INIS)

    Middlebrooks, W.B.; Harrod, D.L.; Gold, R.E.

    1993-01-01

    Various methods are being used to expand heat transfer tubes into the thick tubesheets of nuclear steam generators. The residual stresses in the as-expanded tubes and methods for reducing these stresses are important because of the role which residual stresses play in stress corrosion cracking and stress assisted corrosion of the tubing. Of the various expansion processes, the hydraulic expansion process is most amenable to analytical study. This paper presents results on the residual stresses and strains in hydraulically expanded tubes and the tubesheet as computed by two different finite element codes with three different finite element models and by a theoretical incremental analysis method. The calculations include a sensitivity analysis to assess the effects of the expansion variables and the effect of stress relief heat treatments. (orig.)

  18. New process for the relief of mechanically induced stresses in steam generator tubes

    International Nuclear Information System (INIS)

    Joyeux, J.P.

    1980-01-01

    Heat exchangers include very generally a set of tubes assembled in 'U-type' exchangers or in 'pass-through' exchangers. The tubes are introduced in holes drilled in the tube sheet plate, welded at their extremity and expanded to insure the necessary tightness. The steam generators built by FRAMATOME belong to the U-type and include, depending upon the nominal power of the plant, about three or five thousand inconel tubes. This material has been selected for its resistance to corrosion action at high temperatures. But one drawback of inconel is that residual stress lowers considerably this resistance to corrosion; so it is very important to apply manufacturing processes involving a residual stress level as low as possible. A new process, which involves 'kiss' rolling, is described. (author)

  19. Surry Power Station secondary water chemistry improvement since steam generator replacement and the unit two feedwater pipe rupture

    International Nuclear Information System (INIS)

    Swindell, E.T.

    1988-01-01

    Surry Power Station has two Westinghouse-designed three-loop PWRs of 811 MWe design rating. The start of commercial operation was in July, 1972 in No.1 plant, and May, 1973 in No.2 plant. Both plants began the operation using controlled phosphate chemistry for the steam generators. In 1975, both plants were converted to all volatile treatment on the secondary side due to the tube wall thinning corrosion in the steam generators, which was associated with the phosphate sludge that was building up on the tube sheets and created acidic condition. Thereafter, condenser and air leakage and steam generator denting occurred, and after the operation of 8 years 2 month of No.1 plant and 5 years 9 months of No.2 plant, the steam generators were replaced. A major plant improvement program was designed and implemented from 1979 to 1980. The improvement in new steam generators, the modification for preventing corrosion, the addition of a steam generator blowdown recovery system, the reconstruction of condensers, the installation of full flow, deep bed condensate polishers, the addition of Dionex 8,000 on-line ion chromatograph, a long term maintenance agreement with Westinghouse and so on are reported. (Kako, I.)

  20. Study on reverse flow characteristics under natural circulation in inverted U-tube steam generator

    International Nuclear Information System (INIS)

    Duan Jun; Zhou Tao; Zhang Lei; Hong Dexun; Liu Ping

    2013-01-01

    Natural circulation is important for application in the nuclear power industry. Aiming at the steam generator of AP1000 pressurized water reactor loop, the mathematical model was established to analysis the reverse flow of single-phase water in the inverted U-tubes of a steam generator in a natural circulation system. The length distribution and the mass flow rates in both tubes with normal and reverse flow were determined respectively. The research results show that the reverse flow may result in sharp decrease of gravity pressure head, circulation mass flow rate and heat release rate of natural circulation. It has adverse influence on natural circulation. (authors)

  1. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    Langford, O.M.; Peelman, H.E.

    1978-01-01

    Means and method are described for energizing and regulating a neutron generator tube having a target, an ion source and a replenisher. It providing a negative high voltage to the target and monitoring the target current. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordence with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current

  2. Eddy-current tests on operational evaluation of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Lopez, Luiz Antonio Negro Martin; Ting, Daniel Kao Sun

    2000-01-01

    This paper presents a worldwide research on the technical and economical impacts due to failure in tube bundles of nuclear power plant steam generators. An Eddy current non destructive test using Foucault currents for the inspection and failure detection on the tubes, and also the main type of defects. The paper also presents the signals generated by a Zetec MIZ-40 test equipment. This paper also presents a brief description of an automatic system for data analysis which is under development by using a fuzzy logic and artificial intelligence

  3. Means and method for controlling the neutron output of a neutron generator tube

    International Nuclear Information System (INIS)

    1977-01-01

    A means and method for energizing and regulating a neutron generator tube is described. It has a target, an ion source and a replenisher. A negative high voltage is applied to the target and the target current monitored. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordance with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current. (C.F.)

  4. 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)

  5. Calculation of a steam generating tube stressed state under temperature oscillations in burnout zone

    International Nuclear Information System (INIS)

    Vorob'ev, V.A.; Loshchinin, V.M.; Remizov, O.V.

