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Sample records for outer sg tube

  1. SG tube identification

    International Nuclear Information System (INIS)

    Hoogstraten, P. van

    1994-01-01

    A ''Tracker'' system is described which is designed to identify any tube in a reactor steam generator quickly and safely. Occupational radiation doses to maintenance workers are reduced by using a Tracker and emergency down times are shortened. The system employs a television camera and light source in a stainless steel box with a large window. Both the camera and spotlight can be panned and tilted to reach any point on the tubesheet and are remotely controlled. An operator at a safe working distance can identify any tube visible on a real time video by comparison with the tubesheet pattern stored earlier in the computer memory. The identified tube can then be spotlighted and dealt with quickly by a maintenance worker inside the channel head. (UK)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  3. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  4. Plugging criteria for WWER SG tubes

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  5. Evaluation of reliability of EC inspection of VVER SG tubes

    International Nuclear Information System (INIS)

    Stanic, D.

    2001-01-01

    Evaluation of eddy current data collected during inspection of VVER steam generators is very complex task because of numerous parameters which have affect on eddy current signals. That was the reason that recently ago INETEC has started related scientific project in order to evaluate the reliability of eddy current (EC) inspection of VVER steam generator (SG) tubing. In the scope of project the following objectives will be investigated: 1. Determination of POD (Probability of detection) of various types degradation cracks, where their basic parameters are variables (basic parameters are depth, length, width, orientation, number) on three different sets of tubes (clean ideal tubes, tubes with pilgering, tubes electroplated with copper) 2. Sizing quality (accuracy, repeatability) (same data sets as defined in 1.) 3. Effect of fill factor on POD and sizing quality. 4. Effect of tube bends on POD and sizing quality. 5. Effect of other tube geometry variations on POD and sizing quality (tube ovality, transition zone region, expanded (rolled) part of tube, dents, dings). Investigation will start with bobbin probe technique which is the most used technique for general purpose VVER tube examination. Since INETEC is the only world company which successfully developed and applied rotating probe technique for VVER SG tubes, scope of the project will be extended on rotating probe technique utilizing 'pancake' and 'point' coil. Method reliability will be investigated first on the huge set of EDM notches representing various defect morphologies and simulating different factors, and the second part will be investigated on sets of degradation defects obtained by artificial corrosion. In the scope of the project the measures for enhancing the method reliability have to be determined. This considers the proper definition of parameters of examination system, as well as establishment of the suitable analysis procedures. This article presents the temporary results of the first part of

  6. Burst pressure and leak rate from fretted SG tubes

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  7. Dedicated new descaling method to characterize corrosion and cation release of SG tubing materials

    International Nuclear Information System (INIS)

    Clauzel, Maryline; Guillodo, Michael; Foucault, Marc; Engler, Nathalie; Chahma, Farah

    2012-09-01

    PWR steam generators (SGs), due to the huge wetted surface, are the main source of corrosion product release in the primary coolant circuit. Corrosion products may be transported throughout the whole circuit, activated in the core, and redeposited all over circuit surfaces, resulting in an increase of activity buildup. Understanding the phenomena leading to corrosion product release from SG tubing materials is of primary importance to minimize the global dose integrated by workers and to optimize the reactor shutdown duration and environment releases. Lab scale testing devices are a way to investigate cation release and propose mitigation measures. The descaling technique is based on the specific dissolution of the oxides making possible, by gravimetry, to directly evaluate the total quantity of corroded metal and the quantity of released elements. This technique allows for a statistical study as several SG coupons are exposed in one single test and is usually well-adapted to tubing materials having high or medium cation release behaviors, but has been proven too less accurate for the most recent manufactured SG tubes having low cation release rates. An optimized descaling technique has been developed to allow for the study of low-releasing SG tubing materials. Several steps of the process have been reconsidered. The electropolishing of the coupon is now performed after a careful determination of the thickness of the perturbed layer on the tube outer and/or inner surface to completely remove it so as to limit as much as possible the release of electro-polished faces which are not matter of the study. The number of coupons exposed in the autoclave has been reduced to avoid any saturation of the water primary chemistry, and two kinds of control coupons have been prepared instead of one in the former descaling method to take into account the uncertainties due to the descaling process as well as the CP possible redeposition on the coupons during exposure. Another

  8. Demonstration for the Applicability of the EPRI ETSS on the SG Tube Wear Defects Formed at the Tube Support Structure

    International Nuclear Information System (INIS)

    Shin, Ki Seok; Cheon, Keun Young; Nam, Min Woo; Min, Kyong Mahn

    2013-01-01

    In this paper, the authorized EPRI ETSS 27906.2 applied to the detection of tapered wear volumetric indications and depth sizing within the free span area, loose part not present was reviewed and applied to the site SG tubes for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation. The experiment to demonstrate the applicability of EPRI ETSS was performed by the employment of the newly prepared STD tube and resulted in ensuring the effectiveness and equivalency of the EPRI ETSS as well. The authorized EPRI ETSS 27906.2 for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation was reviewed and applied to the site SG tubes. The testing results were reviewed with the influences of SG tube material and the support structure. The impact of the tube materials was insignificant and that of the tube support structure showed somewhat conservative results. The testing resulted in successful demonstration of applicability of the EPRI ETSS on the SG tube wear defects at the tube support. One of the major flaw mechanisms detected in the currently operating domestic OPR-1000 pressurized water reactors(PWR's) steam generator(SG) tubes is wear defect. In general, wear defect has been constantly detected in the upper tube bundle imposed to the flow induced vibration interaction between tube and its support structure, and the quantity of the affected tubes has also shown the tendency to increase as plant operation life is added. In order to take appropriate measures and maintain the structural integrity for the SG tubes, wear defect is currently categorized as active damage mechanism and the tubes containing 40% or greater wear depth of the nominal tube wall thickness shall be plugged per SGMP(SG Management Program) Recently, a fairly large amplitude of wear defects on the Batwing(BW), one of the upper tube support structures in the SG tubes

  9. Demonstration for the Applicability of the EPRI ETSS on the SG Tube Wear Defects Formed at the Tube Support Structure

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ki Seok; Cheon, Keun Young; Nam, Min Woo [Korea Hydro and Nuclear Power Co. Ltd, Daejeon (Korea, Republic of); Min, Kyong Mahn [Universal Monitoring and Inspection Inc., Daejeon (Korea, Republic of)

    2013-10-15

    In this paper, the authorized EPRI ETSS 27906.2 applied to the detection of tapered wear volumetric indications and depth sizing within the free span area, loose part not present was reviewed and applied to the site SG tubes for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation. The experiment to demonstrate the applicability of EPRI ETSS was performed by the employment of the newly prepared STD tube and resulted in ensuring the effectiveness and equivalency of the EPRI ETSS as well. The authorized EPRI ETSS 27906.2 for getting the actual value of the wear depth and providing structural integrity interpretation based on engineering evaluation was reviewed and applied to the site SG tubes. The testing results were reviewed with the influences of SG tube material and the support structure. The impact of the tube materials was insignificant and that of the tube support structure showed somewhat conservative results. The testing resulted in successful demonstration of applicability of the EPRI ETSS on the SG tube wear defects at the tube support. One of the major flaw mechanisms detected in the currently operating domestic OPR-1000 pressurized water reactors(PWR's) steam generator(SG) tubes is wear defect. In general, wear defect has been constantly detected in the upper tube bundle imposed to the flow induced vibration interaction between tube and its support structure, and the quantity of the affected tubes has also shown the tendency to increase as plant operation life is added. In order to take appropriate measures and maintain the structural integrity for the SG tubes, wear defect is currently categorized as active damage mechanism and the tubes containing 40% or greater wear depth of the nominal tube wall thickness shall be plugged per SGMP(SG Management Program) Recently, a fairly large amplitude of wear defects on the Batwing(BW), one of the upper tube support structures in the SG

  10. Nuclear safety inspection in treatment process for SG heat exchange tubes deficiency of unit 1, TNPS

    International Nuclear Information System (INIS)

    Zhang Chunming; Song Chenxiu; Zhao Pengyu; Hou Wei

    2006-01-01

    This paper describes treatment process for SG heat exchange tubes deficiency of Unit 1, TNPS, nuclear safety inspection of Northern Regional Office during treatment process for deficiency and further inspection after deficiency had been treated. (authors)

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

    International Nuclear Information System (INIS)

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

    2010-09-01

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

  12. Development of a light weighted mobile robot for SG tube inspection in NPP

    International Nuclear Information System (INIS)

    Seo, Yong Chil; Jeong, Kyung Min; Shin, Hochul; Gweng, Jung Ju; Lee, Sung Uk; Jeong, Seung Ho; Choi, Young Soo; Kim, Seung Ho; Shin, Chun Sup; Park, Ki Tae

    2012-01-01

    Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water, because any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulation. In service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal SG chambers limits free access of human workers, remote manipulators are required. In South Korea, Manipulators such as the Zet ec SM series and the Westinghouse ROSA series have bee used. Such manipulators are rigidly mounted to man ways or tube sheets of SG. Confusions of the inspected tubes may occur from deflection of the manipulators. To reduce the deflections of the manipulators for covering the large working areas of tube sheets, sufficient rigidity is required and that leads to an increase of the weight. Such weight increase results in some difficulties for handling and more radiation exposure of human workers. Recently light weighed mobile robots have been introduced by Westinghouse and Zet ec. The robots can move keeping in contact with the tube sheets using devices which are commonly called cam locks. They are easier to handle and provide no confusion for the position of the inspected tubes. But when the clamping forces are loosed accidentally, they can be fall down and light repair works can be performed. This paper provides the design results for a lightweight mobile robot which is being developed in cooperation of our institutes

  13. Development of a light weighted mobile robot for SG tube inspection in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Yong Chil; Jeong, Kyung Min; Shin, Hochul; Gweng, Jung Ju; Lee, Sung Uk; Jeong, Seung Ho; Choi, Young Soo; Kim, Seung Ho [KAERI, Daejeon (Korea, Republic of); Shin, Chun Sup; Park, Ki Tae [Korea Plant Service and Engineering, Busan (Korea, Republic of)

    2012-10-15

    Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water, because any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulation. In service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal SG chambers limits free access of human workers, remote manipulators are required. In South Korea, Manipulators such as the Zet ec SM series and the Westinghouse ROSA series have bee used. Such manipulators are rigidly mounted to man ways or tube sheets of SG. Confusions of the inspected tubes may occur from deflection of the manipulators. To reduce the deflections of the manipulators for covering the large working areas of tube sheets, sufficient rigidity is required and that leads to an increase of the weight. Such weight increase results in some difficulties for handling and more radiation exposure of human workers. Recently light weighed mobile robots have been introduced by Westinghouse and Zet ec. The robots can move keeping in contact with the tube sheets using devices which are commonly called cam locks. They are easier to handle and provide no confusion for the position of the inspected tubes. But when the clamping forces are loosed accidentally, they can be fall down and light repair works can be performed. This paper provides the design results for a lightweight mobile robot which is being developed in cooperation of our institutes.

  14. Optimum Frequency for Eddy Current Testing Method of SMART SG tubes

    International Nuclear Information System (INIS)

    Lee, Yoon Sang; Jung, Hyun Kyu; Choung, Yun Hang

    2009-01-01

    The SMART SG tubes will be made of Alloy 690. The outside diameter will be 17 mm and the thickness will be 2.5 mm. They will be assembled helically around, and their innermost diameter will be about 600 mm and the total length will be about 32 meters. For safety, SMART SG tubes are designed for use with thick tubes such as 2.5 mm thickness compared to about 1 mm thickness of normal Korean standard pressurized water reactor tubes. Due to using thick tubes such as 2.5 mm varieties, it was doubted that the Eddy Current Testing Method (ECT) would be a feasible method. Therefore we are trying to simulate the bobbin probe signal for SMART SG tubes and comparing it to PWR SG ECT probe signal using VIM software, checking for the applicability of ECT. Also we are trying to compare the ECT signal of 2.5 mm thick stainless tubes to check if they are possible substitute material

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

    International Nuclear Information System (INIS)

    Poern, K.

    1997-01-01

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

  16. Structural integrity evaluation of SG tube with surface wear-type defects

    International Nuclear Information System (INIS)

    Kim, Jong Min; Huh, Nam Su; Chang, Yoon Suk; Kim, Young Jin; Hwang, Seong Sik; Kim, Joung Soo

    2006-01-01

    During the last two decades, several guidelines have been developed and used for assessing the integrity of a defective Steam Generator (SG) tube that is generally caused by stress corrosion cracking or wall-thinning phenomenon. However, as some of SG tubes are also failed due to fretting and so on, alternative failure estimation schemes are required for relevant defects. In this paper, parametric three-dimensional Finite Element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of SG tubes with different defect configurations; elliptical wear, tapered and flat wear type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of SG tube. After investigating the effect of key parameters such as defect depth, defect length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wear region. Comparison of failure pressures predicted by the proposed estimation scheme with corresponding burst test data showed a good agreement

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  18. A Mobile Robotic System for the Inspection and Repair of SG Tubes in NPPs

    Directory of Open Access Journals (Sweden)

    Yong-Chil Seo

    2016-04-01

    Full Text Available The reliability and performance of a steam generator (SG is one of the serious concerns in the operation of pressurized water nuclear power plants. Because of high levels of radiation, robotic systems have been used to inspect and repair SG tubes. In this paper, we present a mobile robotic system that positions the inspection and repair tools while hanging down from the tube sheets where the tubes are fixed. All of the driving mechanisms of the mobile robot are actuated by electric motors to start its works, providing that the electric power is prepared without the additional need for an on-site air services. A special tube-holding mechanism with a high holding force has been developed to prevent falling from the tube sheets, even in the case of an electric power failure. We have also developed a quick installation guide device that guides the mobile robot to desired initial positions in the tube sheet exactly and quickly, which helps to reduce the radiation exposure of human workers during the installation work. This paper also provides on-site experimental results and lessons learned.

  19. Multi-frequencies ECT algorithms to remove sodium noise in ISI of ferromagnetic SG tubes of FBR

    International Nuclear Information System (INIS)

    Mihalache, Ovidiu

    2012-01-01

    The paper presents developments and application of multi-frequency eddy current to be used during In-Service Inspection (ISI) of ferromagnetic steam generator (SG) tubes of Fast Breeder Reactors (FBR). Signal enhancement by means of multi-frequency ECT techniques are validated through 3D simulations of both signals and noise due to sodium forms around SG tube or SP. The purpose of such algorithms is to remove from ECT signal the electromagnetic noise resulting from sodium accumulated outside of SG tubes after SG vessel draining. Finite element method (FEM) simulations are used to analyse different sodium build-up scenarios observed experimentally, and to determine optimal multi-frequency ECT algorithms to suppress the most efficiently sodium noise. Also a new 'window multi-frequency' algorithm is applied and validated using 3-dimensional FEM simulations of SP and sodium forms. (author)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  1. Development of a SG Tube Inspection/maintenance Robot

    International Nuclear Information System (INIS)

    Shin, Ho Cheol; Jung, Kyung Min; Choi, Chang Hwan; Kim, Seung Ho

    2005-01-01

    A radiation hardened robot system is developed which assists in an automatic non-destructive testing and the repair of nuclear steam generator tubes. And a control system is developed. For easy carriage and installation, the robot system consists of three separable parts: a manipulator, a water chamber entering and leaving device of the manipulator and a manipulator base pose adjusting device. The kinematic analysis using the grid method was performed to search for the optimal manipulator's link parameters, and the stress analysis of the robotic system was also carried out for a structural safety verification. The robotic control system consists of a main personal computer placed near the operator and a local robotic position controller placed near the steam generator. A software program to control and manage the robotic system has been developed on the NT based OS to increase the usability. The software program provides a robot installation function, a robot calibration function, a managing and arranging function for the eddy-current test, a real time 3- D graphic simulation function which offers a remote reality to operators and so on. The image information acquired from the camera attached to the end-effector is used to calibrate the end-effector pose error and the time-delayed control algorithm is applied to calculate the optimal PID gain of the position controller. Eddy-current probe guide devices, a brushing tool, a motorized plugging tool and a U-tube internal visual inspection system have been developed. A data acquisition system was built to acquire and process the eddy-current signals, and a software program for eddy-current signal acquisition and processing. The developed robotic system has been tested in the Ulchin NPP type steam generator mockup in a laboratory. The final function test was carried out at the Kori Npp type steam generator mockup in the Kori training center

  2. A Conceptual Design of Light-weighted Mobile Robot for the Integrity of SG Tubes in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Yong Chil; Jeong, Kyung Min; Shin, Ho Chul; Lee, Sung Uk; Cho, Jae Wan; Choi, Young Soo; Kim, Seung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shin, Chun Sup; Park, Ki Tae [Korea Plant Serviceand Engineering, Busan (Korea, Republic of)