    1982-01-01

    The technique for evaluating the steam generating tube stressed state under the wall temperature oscillations in the burnout zone is described. The technique is based on analytical solutions for transfer functions connecting the amplitude of surface temperature oscillation with the amplitude and frequency of heat transfer coefficient oscillation and amplitude of thermoelastic stress oscillation with that of temperature oscillation. The results of calculations according to considered technique are compared with that of the problem numerical solution. The conclusion is made that the technique under consideration may be applied for evaluation of steam generator evaporating tube lifetime [ru

  6. Predicted wear on the tube outside surface due to foreign object in the secondary side of steam generator

    International Nuclear Information System (INIS)

    Kim, Hyung Nam; Cho, Nam Cheoul

    2012-01-01

    It is necessary to evaluate the effects of foreign objects on steam generator tubes and to use this information to take appropriate safety precautions to prevent nuclear accidents. Foreign objects may include loose parts from the feed water system and items lost by workers during o/h, and may flow into the secondary side of steam generators during operation. A foreign object could damage steam generator tube walls if there is relative motion between the tube and the foreign object. This is especially true for foreign objects that land on the tube sheet because the velocity of cross flow, which creates a contact force between the tube and foreign object, is relatively high there. During steam generator overhauls, foreign objects are detected by non destructive methods such as the visual test and/or the eddy current test. Confirmed foreign objects should be removed for nuclear safety. The Foreign Object Search and Retrieval System (FOSAR) can be used to remove foreign objects from the steam generators with a square tube array. However, the FOSAR cannot be used (or can be used in only a very restricted area, such as the outside of the tube bundle) in the steam generators with a triangular tube array. In order to continue nuclear power plant operations without removing foreign objects, the integrity of the steam generator tube must be verified. This paper introduces a practical method developed to evaluate the effects of foreign objects detected on tube sheets in the secondary sides of steam generators

  7. Experimental residual stress evaluation of hydraulic expansion transitions in Alloy 690 steam generator tubing

    International Nuclear Information System (INIS)

    McGregor, R.; Doherty, P.; Hornbach, D.; Abdelsalam, U.

    1995-01-01

    Nuclear Steam Generator (SG) service reliability and longevity have been seriously affected worldwide by corrosion at the tube-to-tubesheet joint expansion. Current SG designs for new facilities and replacement projects enhance corrosion resistance through the use of advanced tubing materials and improved joint design and fabrication techniques. Here, transition zones of hydraulic expansions have undergone detailed experimental evaluation to define residual stress and cold-work distribution on and below the secondary-side surface. Using X-ray diffraction techniques, with supporting finite element analysis, variations are compared in tubing metallurgical condition, tube/pitch geometry, expansion pressure, and tube-to-hole clearance. Initial measurements to characterize the unexpanded tube reveal compressive stresses associated with a thin work-hardened layer on the outer surface of the tube. The gradient of cold-work was measured as 3% to 0% within .001 inch of the surface. The levels and character of residual stresses following hydraulic expansion are primarily dependent on this work-hardened surface layer and initial stress state that is unique to each tube fabrication process. Tensile stresses following expansion are less than 25% of the local yield stress and are found on the transition in a narrow circumferential band at the immediate tube surface (< .0002 inch/0.005 mm depth). The measurements otherwise indicate a predominance of compressive stresses on and below the secondary-side surface of the transition zone. Excellent resistance to SWSCC initiation is offered by the low levels of tensile stress and cold-work. Propagation of any possible cracking would be deterred by the compressive stress field that surrounds this small volume of tensile material

  8. Evaluation of the eddy-current method for the inspection of steam generator tubing: denting

    International Nuclear Information System (INIS)

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

    1977-01-01

    Continued evaluation of existing eddy-current in-service inspection (ISI) methods for steam generator tubing has emphasized the effects of denting. Denting is a circumferential deformation of the tube in the support region as a result of the formation of a magnetite, Fe 3 O 4 , corrosion product on the carbon steel tube support plate in the crevice between the support plate and the tube wall. The effects of denting on eddy current inspection are twofold: (1) The detection and measurement of defects in the dented region is hampered by the distortion of the eddy-current response signals caused by the dents and (2) large dents may prevent the passage of high fill factor probes forcing the inspection teams to use a smaller probe in undented regions. Unless appropriate measures are taken, less reliable detection of defects and estimation of their depth can result. The report presents an experimental evaluation of existing eddy-current ISI methods when used to inspect dented regions. Tubes were mechanically dented to simulate various degrees of service induced denting. These tubes were then inserted in carbon steel tube support plates and the crevice region between the support plate and tube was subsequently packed with powdered magnetite. Diametral dent sizes considered during this program varied between 1 and 10 mils. Eddy-current signal patterns and strip chart recordings were obtained using a variety of test parameters, such as frequency, phase rotation and coil configuration, on tube samples containing various degrees of denting and with several types of defects in the dented system

  9. 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

  10. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R. [Argonne National Lab., IL (United States)

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  11. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

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

    International Nuclear Information System (INIS)

    Sviridenko, I.I.