    2010-10-15

    Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water. It is because that any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulatory. In-service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal of SG chambers limits free access of human worker, remote manipulators are required. In South Korea, Manipulators such as the Zetec SM series and the Westinghouse ROSA series have been used. Such manipulators are rigidly mounted to manways or tube sheets of SG. Confusions for the inspected tubes may occur from deflection of the manipulators. To reduce the deflections of the manipulators for covering the large working areas of tube sheets, sufficient rigidity is required and it leads to the increase of the weight. Such weight increase results in some difficulties for handling and more radiation exposure of human workers. Recently light-weighed mobile robots have been introduced by Westinghouse and Zetec. The robots can move keeping in contact with the tube sheets using devices which are commonly called cam-locks. They are easier to handle and provide no confusion for the position of the inspected tubes. But when the clamping forces are loosed accidently, they can be fall down and light repair works can be performed. This paper provides the design results for a light weighted mobile robot which is recently being developed in cooperation of our institutes

  3. A Conceptual Design of Light-weighted Mobile Robot for the Integrity of SG Tubes in NPP

    International Nuclear Information System (INIS)

    Seo, Yong Chil; Jeong, Kyung Min; Shin, Ho Chul; Lee, Sung Uk; Cho, Jae Wan; Choi, Young Soo; Kim, Seung Ho; Shin, Chun Sup; Park, Ki Tae

    2010-01-01

    Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water. It is because that any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulatory. In-service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal of SG chambers limits free access of human worker, remote manipulators are required. In South Korea, Manipulators such as the Zetec SM series and the Westinghouse ROSA series have been used. Such manipulators are rigidly mounted to manways or tube sheets of SG. Confusions for the inspected tubes may occur from deflection of the manipulators. To reduce the deflections of the manipulators for covering the large working areas of tube sheets, sufficient rigidity is required and it leads to the increase of the weight. Such weight increase results in some difficulties for handling and more radiation exposure of human workers. Recently light-weighed mobile robots have been introduced by Westinghouse and Zetec. The robots can move keeping in contact with the tube sheets using devices which are commonly called cam-locks. They are easier to handle and provide no confusion for the position of the inspected tubes. But when the clamping forces are loosed accidently, they can be fall down and light repair works can be performed. This paper provides the design results for a light weighted mobile robot which is recently being developed in cooperation of our institutes

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. Inspection qualification as a tool to risk based ET ISI of VVER type SG tubes

    International Nuclear Information System (INIS)

    Horacek, L.

    2002-01-01

    A Pilot study on Eddy current inspection qualification of VVER 440 steam generator tubes, discussed in this paper, followed the ENIQ methodology principles and covered briefly the assumed scope of ET qualification, relevant elaborated qualification documents, known ISI limitations and a review of input information on component and defects determined for Eddy current inspection qualification of VVER 440 steam generator tubes. The information includes the fabrication of the test blocks with SG tube segments provided by intended defect simulations of realistic SCC type and basic data on the realistic SCC type defects manufacturing technology. Lessons learned from the development of manufacturing technology of SSC type of defects, regional blind tests, elaboration of the preliminary technical justification for Eddy current automated inspections, potential optimisation of inspection procedures, laboratory and practical open trials are summarised in the paper. The results of the Pilot study also especially in relation to POD curve being determined seem to be useful for practical operational ISI programme and Risk informed ISI decisions and the establishment of plugging criteria of VVER 440 and VVER 1000 type steam generator tubes. (orig.)

  6. The straw tube technology for the LHCb outer tracking system

    OpenAIRE

    Bachmann, S; Bagaturia, I; Deppe, H; Eisele, F; Haas, T; Hajduk, L; Langenegger, U; Michalowski, J; Nawrot, A; Polok, G; Pellegrino, A; Schuijlenburg, H; Schwierz, R; Sluijk, T; Spelt, J

    2004-01-01

    For the outer tracking system of the LHCb spectrometer 53.760 straws of 2.5 m length will be used. They are arranged in detector modules of 5 m length and 0.34 m width. The envisaged spatial resolution over the entire active area is 200$mu$m resulting in stringent requirements on the accuracy for the module construction. In this paper we discuss the optimisation of the straws, design and construction of detector modules. The long term operation properties of straws in two different counting g...

  7. Feasibility Test with a STS304 tube of the Eddy Current Test using a Bobbin Probe for the SMART SG Tube Inspection

    International Nuclear Information System (INIS)

    Lee, Yoon Sang; Jung, Hyun Kyu; Choung, Yun Hang

    2010-01-01

    The SMART SG tubes will be made of Alloy 690. The outside diameter will be 17 mm and the thickness will be 2.5 mm. They will be assembled helically around, and their innermost diameter will be about 600 mm and the total length will be about 32 meters. For the sake of safety, SMART SG tubes are designed for use with thick tubes such as 2.5 mm thickness compared to about 1 mm thickness of normal Korean standard pressurized water reactor tubes. Due to using thick tubes such as the 2.5 mm varieties, it was doubted that the Eddy Current Testing Method (ECT) would be a feasible method. Therefore we are trying to check the feasibility of the ECT using the substitute material STS304 tube instead of Alloy 690 tubes with the bobbin type ECT probe. The previous paper reported the feasibility of the ECT using modeling, but this paper will report the preliminary experimental results and comparison with the previous results of the modeling for the STS304 tube

  8. Electrical properties of various types of straw tubes considered for the LHCb outer tracker

    CERN Document Server

    Gromov, V

    2001-01-01

    Because of the appreciable length (up to 3.6 m) of the straw tube modules of the Outer Tracker, transmission line effects will have impact on their operational properties. These effects were clearly observed in a 1.6 m long prototype. A few types of straw tubes have been examined from the point of view of electrical properties, with emphasis on the study of signal transmission and cross-talk.

  9. LHCb: Ageing Phenomena in the Straw Tube Tracker (Outer Tracker) of the LHCb experiment

    CERN Multimedia

    Bachmann, S

    2009-01-01

    The outer tracking system of the LHCb spectrometer is built in the straw tube technology. In tota it consists of 53760 straw of 2.5m length. Thorough investigations have been performed to study the detector performance under long-term irradiations. Problems occuring caused by ageing are discussed and solutions are presented.

  10. New configuration for efficient and durable copper coating on the outer surface of a tube

    Directory of Open Access Journals (Sweden)

    Irfan Ahmad

    2017-03-01

    Full Text Available A well-adhered copper coating on stainless steel power coupler parts is required in superconducting radio frequency (SRF accelerators. Radio frequency power coupler parts are complex, tubelike stainless steel structures, which require copper coating on their outer and inner surfaces. Conventional copper electroplating sometimes produces films with inadequate adhesion strength for SRF applications. Electroplating also requires a thin nickel strike layer under the copper coating, whose magnetic properties can be detrimental to SRF applications. Coaxial energetic deposition (CED and sputtering methods have demonstrated efficient conformal coating on the inner surfaces of tubes but coating the outer surface of a tube is challenging because these coating methods are line of sight. When the substrate is off axis and the plasma source is on axis, only a small section of the substrate’s outer surface is exposed to the source cathode. The conventional approach is to rotate the tube to achieve uniformity across the outer surface. This method results in poor film thickness uniformity and wastes most of the source plasma. Alameda Applied Sciences Corporation (AASC has developed a novel configuration called hollow external cathode CED (HEC-CED to overcome these issues. HEC-CED produces a film with uniform thickness and efficiently uses all eroded source material. The Cu film deposited on the outside of a stainless steel tube using the new HEC-CED configuration survived a high pressure water rinse adhesion test. HEC-CED can be used to coat the outside of any cylindrical structure.

  11. Mechanical Tests Plan after Neutron Irradiation for SMART SG Tube Materials in a Hot Cell

    International Nuclear Information System (INIS)

    Ahn, Sang Bok; Baik, Seung Jai; Kim, Do Sik; Yoo, Byung Ok; Jung, Yang Hong; Song, Woong Sub; Choo, Kee Nam; Park, Jin Seok; Lee, Yong Sun; Ryu, Woo Seog

    2010-01-01

    An advanced integral PWR, SMART (System- Integrated Modular Advanced ReacTor) is being developed in KAERI. It has compact size and a relatively small power rating compared to a conventional reactor. The main components such as the steam generators, main circulation pumps are located in the reactor vessel. Therefore they are damaged from neutron irradiations generated from nuclear fuel fissions during operation. The SMART SG tubes which are 17 mm in a diameter and 2.5 mm in a thickness will be made of Alloy 690. To ensure the operation safety the post irradiation examinations is necessary to evaluate the deterioration levels of various original properties. Specially the amount of mechanical properties change should be reflected and revised to design data. For that tensile, fracture, hardness test are planned and under preparations. In this paper the detailed plans are reviewed. Three kinds of materials having different heat treatment procedures are prepared to fabricate specimens. The capsules installed the specimens are going to be irradiated in HANARO. Finally the tests for them will be performed in IMEF, Irradiated Materials Examination Facility at KAERI

  12. 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)

  13. Sensitive technique for detecting outer defect on tube with remote field eddy current testing

    International Nuclear Information System (INIS)

    Kobayashi, Noriyasu; Nagai, Satoshi; Ochiai, Makoto; Jimbo, Noboru; Komai, Masafumi

    2008-01-01

    In the remote field eddy current testing, we proposed the method of enhancing the magnetic flux density in the vicinity of an exciter coil by controlling the magnetic flux direction for increasing the sensitivity of detecting outer defects on a tube and used the flux guide made of a magnetic material for the method. The optimum structural shape of the flux guide was designed by the magnetic field analysis. On the experiment with the application of the flux guide, the magnetic flux density increased by 59% and the artificial defect detection signal became clear. We confirmed the proposed method was effective in a high sensitivity. (author)

  14. Study of steam condensation in SG tubes with large amount of nitrogen to be accumulated

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S.A.; Sitnik, Y.K. [EDO Gidropress, Podolsk (Russian Federation)

    1997-12-31

    The effect of nitrogen during SG heat transfer under SBLOCA conditions have been studied. Depressurization of the primary side leads to release of nitrogen dissolved in the hydroaccumulator water. Nitrogen can accumulate in SGs and affect adversely heat transfer under reflux condenser conditions. The main objective of the study has been to show that nitrogen does not prevent heat transfer in SGs of the VVER-640 which is reactor plant of new generation. (orig.).

  15. Study of steam condensation in SG tubes with large amount of nitrogen to be accumulated

    Energy Technology Data Exchange (ETDEWEB)

    Logvinov, S A; Sitnik, Y K [EDO Gidropress, Podolsk (Russian Federation)

    1998-12-31

    The effect of nitrogen during SG heat transfer under SBLOCA conditions have been studied. Depressurization of the primary side leads to release of nitrogen dissolved in the hydroaccumulator water. Nitrogen can accumulate in SGs and affect adversely heat transfer under reflux condenser conditions. The main objective of the study has been to show that nitrogen does not prevent heat transfer in SGs of the VVER-640 which is reactor plant of new generation. (orig.).

  16. Simultaneous and long-lasting hydrophilization of inner and outer wall surfaces of polytetrafluoroethylene tubes by transferring atmospheric pressure plasmas

    International Nuclear Information System (INIS)

    Chen, Faze; Song, Jinlong; Huang, Shuai; Xu, Wenji; Sun, Jing; Liu, Xin; Xu, Sihao; Xia, Guangqing; Yang, Dezheng

    2016-01-01

    Plasma hydrophilization is a general method to increase the surface free energy of materials. However, only a few works about plasma modification focus on the hydrophilization of tube inner and outer walls. In this paper, we realize simultaneous and long-lasting plasma hydrophilization on the inner and outer walls of polytetrafluoroethylene (PTFE) tubes by atmospheric pressure plasmas (APPs). Specifically, an Ar atmospheric pressure plasma jet (APPJ) is used to modify the PTFE tube’s outer wall and meanwhile to induce transferred He APP inside the PTFE tube to modify its inner wall surface. The optical emission spectrum (OES) shows that the plasmas contain many chemically active species, which are known as enablers for various applications. Water contact angle (WCA) measurements, x-ray photoelectron spectroscopy (XPS) and atomic force microscopy (AFM) are used to characterize the plasma hydrophilization. Results demonstrate that the wettability of the tube walls are well improved due to the replacement of the surface fluorine by oxygen and the change of surface roughness. The obtained hydrophilicity decreases slowly during more than 180 d aging, indicating a long-lasting hydrophilization. The results presented here clearly demonstrate the great potential of transferring APPs for surface modification of the tube’s inner and outer walls simultaneously. (paper)

  17. Status of IGA in Japanese plants and results of S/G pulled tube examinations

    International Nuclear Information System (INIS)

    Takamatsu, H.

    1986-01-01

    Currently there are 14 operating PWRs in Japan. Five of the plants have been affected by the intergranular stress corrosion cracking (IGSCC). The experience of the 14 plants with regard to tube plugging is summarized. The five affected plants are: Mihama 2, Takahama 1, Takahama 2, Ohi 1 and Genkai 1. The first two plants experienced IGSCC in the tube sheet crevice region (55 and 124 tubes plugged) and a slight problem with IGSCC in tube support plate crevices (13 and 8 tubes plugged). The other three plants have experienced extensive IGSCC in their tube support plate crevices (1330, 773 and 667 tubes plugged, respectively). The Japanese consider that their IGSCC (IGA) problem is caused by a combination material and environmental factors. All three factors must be present for attack to take place. If the two environmental factors can be controlled, then the attack will stop. The preventive measures being employed by the Japanese are given. These include: prevention of free caustic (i.e., sludge lancing, crevice cleaning, boron injection and improvement in water treatment) and maintenance of deoxidation environment (i.e., hydrazine soaking before operation and enriching the concentration of hydrazine in the secondary water)

  18. Tube Inner Coating of Non-Conductive Films by Pulsed Reactive Coaxial Magnetron Plasma with Outer Anode

    Directory of Open Access Journals (Sweden)

    Musab Timan Idriss Gasab

    2018-03-01

    Full Text Available The double-ended coaxial magnetron pulsed plasma (DCMPP method with auxiliary outer anode was introduced in order to achieve the uniform coating of non-conductive thin films on the inner walls of insulator tubes. In this study, titanium (Ti was employed as a cathode (sputtering target, and a glass tube was used as a substrate. In an argon (Ar and oxygen (O2 gas mixture, magnetron plasma was generated. Oxygen gas was introduced to deposit a titanium oxide (TiO2 film. A comparison between films coated with and without an auxiliary outer anode was made. As a result, it was clearly shown that the DCMPP method using an auxiliary outer anode enhanced the uniformity of the deposited non-conductive film compared to the conventional DCMPP method. Moreover, the optimum conditions under which the thin TiO2 film was deposited on the inner wall of the glass tube were revealed. From the results, it was supposed that the auxiliary outer anode contributed to the uniformity of the distributions of deposited negative charge on the non-conductive film and consequently the electric field and the plasma density uniform.

  19. Wastage Characteristics of a Modified 9Cr-1Mo Steel Tube Material for a SFR SG

    International Nuclear Information System (INIS)

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

    2009-01-01

    The development of a sodium heated steam generator with a safety and reliability is an essential requirement from the viewpoint of the economical efficiency of a sodium cooled fast reactor. In most cases these steam generators which are in the process of development, or operating, are of a shell-in tube type, with a high pressure water/steam inside the tubes and low pressure sodium on the shell-side, with a single wall tube as a barrier between these fluids. 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, important phenomena, 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 with 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. The objective of this study is a basic investigating of the sodium-water reaction phenomena by small water/steam leaks. For this, wastage tests for modified 9Cr-1Mo steel are being prepared

  20. On the possibilities for efficient simulation of leak rates through SG tubes

    International Nuclear Information System (INIS)

    Sorsek, I.; Cizelj, L.

    1995-01-01

    In this paper, the problem of predicting excessive leak rates through the through wall cracks in tubes at the tube - tube support plate intersections is discussed in some detail. Basically, we are able to define the leak rate through an individual defect. On the steam generator level, which actually means the sum of all individual leak rates, a new approach is introduced. The main characteristic of the new approach is seeking for a probability of exceeding the allowable leak rate rather than estimating more or less conservative highest expected leak rate value. This however introduces extensive computational effort which practically prevents the use of direct Monte Carlo simulations. Some possibilities to reduce the computational effort are discussed and their preliminary results compared. Also, some exact solutions were found and compared with numerical solutions achieved with the first order reliability method. Directions for future work in this important topic are given. (author)

  1. Simulation of the condensation experiment for the SG primary of SMART with MIDAS/SMR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seong Won [Korea Radiation Technology Institute Co., Daejeon (Korea, Republic of)

    2012-05-15

    SMART is new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. MIDAS/SMR V1.0.3 code has been used for estimating the severe accident from SMART. SMART has the SG tube with helical shape that is different from that of PWR. MIDAS code has the condensation model for the both sides of surfaces from the various kinds of geometry. But it does not have the condensation model for the helical type of tube. In this study, the condensation experiment for the outer surface of the SG tube in SMART that was performed by Po-Hang university was simulated with MIDAS/SMR under the assumption of straight pipe. The simulation results showed well predictions of the amount of heat removal from the SG tube primary side and thermal hydraulic parameters

  2. Simulation of the condensation experiment for the SG primary of SMART with MIDAS/SMR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

    2012-01-01

    SMART is new concept of reactor that all the main components such as the steam generator, the coolant pumps and the pressurizer are located inside the reactor vessel. MIDAS/SMR V1.0.3 code has been used for estimating the severe accident from SMART. SMART has the SG tube with helical shape that is different from that of PWR. MIDAS code has the condensation model for the both sides of surfaces from the various kinds of geometry. But it does not have the condensation model for the helical type of tube. In this study, the condensation experiment for the outer surface of the SG tube in SMART that was performed by Po-Hang university was simulated with MIDAS/SMR under the assumption of straight pipe. The simulation results showed well predictions of the amount of heat removal from the SG tube primary side and thermal hydraulic parameters

  3. 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

  4. 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.