    2005-01-01

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

  13. Theoretical-experimental assessment of the variables affecting fretting of Atucha I nuclear power plant utility steam generators tubes

    International Nuclear Information System (INIS)

    Kulichevsky, Raul M.

    1995-01-01

    Fretting wear of Steam Generator tubes caused by flow induced vibrations generates uncertainty on their integrity. The knowledge of the controlling variables of the wear process may give a criterion to evaluate the tubes residual life. Information on vibratory response and dynamic interaction between tubes and their supports are prerequisites for understanding the relationship between fretting wear and tube vibration. Experimental results of the vibratory response of an Atucha-I nuclear power plant type U-tube, the influence of tube/support clearance on this response and a study of tube/support dynamic interaction, which allow the verification of a finite element model of this type of tubes, are presented in this work. Also wear results for the Incoloy 800/DIN 1.4550 austenitic stainless steel pair of materials and a first evaluation of the wear constant of this pair are presented. (author)

  14. Effect of heat transfer tube leak on dynamic characteristic of steam generator

    International Nuclear Information System (INIS)

    Sun Baozhi; Shi Jianxin; Li Na; Zheng Lusong; Liu Shanghua; Lei Yu

    2015-01-01

    Taking the steam generator of Daya Bay Nuclear Power Station as the research object, one-dimensional dynamic model of the steam generator based on drift flux theory and leak model of heat transfer tube were established. Steady simulation of steam generator under different conditions was carried out. Based on verifying the drift flux model and leak model of heat transfer tube, the effect of leak location and flow rate under different conditions on steam generator's key parameters was studied. The results show that the drift flux model and leak model can reflect the law of key parameter change accurately such as vapor mass fraction and steam pressure under different leak cases. The variation of the parameters is most apparent when the leak is at the entrance of boiling section and vapor mass fraction varies from 0.261 to 0.163 when leakage accounts for 5% of coolant flow rate. The successful prediction of the effect of heat transfer tube leak on dynamic characteristics of the steam generator based on drift flux theory supplies some references for monitoring and taking precautionary measures to prevent heat transfer tube leak accident. (authors)

  15. Ultrasonic wall thickness gauging for ferritic steam generator tubing as an in-service inspection tool

    International Nuclear Information System (INIS)

    Haesen, W.M.J.; Tromp, Th.J.

    1980-01-01

    In-service inspection of LWR steam generators is more or less a standard routine operation. The situation can be very different for LMFBRs. For the SNR 300 (Kalkar Power Station) the situation is different because the steam generators have ferritic tubing. The tube walls are comparatively thick, 2 to 4.5 mm. During inservice examinations the steam generators will be drained on both sides, however on the sodium side a sodium film will be present. Furthermore the SNR 300 will have two types of steam generator. A straight tube design and a helical coil design will be used. Both types consist of a evaporator and superheater. The steam generators are of course not radioactive. It is obvious that in this case the eddy current (EC) technique is not an enviable inservice inspection tool. Basically EC is a surface flaw detection technique. Only the saturation magnetisation method will improve the EC technique sufficiently for ferritic material. However the 'in bore examination' with the saturation technique was, in case of the SNR 300 steam generator tubing, considered impossible since the inner diameters are fairly small. Furthermore sodium traces may influence the EC method. Although multifrequency methods can solve this problem, EC is not considered as a useful tool for examining ferritic tubing. Another method is to employ the 'stray flux' method which is under development with the TNO organization in Holland. The EC and stray flux method do have one drawback, these methods do not detect gradual changes in wall thickness. Ultrasonic examinations will be used in the SNR 300 as the main inspection tool for the steam generators. In this paper the reasons why ultrasonic examination was selected are explained. The results of the development work on this subject are discussed

  16. Application of laser-based profilometry to tubing in power generating utilities

    Science.gov (United States)

    Doyle, James L.

    1995-05-01

    Over the past several years lasers have been employed in an ever widening number of applications in an incredibly diverse set of markets. In the area of nondestructive testing, however, laser-based systems have only recently made inroads into the commercial markets. About ten years ago QUEST Integrated, Inc., began working with the U.S. Navy to adapt the principal of laser triangulation to solve a serious maintenance related problem. The internal surfaces of marine boiler tubes were experiencing pitting and corrosion which had resulted in catastrophic shipboard failures. At that time, conventional visual methods only allowed operators to inspect the first eighteen inches of the tube using a rigid borescope. If any pits were located, a mechanical stylus mechanism was used to obtain an approximate depth measurement of the pit. The condition of the balance of the tube was then extrapolated based on this extremely limited amount of information. Often the worst pitting was found in the bends of the tube, which could not be inspected by the visual method. Finally, a catastrophic boiler failure on an aircraft carrier resulted in the initiation of a search by the U.S. Navy for a better solution. Quest was contracted to develop an articulated probe which could negotiate the full length of a boiler tube with multiple bends, and generate a complete digital map of the inside surface. A key requirement of this probe would be rapid and quantitative measurement of internal features such as ID pits and corrosion. In 1987 QUEST delivered the first laser- optic tube inspection system to the U.S. Navy for use in marine boiler tubes. The Laser Optic Tube Inspection System (LOTISTM) was immediately put to use and paid for itself many times over in reduced maintenance costs. Over the next six years several generations of LOTIS were developed for the U.S. Navy, each one providing more capabilities, improved inspection speeds, and more user friendly operator interface. Today, LOTIS is

  17. Damage mechanisms and estimation of the frequency of leaks of steam generator tubes in German PWRs

    International Nuclear Information System (INIS)

    Reck, H.

    1992-01-01

    Operating experience of steam generator tubes in German PWRs has shown that so far there have only relatively few cases of damage been registered. The only steam generators with a high failure rate were exchanged in 1983. The material of the affected tubes was Inconel 600. The types of failure that occurred in the late 70's and early 80's were mainly wastage corrosion, which was thought to be the result of phosphate operating. After optimising the water chemistry and changing to ''high AVT'' operating, the failure rate decreased considerably. In total, about 973 of the 193335 tubes that were in operation were plugged because of wall-thinning or leaks. There have been 6 leaks, with the highest leakage volume being 40 liters per hour. 7 refs., 6 figs., 6 tabs

  18. 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

  19. Hideout of sea water impurities in steam generator tube deposits: laboratory and field studies

    International Nuclear Information System (INIS)

    Balakrishnan, P.V.; Turner, C.W.; Thompson, R.; Sawochka, S.

    1996-01-01

    Sea water impurities hide out within thin (∼10 μm) deposits on steam generator tubes, as demonstrated by both laboratory studies using segments of fouled steam generator tubes pulled in 1992 from Crystal River-3 nuclear power station and field hideout return studies performed during recent plant shutdowns. Laboratory tests performed at 279 o C (534 o F) and heat fluxes ranging from 35 to 114 kW/m 2 (11,100 - 36,150 Btu/h.ft 2 ), conditions typical of the lower tubesheet to the first support plate region of a once-through steam generator, showed that impurity hideout can occur in thin free-span tube deposits. The extent of hideout increased with increasing heat flux. Soluble species, such as sodium and chloride ions, returned promptly to the bulk water from the deposits when the heat flux was turned off, whereas less soluble species, such as calcium sulfate and magnesium hydroxide, returned more slowly. Recent field hideout return studies performed at Crystal River-3 where the water level in the steam generators was maintained below the first tube support plate during the shutdown, thus wetting only the thin deposits in the free span and the small sludge pile, corroborate the laboratory findings, showing that hideout does indeed occur in the free-span regions of the tubes. These findings suggest that hideout within tube deposits has to be accounted for in the calculation of crevice chemistry from hideout return studies and in controlling the bulk chemistry using the molar ratio criterion. (author). 3 refs., 4 tabs., 3 figs

  20. Investigation of reliability of EC method for inspection of VVER steam generator tubes

    International Nuclear Information System (INIS)

    Corak, Z.

    2004-01-01

    Complete and accurate non-destructive examinations (NDE) data provides the basis for performing mitigating actions and corrective repairs. It is important that detection and characterization of flaws are done properly at an early stage. EPRI Document PWR Steam Generator Examination Guidelines recommends an approach that is intended to provide the following: Ensure accurate assessment of steam generator tube integrity; Extend the reliable, cost effective, operating life of the steam generators, and Maximize the availability of the unit. Steam Generator Eddy Current Data Analysis Performance Demonstration represents the culmination of the intense two-year industry effort in the development of a performance demonstration program for eddy current testing (ECT) of steam generator tubing. It is referred to as the Industry Database (IDB) and provides a capability for individual organizations to implement SG ECT performance demonstration programs in accordance with the requirements specified in Appendices G and H of the ISI Guidelines. The Appendix G of EPRI Document PWR Steam Generator Examination Guidelines specifies personnel training and qualification requirements for NDE personnel who analyze NDE data for PWR steam generator tubing. Its purpose is to insure a continuing uniform knowledge base and skill level for data analysis. The European methodology document is intended to provide a general framework for development of qualifications for the inspection of specific components to ensure they are developed in a consistent way throughout Europe while still allowing qualification to be tailored in detail to meet different nation requirements. In the European methodology document one will not find a detailed description of how the inspection of a specific component should be qualified. A recommended practice is a document produced by ENIQ to support the production of detailed qualification procedures by individual countries. VVER SG tubes are inspected by EC method but a

  1. Evaluation of machine learning tools for inspection of steam generator tube structures using pulsed eddy current

    Science.gov (United States)

    Buck, J. A.; Underhill, P. R.; Morelli, J.; Krause, T. W.