  5. Assessment of the Polyacrylic Acid for an Ammonia Water Treatment and for Alloy 800NG SG Tube Material in Pressurized Water Reactors

    International Nuclear Information System (INIS)

    Lamouroux, Christine; You, Dominique; Plancque, Gabriel; Roy, Marc; Laire, Charles; Schnongs, Philippe

    2012-09-01

    To prevent the Steam Generators (SG) fouling by corrosion products or the Tube Support Plate (TSP) blockage the on-line injection of a dispersant such the Polyacrylic Acid (PAA) could be a relevant water treatment. Long-term trials performed in PWRs have shown that the PAA, injected at the SG inlet, facilitate the evacuation of the iron oxides by the SG blowdown. Given the ammonia treatment of the secondary water of the Belgian PWRs, the R and D program carried out was devoted to: - Verify the innocuousness of the PAA and its degradation products versus Alloy 800NG SCC susceptibility in case of over concentrations and sludge presence, - Assess the potential impact of the PAA and its thermal degradation products on the specific NH 3 water treatment. The main results can be summarized as following: The corrosion tests performed with PAA in case of over concentrations and sludge couldn't point out any negative effect of the dispersant on the SCC susceptibility of tubing materials such as Alloy 800NG. No significant modification of the tube oxide layer has been observed. At the SG operating temperature, the PAA is decomposed and a large spectrum from high to lower molecular weights polymers than the initial PAA arises. The fragmentation of the polymer into low molecular weight polyacrylic acids is obtained within 20 minutes and the average molecular weight is reduced by 50% from the original one. The thermal degradation products, their quantity and their kinetic of appearance, have been determined. The generated acetate concentration during the on-line dispersant application should remain low compared to the current values observed in the SG water. From the numerical simulation based on acetate concentration and on the kinetic law deduced from the experimental work, it can be concluded that in a 2-phase medium, the margin on the water pH compared to the neutral pH remains high. At 180 deg. C, no impact on the water pH is identified, taking into account realistic

  6. An Analysis of Design Characteristics of ECT Bobbin Probe for S/G Tube

    International Nuclear Information System (INIS)

    Nam, Min-Woo; Cho, Chan-Hee; Jee, Dong-Hyun; Jung, Jee-Hong; Lee, Hee-Jong

    2006-01-01

    The bobbin probe technique is basically one of the important ECT methods for the steam generator tube integrity assessment that is practiced during each plant outage. The bobbin probe is one of the essential components which consist of the whole ECT examination system, and provides us a decisive data for the evaluation of tube integrity in compliance with acceptance criteria described in specific procedures. The selection of examination probe is especially important because the quality of acquired ECT data is determined by the probe design characteristics, geometry and operation frequencies, and has an important effect on examination results. In this study, the relationship between electric characteristic changes and differential coil gap variation has been investigated to optimize the ECT signal characteristics of the bobbin probe. With the results from this study, we have elucidated that the optimum coil gap is 1.2 - 1.6mm that give the best result for O.D. volumetric defects in ASME calibration standards

  7. Characteristics Testing of the ECT Bobbin Probe for S/G Tube Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Nam, Min Woo; Lee, Hee Jong; Cho, Chan Hee; Yoo, Hyun Joo [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    The bobbin probe technique is basically one of the important ECT methods for the steam generator tube integrity assessment that is practiced during each plant outage. The bobbin probe also is the essential component which consists of the whole ECT examination system, and provides a decisive data for the evaluation of tube integrity in compliance with acceptance criteria described in specific procedures. The selection of probe is especially important because the quality of acquired ECT data is determined by the probe design characteristics, such as geometry and operation frequency, and has an important effect on examination results. The Electric Power Research Institute (EPRI) has recently defined the procedures for the qualification of eddy current hardware and technique. These procedures provide two basic methods for qualification. Flawed tube removed from operation, or artificial flaw is required for the original qualification of technique combined with related flaw mechanism. In case where the original qualification has been completed, the concept of equivalency may be used to extend the original qualification to similar probe designs. The qualified acquisition technique may be modified to substitute or replace instruments or probes without re-qualification provided that the range of essential variables defined in the examination technique specification sheet are met. In this case, both the original and replaced instrument or probe shall be characterized utilizing EPRI Guideline supplement 'H1'. This study is the result of the comparative performance evaluation of bobbin coil eddy current probes manufactured by KEPRI and a foreign manufacturer. As a result of this study, although there were minor differences between the two probe types, it was evaluated that the two probes were almost identical in the significant performance characteristics described in the EPRI guideline

  8. Statistical analysis of the influence of lancing on the secondary corrosion affecting SG tube ends in the CP0 series of PWRs

    International Nuclear Information System (INIS)

    Souchois, T.

    1995-05-01

    The main method of tube sheet cleaning during unit outages is high pressure lancing. A new tool called CECIL has been in use for this purpose since 1991 on the CP0 series of PWRs. This paper presents a statistical analysis of inspection data providing a basis for determining whether the type of lancing tool used has a 'statistically' significant effect, on the one hand, on the progression of the secondary corrosion affecting the CP0 type PWR SG tube ends and, on the other hand, on the appearance of corrosion on sound tubes after one operating cycle. The study results showed the CECIL cleaning method to be more efficient in tat larger amounts of accumulated deposits were removed and the risks of early corrosion were 14 times lower, with slower degradation in the course of subsequent cycles. (author). 2 refs., 8 figs., 19 tabs

  9. Aging Studies for the Large Honeycomb Drift Tube System of the Outer Tracker of HERA-B

    CERN Document Server

    Albrecht, H; Beck, M; Belkov, A; Berkhan, K; Bohm, G; Bruinsma, M; Buran, T; Capeans, M; Chamanina, J; Chen, BX; Deckers, H; Dehmelt, K; Dong, X; Eckmann, R; Emelianov, D; Fourletov, S; Golutvin, I; Hohlmann, M; Hoepfner, Kerstin; Hulsbergen, W; Jia, Y; Jiang, C; Kapitza, H; Karabekyan, S; Ke, Z; Kiryushin, Y; Kolanoski, H; Korpar, S; Krizan, P; Krucker, D; Lanyov, A; Liu, Y Q; Lohse, T; Loke, R; Mankel, R; Medin, G; Michel, E; Moshkin, A; Ni, J; Nowak, S; Ouchrif, M; Padilla, C; Pose, D; Ressing, D; Saveliev, V; Schmidt, B; Schmidt-Parzefall, W; Schreiner, A; Schwanke, U; Schwarz, Andreas S; Siccama, I; Solunin, S; Somov, S; Souvorov, V; Spiridonov, A; Staric, M; Stegmann, C; Steinkamp, O; Tesch, N; Tsakov, I; Uwer, U; Vassiliev, S; Vukotic, I; Walter, M; Wang, J J; Wang, Y M; Wurth, R; Yang, J; Zheng, Z; Zhu, Z; Zimmerman, R

    2003-01-01

    The HERA-B Outer Tracker consists of drift tubes folded from polycarbonate foil and is operated with Ar/CF4/CO2 as drift gas. The detector has to stand radiation levels which are similar to LHC conditions. The first prototypes exposed to radiation in HERA-B suffered severe radiation damage due to the development of self-sustaining currents (Malter effect). In a subsequent extended R&D program major changes to the original concept for the drift tubes (surface conductivity, drift gas, production materials) have been developed and validated for use in harsh radiation environments. In the test program various aging effects (like Malter currents, gain loss due to anode aging and etching of the anode gold surface) have been observed and cures by tuning of operation parameters have been developed.