    2017-02-01

    Degradation of nuclear steam generator (SG) tubes and support structures can result in a loss of reactor efficiency. Regular in-service inspection, by conventional eddy current testing (ECT), permits detection of cracks, measurement of wall loss, and identification of other SG tube degradation modes. However, ECT is challenged by overlapping degradation modes such as might occur for SG tube fretting accompanied by tube off-set within a corroding ferromagnetic support structure. Pulsed eddy current (PEC) is an emerging technology examined here for inspection of Alloy-800 SG tubes and associated carbon steel drilled support structures. Support structure hole size was varied to simulate uniform corrosion, while SG tube was off-set relative to hole axis. PEC measurements were performed using a single driver with an 8 pick-up coil configuration in the presence of flat-bottom rectangular frets as an overlapping degradation mode. A modified principal component analysis (MPCA) was performed on the time-voltage data in order to reduce data dimensionality. The MPCA scores were then used to train a support vector machine (SVM) that simultaneously targeted four independent parameters associated with; support structure hole size, tube off-centering in two dimensions and fret depth. The support vector machine was trained, tested, and validated on experimental data. Results were compared with a previously developed artificial neural network (ANN) trained on the same data. Estimates of tube position showed comparable results between the two machine learning tools. However, the ANN produced better estimates of hole inner diameter and fret depth. The better results from ANN analysis was attributed to challenges associated with the SVM when non-constant variance is present in the data.

  2. Device for removing and recuperating sludge deposited on the tube plate of a steam generator

    International Nuclear Information System (INIS)

    Bes, Louis.

    1982-01-01

    The cleaning device includes a descaling ramp with high pressure jets permanently fixed inside the steam generator, a system for driving the sludge formed towards the centre of the tube plate and a valve for removing the sludge giving into a hollow central column [fr

  3. EXTRACSION: a system for automatic Eddy Current diagnosis of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Georgel, B.; Zorgati, R.

    1992-01-01

    Improving speed and quality of Eddy Current non-destructive testing of steam generator tubes leads to automation of all process that contribute to diagnosis. This paper describes how signal processing, pattern recognition and artificial and artificial intelligence are used to build a software package that is able to automatically provide an efficient diagnosis. (author)

  4. Principal-Generated YouTube Video as a Method of Improving Parental Involvement

    Science.gov (United States)

    Richards, Joey

    2013-01-01

    The purpose of this study was to evaluate the involvement level of parents and reveal whether principal-generated YouTube videos for regular communication would enhance levels of parental involvement at one North Texas Christian Middle School (pseudonym). The following questions guided this study: 1. What is the beginning level of parental…

  5. Experimental facility design for study of fretting in steam generator tubes

    International Nuclear Information System (INIS)

    Balbiani, J.P.; Bergant, M.; Yawny, A.

    2012-01-01

    The design of an experimental facility for fretting wear testing of steam generator tubes under pressurized water up to 340 o C, is presented. The main component of the device consists in an autoclave which permits to recreate steam generator operating conditions. CAD CATIA V5R18, CAE ABAQUS and ASME Sec. VII Div. 1 (Rules for Construction of Pressure Vessels) were used along the design process. The design of the autoclave included the pressure vessel itself and the necessary flanges and nozzles. In addition, an axial dynamic sealing system was designed to allow for actuation from outside the pressure boundary. Complementary, typical tube - support contact conditions were analyzed and the principal variables affecting their mutual interaction determined. In addition, a simple device which allows performing fretting wear testing on steam generator tubes in air at room temperature was fabricated and the feasibility of a quantitative assessment of different aspects related with the fretting induced damage was explored. Characterization techniques available at Centro Atomico Bariloche, like light microscopy, scanning electron microscopy (SEM), energy dispersive analysis of X-ray (EDAX) and surface damage analysis by optic profilometry were shown to be appropriate for this aim. The designed facility will allow evaluating fretting damage of tubes - support combinations that might be used on the steam generator of the prototype reactor CAREM-25. It is also expected it could be applied to characterize fretting severity in other applications (nuclear fuel elements) (author)

  6. Extraction: a system for automatic eddy current diagnosis of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Georgel, B.; Zorgati, R.

    1994-01-01

    Improving speed and quality of Eddy Current non-destructive testing of steam generator tubes leads to automatize all processes that contribute to diagnosis. This paper describes how we use signal processing, pattern recognition and artificial intelligence to build a software package that is able to automatically provide an efficient diagnosis. (authors). 2 figs., 5 refs

  7. 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 ).

  8. Design Concept of Array ECT Sensor for Steam Generator Tubing Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chan Hee; Lee, Tae Hun; Yoo, Hyun Ju [Korea Hydro and Nuclear Power Co. Ltd. CRI, Daejeon (Korea, Republic of)

    2015-05-15

    The eddy current testing, which is one of the nondestructive examination methods, is widely used for the inspection of heat exchangers including steam generator tubing in the nuclear power plant. It uses electromagnetic induction to detect flaws in conductive materials. Two types of eddy current probes are conventionally used for the inspection of steam generator tubing according to the main purpose. One is the bobbin probe technology and the other is the rotating probe. During the inspection, they have restrictions for the flaw detection or the inspection speed. An array probe can be alternative to the bobbin and rotating probes. The design concept of array coils with high sensitivity is described in this paper. It is expected that the eddy current testing using this type of array sensors may provide high detectability and resolution for flaws in steam generator tubing. Eddy current technology has some barriers for the inspection of steam generator tubing in the nuclear power plant. Bobbin probes offer poor circumferential crack detection and rotating probes are time and money consuming due to the mechanical rotation. Array probe inspection technique can replace bobbin and rotating probe techniques due to its sensitivity for flaw detection and inspection speed. In general, circular-shaped coils are considered in an array eddy current probe.

  9. Steam-generator tube failures: world experience in water-cooled nuclear power reactors in 1974

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-01-01

    Steam-generator tube failures were reported at 25 of 59 water-cooled nuclear power reactors surveyed in 1974, compared to 11 of 49 in 1973. A summary is presented of these failures, most of which, where the cause is known, were the result of corrosion. Water chemistry control, inspection and repair procedures, and failure rates are discussed

  10. The development and application of overheating failure model of FBR steam generator tubes. 2

    International Nuclear Information System (INIS)

    Miyake, Osamu; Hamada, Hirotsugu; Tanabe, Hiromi

    2001-11-01

    The JNC technical report 'The Development and Application of Overheating Failure Model of FBR Steam Generator Tubes' summarized the assessment method and its application for the overheating tube failure in an event of sodium-water reaction accident of fast breeder reactor's steam generators (SGs). This report describes the following items studied after the publication of the above technical report. 1. On the basis of the SWAT-3 experimental data, realistic local heating conditions (reaction zone temperature and related heat transfer conditions) for the sodium-water reaction were proposed. New correlations are cosine-shaped temperature profiles with 1,170 C maximum for the 100% and 40% Monju operating conditions, and those with 1,110 C maximum for the 10% condition. 2. For the cooling effects inside of target tubes, LWR's studies of critical heat flux (CHF) and post-CHF heat transfer correlations have been examined and considered in the assessment. The revised assessment adopts the Katto's correlation for CHF, and the Condie-Bengston IV correlation for post-CHF. 3. Other additional examination for the assessment includes treatments of the whole heating effect (other than the local reaction zone) due to the sodium-water reaction, and the temperature-dependent thermal properties of the heat transfer tube material (2.25Cr-1Mo steel). The revised overheating tube failure assessment method has been applied to the Monju SG studies. It is revealed consequently that no tube failure occurs in 100%, 40%, and 10% operating conditions when an initial leak is detected by the cover gas pressure detection system. The assessment for the SG system improved for the detection and blowdown systems shows even better safety margins against the overheating tube failure. (author)

  11. Proceedings of steam generator sludge deposition in recirculating and once through steam generator upper tube bundle and support plates

    International Nuclear Information System (INIS)

    Baker, R.L.; Harvego, E.A.

    1992-01-01

    The development of remedial measures of shot peening have given nuclear utilities viable measures to address primary water stress corrosion cracking to extend steam generator life. The nuclear utility industry is now faced with potential replacement of steam generators in nuclear power plants due to stress corrosion cracking and intergranular attach in crevice locations on the secondary side of steam generators at tube support plates and at the crevice at the top of the tube sheet. Significant work has been done on developing and understanding of the effects of sludge buildup on the corrosion process at these locations. This session was envisioned to provide a forum for the development of an understanding of the mechanisms which control the transport and deposition of sludge on the secondary side of steam generators. It is hoped that this information will aid utilities in monitoring the progression of fouling of these crevices by further knowledge in where to look for the onset of support plate crevice fouling. An understanding of the progression of fouling from upper tube support plates to those lower in the steam generator where higher temperatures cause the corrosion process to initiate first can aid the nuclear utility industry in developing remedial measures for this condition and in providing a forewarning of when to apply such remedial measures

  12. Ruptured eardrum

    Science.gov (United States)

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

  13. Generation and growth rates of nonlinear distortions in a traveling wave tube

    International Nuclear Information System (INIS)

    Woehlbier, John G.; Dobson, Ian; Booske, John H.

    2002-01-01

    The structure of a steady state multifrequency model of a traveling wave tube amplifier is exploited to describe the generation of intermodulation frequencies and calculate their growth rates. The model describes the evolution of Fourier coefficients of circuit and electron beam quantities and has the form of differential equations with quadratic nonlinearities. Intermodulation frequencies are sequentially generated by the quadratic nonlinearities in a series solution of the differential equations. A formula for maximum intermodulation growth rates is derived and compared to simulation results

  14. Current state and development trends of inspections of steam generator tubes by eddy currents method

    International Nuclear Information System (INIS)

    Kubis, S.; Herman, M.