  10. 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)

  11. Film flow analysis for a vertical evaporating tube with inner evaporation and outer condensation

    International Nuclear Information System (INIS)

    Park, Il Seouk

    2008-01-01

    A numerical study for the flow, heat and mass transfer characteristics of the evaporating tube with the films flowing down on both the inside and outside tube walls has been carried out. The condensation occurs along the outside wall while the evaporation occurs at the free surface of the inside film. The transport equations for momentum and energy are parabolized by the boundary-layer approximation and solved by using the marching technique. The calculation domain of 2 film flow regions (evaporating and condensation films at the inside and outside tube wall respectively) and tube wall is solved simultaneously. The coupling technique for the problem with the 3 different regions and the 2 interfaces of them has been developed to calculated the temperature field. The velocity and temperature fields and the amount of the condensed and evaporated mass as well as the position where the evaporating film is completely dried out are successfully predicted for various inside pressures and inside film inlet flow rates

  12. Fluid velocity in outer channels of tube bundle of PGV-1 steam generator prototype

    International Nuclear Information System (INIS)

    Salgo, C.

    1979-01-01

    This paper deals with the determination of the fluid velocity and pressure in every channel outside the tube bundle of the prototype model of the steam generator PGV-1, designed jointly by CNEN and NIRA. The results are obtained by the numerical solution of a system of algebraic and differential equations deduced by a mathematical ''channel'' model. Such results agree with the experiments performed on a model of the prototype PGV-1 by Alsthom Technique des Fluides

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-02-15

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

  14. Increasing reliability of defect characterization on sg tubings using a combination of signal processing and expert system

    International Nuclear Information System (INIS)

    Benoist, B.; David, B.; Pigeon, M.

    1989-01-01

    An expert system is developed for automatic analysis of eddy current signals provided by the multifrequency control of steam generators tubing. This article describes on one hand the aim and the results of the elimination of pilgrim noise, on the other hand the expert system which uses signal analysis and signal processing in unison

  15. On Eddy current examination (ECE) of Incoloy 800 SG tube using OD encircling and ID bobbin coil

    International Nuclear Information System (INIS)

    Kapoor, K.; Sunder Krishna, K.; Bakshu, S.A.

    2015-01-01

    The purpose of this paper is to present and compare the results of ECE carried out on steam generator tubes from OD side and ID side. During the manufacturing of the tubes Eddy current testing is being carried out using OD encircling probe as per ASTM E 571. Here the purpose of the test is to capture the manufacturing defects. The parameters of the test are optimized to achieve best sensitivity to this requirement. These tubes are then installed in the steam generator and once again ECE is carried out during installation (pre-service inspection-PSI) and during in-service inspection (ISI) by using ID bobbin coil. These tests are carried out as per ASME section V article 8 appendix 1. Here the purpose of the test is to detect wall thinning, dent, pits etc due to operation and to locate these defects (OD side or ID side). Here the operating parameters are optimized for phase separation of defects from OD and ID. These parameters are quite different from those used during the manufacturing ECE. Interpretation of the signals detected in PSI/ISI in must be done with care to correlate with defect indications detected during manufacturing. In the present study, tubes with certain manufacturing defects, detected with OD encircling test were subjected to ID bobbin coil examination. Also certain tubes with signal picked up during test from ID were examined by using the OD encircling probe. This comparison of the results provides a clear picture about the sensitivity and deficiency of the either type of test. (author)

  16. Life management of SG for WWER plants

    International Nuclear Information System (INIS)

    Trunov, N. B.; Dragunov, Yu. G.; Banyuk, G. F.

    2004-01-01

    Nowadays, 252 steam generators (SG) of horizontal type are in operation at WWER plants constructed by the Russian designs. In connection with end of the specified service life of the reactor plant equal to 30 years the activities are performed on service life extension of the main equipment including the SG. At some Units, throughout the design service life of SG there were problems resulting in necessity of SG replacement. At the same time the SGs at some Units are in successful operation above the design service life. This report deals with the peculiarities of operation of the horizontal SGs and the problems to be highlighted as the most important for service life extension. The main component to determine possibility for SG service life extension is the SG tubing. As the operating experience shows it is water chemistry of the secondary circuit that is the main factor influencing operability of the SG tubing. Therefore, differences in water chemistry organization leads to significant differences in operability of the SG tubing at various Units and in some cases within one Unit. Owing to the fact that the cases of water chemistry disturbance and the process of tubes fouling with the corrosion products of the main condensate system are not excluded, the damages continue to occur. Tube integrity shall be inspected by eddy current method using the various instrument complexes. This method has certain disadvantages but allows to estimate the degree and direction of degradation processes. The results of eddy current test (ECT) can be used to determine the plugging criterion for defective tubes. The significant number of defective tubes at some Units makes a choice of the plugging criterion to be an important problem, on which solution the SG safety, reliability and service life depends. The report deals with directions of activities in service life management for the SG at WWER plants. Main activities are improvement of water chemistry and non-destructive tests.(author)

  17. Self-wastage Behavior of Modified 9Cr-1Mo Steel as Heat Transfer Tube Material for a SFR SG

    International Nuclear Information System (INIS)

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

    2008-01-01

    Sodium cooled fast reactors adopt sodium heated steam generators in a secondary sodium circuit to raise the steam to drive the turbine. In most cases these steam generators are of a shell-in tube type, with a high pressure water/steam inside the tubes and low pressure sodium on the shell-side, with a single wall tube as a barrier between these fluids. 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, there results an important phenomena, so-called 'self-wastage' which may cause damage to the inside of the leakage site itself. If a steam generator is operated for some time with this condition, it is possible that it will damage the leak hole itself, which may eventually become a much larger opening. There is a danger that the resultant leak rate caused by a self-wastage might create the state of a small leak, or even an intermediate leak which would then give rise to the problems of a multi-target wastage. It has been observed in this study and others that the diameter of the nozzle hole grows to become a larger size 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. The objective this study is a basic investigating of the sodium water reaction phenomena by small water/steam leaks

  18. Effects of Relative SG Tube Pitches on the Performance Characteristics of a Small Modular Reactor driven by Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Youngjin; Yi, Kunwoo; Lee, Byungjin [KEPCO Engineering and Construction, Daejeon (Korea, Republic of)

    2016-10-15

    In this research, the capacity and basic dimensions for SMRs driven by a natural circulation are preliminarily assumed to determine the SMR configuration for the conceptual design, and each of the pre-set values is explained below. Firstly, the PZR configuration is not considered because it is not included to the main flow of the primary coolant. One of the SMR requirements is that SMR shall carry on the road. Hence, the vehicle geometrical limits are 15 meters for the length, and 3.5 meters for the height, approximately. With these limits for the dimensions of the SMR, RV length is assumed about 13.8 meters and RV diameter about 2.5 meters. In IAEA definition for SMRs, the capacity of electric power is no more than 300 MWe. If the efficiency of SMR power plant is assumed to 33% compared to the commercial power plant, the core power is below 1,000 MWth. In this research, the core power is assumed to 200 MWth arbitrarily during normal operation. The primary coolant passes through the outside of tubes, and the heat is transfer to the secondary feedwater. The secondary feedwater passes through the inside of tubes, and the heat from the primary coolant is received to generate the superheated steam. The present work carries out numerical simulations to get an insight for the effects of the diameters of the reactor vessel and riser using the parameters such as the steam generator tube pitches. To sum up, the calculation results show a good agreement with the theoretical equation and the uniform diameter loop has a more uniform temperature distribution and larger mass flow rate.

  19. Estimation of the number of physical flaws from periodic ISI data of SG tubes using effective POD

    International Nuclear Information System (INIS)

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

    2008-01-01

    It is necessary to know the number of flaws and their size distribution in order to calculate the probability of failure or to estimate the amount of leakage through the tube wall of steam generators. But In-Service Inspection (ISI) flaw data is different from the physical flaw data. In case of a single inspection, it is easy to estimate the number of physical flaws using the POD curve. However, we may be faced with some difficulties in obtaining the number of physical flaws from the periodic in-service inspection data. In this study a method for estimating the number of physical flaws from periodic in-service inspection data is proposed. In order to calculate the number of physical flaws with periodic ISI data, both probabilities of detecting and missing flaws should be considered. And flaw initiation and growth history must be known also. The flaw initiation and growth history can be inferred from appropriate probabilistic flaw growth rate. Two inference methods are proposed and compared. One is Monte Carlo simulation method and the other is transition (stochastic) matrix method. The effective POD, the total possibility of detection considering both probabilities of detecting and missing flaws for each flaw size, can be calculated using above two inference methods. And two methods are compared and the usefulness and convenience are evaluated from several applications

  20. Phase diagrams and radial distribution of the electric field components of coaxial discharges with outer dielectric tube at different wave modes

    International Nuclear Information System (INIS)

    Neichev, Z; Benova, E; Gamero, A; Sola, A

    2007-01-01

    The purpose of this work is to investigate phase diagrams and electric field radial distribution of coaxial discharges, sustained by a traveling electromagnetic wave, assuming finite and infinite thickness of the discharge chamber in the model. The calculations are made for azimuthally symmetric and dipolar wave modes. The phase diagrams and the radial profiles of the electric field at various thicknesses of the outer dielectric tube of the chamber and different discharge conditions are obtained. For the purpose of low pressure coaxial plasma modelling, radial profiles of the electric field at different discharge conditions have been investigated experimentally and compared with the theoretical results

  1. A novel portable device to measure the temperature of both the inner and the outer tubes of a parabolic receiver in the field

    Science.gov (United States)

    Hermoso, J. L. Navarro; Espinosa-Rueda, Guillermo; Martinez, Noelia; Heras, Carlos; Osta, Marta

    2016-05-01

    The performance of parabolic trough (PT) receiver tubes (RT) has a direct impact on Solar Thermal Energy (STE) plant production. As a result, one major need of operation and maintenance (O&M) in STE plants is to monitor the state of the receiver tube as a key element in the solar field. However the lack of specific devices so far has limited the proper evaluation of operating receiver tubés thermal performance. As a consequence non-accurate approximations have been accepted until now using infrared thermal images of the glass outer tube. In order to fulfill this need, Abengoa has developed a unique portable device for evaluating the thermal performance and vacuum state of parabolic trough receiver tubes placed in the field. The novel device described in this paper, simultaneously provides the temperature of both the inner steel tube and the outer glass tube enabling a check on manufacturers specifications. The on-field evaluation of any receiver tube at any operating temperature has become possible thanks to this new measuring device. The features and usability of this new measurement system as a workable portable device in operating solar fields provide a very useful tool for all companies in the sector contributing to technology progress. The originality of the device, patent pending P201431969, is not limited to the CSP sector, also having scientific significance in the general measuring instruments field. This paper presents the work carried out to develop and validate the device, also detailing its functioning properties and including the excellent results obtained in the laboratory to determine its accuracy and standard deviation. This information was validated with data collected by O&M teams using this instrument in a commercial CSP plant. The relevance of the device has been evidenced by evaluating a wide sample of RT and the results are discussed in this paper. Finally, all the on field collected data is used to demonstrate the high impact that using

  2. Convective heat transport of high-pressure flows inside active, thick walled-tubes with isothermal outer surfaces: usage of Nusselt correlation equations for an inactive, thin walled-tube

    Energy Technology Data Exchange (ETDEWEB)

    Campo, Antonio [Idaho State Univ., Nuclear Engineering Dept., Pocatello, ID (United States); Sanchez, Alejo [Universidad de los Andes, Depto. de Ingenieria Mecanica, Merida (Venezuela)

    1998-03-01

    A semi-analytical analysis was conducted for the prediction of the mean bulk- and interface temperatures of gaseous and liquid fluids moving laminarly at high pressures inside thick-walled metallic tubes. The outer surfaces of the tubes are isothermal. The central goal of this article is to critically examine the thermal response of this kind of in-tube flows utilizing two versions of the 1-D lumped model: one is differential-numerical while the other is differential-algebraic. For the former, the local Nusselt number characterizing an inactive, isothermal tube was taken from correlation equations reported in the heat transfer literature. For the latter, a streamwise-mean Nusselt number associated with an active, isothermal tube was taken from standard correlation equations that appear in text-books on basic heat transfer. For the two different versions of the 1-D lumped model tested, the computed results consistently demonstrate that the differential-algebraic, provides accurate estimates of both the mean bulk- and the interface temperatures when compared with those temperature results computed with formal 2-D differential models. (author)

  3. Technology development on production of test specimens from irradiated capsule outer-tube and mechanical evaluation test of stainless steel with high dose carried out by the technology

    International Nuclear Information System (INIS)

    Hayashi, Koji; Shibata, Akira; Iwamatsu, Shigemi; Sozawa, Shizuo; Takada, Fumiki; Ohmi, Masao; Nakagawa, Tetsuya

    2008-03-01

    The irradiation capsule 74M-52J was irradiated during total 136 cycles at reactor core of JMTR and the maximum neutron dose reached on 3.9x10 26 n/m 2 at the capsule outer-tube made of a type 304 stainless steel. In order to produce mechanical test specimens from the outer-tube, a punching technique was developed as a simple remote-handling method in a hot-cell. From comparison between the punching and the mechanical cutting methods, it was clarified that the punching technique was applicable to practical use. Moreover, an evaluation test of mechanical properties using specimens sampled from the 74M-52 was performed with in-water high temperature condition, less than 288degC. The result shows that the residual elongation is 18% at 150degC and 13% at 288degC. It was confirmed that the type 304 stainless steel irradiated up to such high dose shows enough ductility. (author)

  4. Combined natural convection and surface radiation in the annular region between a volumetrically heated inner tube and a finite conducting outer tube

    International Nuclear Information System (INIS)

    Gianoulakis, S.; Klein, D.E.

    1993-01-01

    Buoyancy-driven natural-convection heat transfer in enclosures has been the subject of considerable research with applications to electronic packaging, solar collectors, and shipping containers for spent nuclear fuel. A numerical study has been carried out to predict combined natural-convection and radiation heat transfer in the annular region between concentric tubes. The inner tube was volumetrically heated. Both tubes were of finite conductance. The surfaces of the annular region were diffuse and gray. The gas in the annulus was assumed to be nonparticipating. A newly developed hybrid finite element finite difference method was used for the study. This method combines finite element discretization of geometries with finite difference discretized solution procedures for the governing differential equations. This study examined the effects of surface radiative properties and material conductivities on the temperature and velocity fields and on local heat transfer rates. Fluid Raleigh numbers ranging from 10 3 to 10 7 , ratios of solid to fluid region thermal conductivities ranging from 10 to 10 4 , and surface total hemispherical emissivities ranging from 0.0 to 1.0 were examined in this study. It was found that the heat transfer across the annulus was dominated by conduction and radiation for the lower Raleigh number flows. As the fluid Raleigh number increased, convection became a primary mode of heat transfer. As the surface emissivity was increased in the annulus, the average Nusselt number on the inner tube surface decreased

  5. 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

  6. A long-term aging study of honeycomb drift tubes for the HERA-B Outer Tracker using a circulated and purified CF$_{4}$ gas mixture

    CERN Document Server

    Capéans-Garrido, M; Hohlmann, M; Schmidt, B

    2003-01-01

    The Outer Tracker of HERA-B uses a gas mixture containing CF/sub 4/ to obtain high electron drift velocities. The high cost of this gas makes it necessary to circulate the gas mixture which must then be purified to avoid accumulation of air and pollutants. However, the usage of gas purifiers poses the danger of outgassing pollutants from the purifiers themselves into the gas stream. Purifiers could also be attacked chemically by the aggressive products from the cracking of CF/sub 4/ molecules in the plasma avalanches of the detector. This could potentially release further harmful pollutants into the gas stream. To test for such effects, a long-term irradiation study of about 3000 h was carried out with the honeycomb drift tubes that are used in the Outer Tracker. This provided a check of the long-term stability of the gas purifiers before putting them into operation for the full-size detector. We report on the experimental setup, procedures and the results obtained. (8 refs).

  7. Evaluation on double-wall-tube residual stress distribution of sodium-heated steam generator by neutron diffraction and numerical analysis

    International Nuclear Information System (INIS)

    Kisohara, N.; Suzuki, H.; Akita, K.; Kasahara, N.