    1988-01-01

    The requirements are presented for steam generator tube testing and the factors are discussed affecting the accuracy of measurement by the eddy currents method. The individual types of the eddy currents method are described, such as the single-frequency method, the multi-frequency method and the pulse eddy currents method. The demands are summed up on manipulator systems whose design is adapted to the steam generator configuration conditions. (E.S.). 5 figs., 16 possu

  15. Microstructural Evolution and Creep-Rupture Behavior of Fusion Welds Involving Alloys for Advanced Ultrasupercritical Power Generation

    Science.gov (United States)

    Bechetti, Daniel H., Jr.

    Projections for large increases in the global demand for electric power produced by the burning of fossil fuels, in combination with growing environmental concerns surrounding these fuel sources, have sparked initiatives in the United States, Europe, and Asia aimed at developing a new generation of coal fired power plant, termed Advanced Ultrasupercritical (A-USC). These plants are slated to operate at higher steam temperatures and pressures than current generation plants, and in so doing will offer increased process cycle efficiency and reduced greenhouse gas emissions. Several gamma' precipitation strengthened Ni-based superalloys have been identified as candidates for the hottest sections of these plants, but the microstructural instability and poor creep behavior (compared to wrought products) of fusion welds involving these alloys present significant hurdles to their implementation and a gap in knowledge that must be addressed. In this work, creep testing and in-depth microstructural characterization have been used to provide insight into the long-term performance of these alloys. First, an investigation of the weld metal microstructural evolution as it relates to creep strength reductions in A-USC alloys INCONELRTM 740, NIMONICRTM 263 (INCONEL and NIMONIC are registered trademarks of Special Metals Corporation), and HaynesRTM 282RTM (Haynes and 282 are registered trademarks of Haynes International) was performed. gamma'-precipitate free zones were identified in two of these three alloys, and their development was linked to the evolution of phases that precipitate at the expense of gamma'. Alloy 282 was shown to avoid precipitate free zone formation because the precipitates that form during long term aging in this alloy are poor in the gamma'-forming elements. Next, the microstructural evolution of INCONELRTM 740H (a compositional variant of alloy 740) during creep was investigated. Gleeble-based interrupted creep and creep-rupture testing was used to

  16. Support vector regression model based predictive control of water level of U-tube steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Kavaklioglu, Kadir, E-mail: kadir.kavaklioglu@pau.edu.tr

    2014-10-15

    Highlights: • Water level of U-tube steam generators was controlled in a model predictive fashion. • Models for steam generator water level were built using support vector regression. • Cost function minimization for future optimal controls was performed by using the steepest descent method. • The results indicated the feasibility of the proposed method. - Abstract: A predictive control algorithm using support vector regression based models was proposed for controlling the water level of U-tube steam generators of pressurized water reactors. Steam generator data were obtained using a transfer function model of U-tube steam generators. Support vector regression based models were built using a time series type model structure for five different operating powers. Feedwater flow controls were calculated by minimizing a cost function that includes the level error, the feedwater change and the mismatch between feedwater and steam flow rates. Proposed algorithm was applied for a scenario consisting of a level setpoint change and a steam flow disturbance. The results showed that steam generator level can be controlled at all powers effectively by the proposed method.

  17. A Study of Frequency Mixing Approaches for Eddy Current Testing of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Jung, Hee Jun; Song, Sung Jin; Kim, Chang Hwan; Kim, Dae Kwang

    2009-01-01

    The multifrequency eddy current testing(ECT) have been proposed various frequency mixing algorithms. In this study, we compare these approaches to frequency mixing of ECT signals from steam generator tubes; time-domain optimization, discrete cosine transform-domain optimization. Specifically, in this study, two different frequency mixing algorithms, a time-domain optimization method and a discrete cosine transform(DCT) optimization method, are investigated using the experimental signals captured from the ASME standard tube. The DCT domain optimization method is computationally fast but produces larger amount of residue.

  18. Influence of sodium deposits in steam generator tubes on remote field eddy current signals

    Energy Technology Data Exchange (ETDEWEB)

    Thirunavukkarasu, S. [EMSI Section, NDE Division, Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Rao, B.P.C. [EMSI Section, NDE Division, Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)], E-mail: bpcrao@igcar.gov.in; Vaidyanathan, S.; Jayakumar, T.; Raj, Baldev [EMSI Section, NDE Division, Metallurgy and Materials Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2008-04-15

    The presence of sodium deposits in defective regions of steam generator (SG) tubes of fast-breeder reactors is expected to influence the remote field eddy current (RFEC) signals. By exposing five SG tubes having uniform wall loss grooves to a sodium environment in a specially designed test vessel, changes in the shape of RFEC signals were observed and it was possible to approximate the volume of sodium deposited in defects. An invariant signal parameter was determined for quantitative characterization of defects despite the presence of sodium in the defects.

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  20. Signal Characteristics of Eddy Current Test for Intergranular Attack of Steam Generator Tubes

    International Nuclear Information System (INIS)

    Choi, Myung Sik; Lee, Deok Hyun; Han, Jung Ho; Hur, Do Haeng; Cho, Se Gon; Yim, Chang Jae

    2002-01-01

    Because intergranular attack (IGA), one of the localized corrosion forms occurring on steam generator tubes, can not be fabricated by an electric discharge machining method, there are few data for the eddy current test (ECT) characteristics of IGA. In this paper, the characteristics of eddy current signals are evaluated using nonexpanded tubes with IGA defects formed in 0.1 M sodium tetrathionate solution at 40 .deg. C. The detectability and sizing accuracy of IGA were discussed in terms of the coil type and frequency of the ECT probes

  1. Evaluation of sampling plans for in-service inspection of steam generator tubes

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Heasler, P.G.; Baird, D.B.

    1994-02-01

    This report summarizes the results of three previous studies to evaluate and compare the effectiveness of sampling plans for steam generator tube inspections. An analytical evaluation and Monte Carlo simulation techniques were the methods used to evaluate sampling plan performance. To test the performance of candidate sampling plans under a variety of conditions, ranges of inspection system reliability were considered along with different distributions of tube degradation. Results from the eddy current reliability studies performed with the retired-from-service Surry 2A steam generator were utilized to guide the selection of appropriate probability of detection and flaw sizing models for use in the analysis. Different distributions of tube degradation were selected to span the range of conditions that might exist in operating steam generators. The principal means of evaluating sampling performance was to determine the effectiveness of the sampling plan for detecting and plugging defective tubes. A summary of key results from the eddy current reliability studies is presented. The analytical and Monte Carlo simulation analyses are discussed along with a synopsis of key results and conclusions

  2. 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)

  3. 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)

  4. Life Estimation of PWR Steam Generator U-Tubes Subjected to Foreign Object-Induced Fretting Wear

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

    2005-01-01

    This paper presents an approach to the remaining life prediction of steam generator (SG) U-tubes, which are intact initially, subjected to fretting-wear degradation due to the interaction between a vibrating tube and a foreign object in operating nuclear power plants. The operating SG shell-side flow field conditions are obtained from a three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element models of U-tubes to get the natural frequency, corresponding mode shape, and participation factor. The wear rate of a U-tube caused by a foreign object is calculated using the Archard formula, and the remaining life of the tube is predicted. Also discussed in this study are the effects of the tube modal characteristics, external flow velocity, and tube internal pressure on the estimated results of the remaining life of the tube

  5. Metallurgical problems in the exchange tube of a fast reactor steam generator

    International Nuclear Information System (INIS)

    Coriou, M.; Champeix, L.; Weisz, M.

    1980-10-01

    The design of the 1200 MWe Super Phenix power station steam generators is based on the following principles: once through helical tube exchangers which can be completely drained on the sodium side; the single wall exchange tubes are accessible to Foucault current testing during shutdowns. The authors explain the reasons for selecting the 800 Alloy for the exchange tubes. This choice was borne out by the results of several years of studies in the following areas: 6000 test hours with a 45 MWe model; corrosion test under stress in a water-steam and sodium plus caustic soda environment; resistance to creep and fatigue (effects of ageing and annealing, of the chemical compound); industrial feasibility, fabrication, utilization, bending, coiling, welding, testing. Concurrently, the EMl2 qualification finalizing has been pursued for the same application [fr

  6. The PISC programme on defective steam generator tubes inspection summary report

    International Nuclear Information System (INIS)

    Birac, C.; Comby, R.; Maciga, G.; Zanella, G.; Perez Prat, J.; Estorff, U. von

    1995-01-01

    The PISC III Actions are intended to extend the results and methodologies of the previous PISC exercises, i.e. the validation of the capabilities of the various examination techniques when used on real defects in real components under realistic conditions of inspection. The objective of this action is relatively close to that of the heavy structures programmes: the experimental evaluation of the performance of test procedures used for steam generator tubes in nuclear power plants during in-service or pre-service inspections. The exercise is a capability exercise consisting of Round Robin Tests on individual tubes including calibration, training and blind test tubes. In this paper the main conclusions from the RRT conducted in the framework of Action 5 will be presented and discussed. (author). 7 refs, 4 figs, 2 tabs

  7. 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

  8. Predictive analyses of flow-induced vibration and fretting wear in steam generator tubes

    International Nuclear Information System (INIS)

    Axisa, F.

    1989-01-01

    Maintaining the service life of PWR steam generators under highly reliable conditions requires a complex design to prevent various damaging processes, including those related to flow induced vibration. Predictive analyses have to rely on numerical tools to compute the vibratory response of multi-supported tubes in association with experimental data and semi-empirical relationships for quantifying flow-induced excitation mechanisms and tube damaging processes. In the presence of loose supports tube dynamics becomes highly nonlinear in nature. To de