    2012-01-01

    A double-wall-tube is nominated for the steam generator heat transfer tube of future sodium fast reactors (SFRs) in Japan, to decrease the possibility of sodium/water reaction. The double-wall-tube consists of an inner tube and an outer tube, and they are mechanically contacted to keep the heat transfer of the interface between the inner and outer tubes by their residual stress. During long term SG operation, the contact stress at the interface gradually falls down due to stress relaxation. This phenomenon might increase the thermal resistance of the interface and degrade the tube heat transfer performance. The contact stress relaxation can be predicted by numerical analysis, and the analysis requires the data of the initial residual stress distributions in the tubes. However, unclear initial residual stress distributions prevent precious relaxation evaluation. In order to resolve this issue, a neutron diffraction method was employed to reveal the tri-axial (radius, hoop and longitudinal) initial residual stress distributions in the double-wall-tube. Strain gauges also were used to evaluate the contact stress. The measurement results were analyzed using a JAEA's structural computer code to determine the initial residual stress distributions. Based on the stress distributions, the structural computer code has predicted the transition of the relaxation and the decrease of the contact stress. The radial and longitudinal temperature distributions in the tubes were input to the structural analysis model. Since the radial thermal expansion difference between the inner (colder) and outer (hotter) tube reduces the contact stress and the tube inside steam pressure contributes to increasing it, the analytical model also took these effects into consideration. It has been conduced that the inner and outer tubes are contacted with sufficient stresses during the plant life time, and that effective heat transfer degradation dose not occur in the double-wall-tube SG. (authors)

  8. 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)

  9. 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)

  10. Spatial Variation of Hydrodynamic Mass Coefficients for Tube Bundle in a Cylindrical Shell

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Keum Hee; Ryu, Ki Wahn [Chonbuk National University, Jeonju (Korea, Republic of); Park, Chi Yong [KEPCO Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    Wear of the steam generator (SG) tubes affects the performance of nuclear power plants. Generally, the problem is caused by excessive flow-induced vibration (FIV). In analyzing the FIV, many researchers have used a uniform added mass coefficient for all of the SG tubes. However, the outermost SG tubes have more structural problems than inside tubes. The purpose of this study is to find out the added mass coefficients of each tube in a cylindrical shell

  11. 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

  12. Internal heat exchange tubes for industrial furnaces

    Energy Technology Data Exchange (ETDEWEB)

    Hoetzl, M.; Lingle, T.M.

    1992-05-26

    This patent describes a method for cooling the work within an industrial furnace. It comprises providing a longitudinally extending outer tube which extends into the furnace having a closed axial end and an open axial end; providing a preformed inner tube open at both ends within the outer tube; injecting a coolant into the inner tube so that the coolant flows from one axial end of the tube out the opposite end adjacent the closed end of the outer tube, and from the closed end of the outer tube to the open end thereof; circulating a gas within the furnace against the outer tube to effect heat transfer therewith.

  13. A study on LMFBR steam generator design without tube failure propagation in water leak events

    International Nuclear Information System (INIS)

    Futagami, Satoshi; Hayafune, Hiroki; Fujimura, Ken; Sato, Mitsuru

    2009-01-01

    The major target performance of the SG for commercialized FBR is not only economic performance but also property protection performance. The candidate SG design will be selected at the end of JFY 2010. The straight double wall tube SG is one of the SG candidates for commercialized FBR, and other SG concepts were studied in this paper. In proposing an alternative SG, alternative technological measures with a double wall tube were investigated and included reinforcing the tube against wastage and quick detection of initial tube leaks. Alternative SG concept candidates for preventing tube failure propagation and mitigation of water leak accidents were proposed through a combination of technological measures. The candidates were then comparatively evaluated from the point of view of property protection performance, total weight, technological issues, and so on. A coated wall tube SG and protective wall tube SG were decided on as the alternative SGs because of superior property protection performance and with the technological issues. At the end of JFY 2010, the straight double wall tube SG will be decided upon as the result of R and D activities, and alternative SGs evaluated in feasibility studies. A plan for studying feasibility with the technological issues of the alternative SG was proposed. (author)

  14. 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)

  15. Field experience with KWU SG chemical cleaning process

    International Nuclear Information System (INIS)

    Odar, S.

    1989-01-01

    The ingress of corrosion products into PWR steam generators (SG's) their deposition and the subsequent concentration of salt impurities can induce a variety of mechanisms for corrosion attack on SG tubing. Already, some plants have had to replace their steam generators due to severe corrosion damage and others are seriously considering the same costly action in the near future. One of the most effective ways to counteract corrosion mechanisms and thus to reduce the likelihood of SG replacement becoming necessary is to clean the SG's and to keep them clean. For many years, the industry has been involved in developing different types of cleaning techniques. Among these, chemical cleaning has been shown to be especially effective. In this article, the KWU chemical cleaning process, for which there is considerable application experience, is described. The results of field applications will be presented together with material compatibility data and information on cleaning effectiveness. (author)

  16. 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

  17. Safety significance of steam generator tube degradation mechanisms

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-01

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

  18. Outer magnetosphere

    International Nuclear Information System (INIS)

    Schardt, A.W.; Behannon, K.W.; Lepping, R.P.; Carbary, J.F.; Eviatar, A.; Siscoe, G.L.

    1984-01-01

    Similarities between the Saturnian and terrestrial outer magnetosphere are examined. Saturn, like earth, has a fully developed magnetic tail, 80 to 100 RS in diameter. One major difference between the two outer magnetospheres is the hydrogen and nitrogen torus produced by Titan. This plasma is, in general, convected in the corotation direction at nearly the rigid corotation speed. Energies of magnetospheric particles extend to above 500 keV. In contrast, interplanetary protons and ions above 2 MeV have free access to the outer magnetosphere to distances well below the Stormer cutoff. This access presumably occurs through the magnetotail. In addition to the H+, H2+, and H3+ ions primarily of local origin, energetic He, C, N, and O ions are found with solar composition. Their flux can be substantially enhanced over that of interplanetary ions at energies of 0.2 to 0.4 MeV/nuc

  19. 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)

  20. Pressure tube type research reactor

    International Nuclear Information System (INIS)

    Ueda, Hiroshi.

    1976-01-01

    Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

  1. Experimental studies on the evaporative heat transfer and pressure drop of CO{sub 2} and CO{sub 2}/propane mixtures flowing upward in smooth and micro-fin tubes with outer diameter of 5 mm for an inclination angle of 45

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jin Min; Kim, Min Soo [School of Mechanical and Aerospace Engineering, Seoul National University, Seoul 151-744 (Korea); Kim, Yong Jin [School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332 (United States)

    2010-08-15

    Heat transfer characteristics show different tendency according to the tube orientations such as horizontal, vertical, and inclined positions. In this study, evaporative heat transfer characteristics and pressure drop of CO{sub 2} and CO{sub 2}/propane mixtures flowing upward are investigated in inclined smooth and micro-fin tubes. Smooth and micro-fin tubes with outer diameter of 5 mm and length of 1.44 m with inclination angle of 45 were chosen as test tubes. Average inner diameters of test tubes are 4.0 mm (smooth tube) and 4.13 mm (micro-fin tube). The tests were conducted at mass fluxes from 212 to 656 kg/m{sup 2} s, saturation temperatures from -10 to 30 C and heat fluxes from 15 to 60 kW/m{sup 2} for CO{sub 2}. In addition, for CO{sub 2}/propane mixtures, the test was carried out at inlet temperatures from -10 to 30 C for several compositions (75/25, 50/50, 25/75 wt%) with the same mass fluxes, heat fluxes applied for CO{sub 2}. Heat transfer coefficients in inclined tube are approximately 1.8-3 times higher than those in horizontal tube and the average pressure drop of inclined tube exists between that of horizontal and vertical tubes. (author)

  2. 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)

  3. 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)

  4. Numerical simulation of condensation phase change flow in an inclined tube with bend

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Jong Chull; Shin, Byung Soo; Do, Kyu Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Lee, Yong Kap [Anflux Co., Seoul (Korea, Republic of)

    2012-10-15

    The new PWR design named APR+ incorporates a passive auxiliary feedwater system (PAFS) as shown in Fig.1. The PAFS consists of two separate divisions. Each division is equipped with one passive condensation heat exchanger (PCHX), isolation or drain or vent valves, check valves, instrumentation and control, and pipes. It is aligned to feed condensed water to its corresponding steam generator (SG). During the PAFS normal operation, steam being produced in the SG secondary side by the residual heat moves up due to buoyancy force and then flows into the PCHX where steam is condensed on the inner surface of the tubes of which the outer surfaces are cooled by the water stored in the passive condensation cooling tank (PCCT). The condensate is passively fed into the SG economizer by gravity. Because the thermal hydraulic characteristics in the PCHT determine the condensation mass rate and the possibility of system instability and water hammer, it is important to understand the condensation phase change flow in the PCHT. This paper presents a numerical simulation of the condensation phase change flow in the PCHX adopted for the APR+ PAFS.

  5. PLUSS-A weldless leaktight sleeve for alloy 600/690 steam generator tubes

    International Nuclear Information System (INIS)

    Potz, F.; Bohmann, W.

    1998-01-01

    The ABB PLUSS sleeving represents a new SG tube repair technique qualified and approved to replace in the future most of the plugging as well as welded sleeving. Basically the advantages of an innovative combination of both alloys 600/690 and 800 are taken into consideration. The upper sleeve/SG tube-joint is hydraulically expanded stressing the SG tube only within the elastic range. The lower joint is hard rolled. The installation processes are simple and reproducible, fast, computerized and individually recorded. The operating temperature range of the sleeved SG-tube is effectively reduced so that any further corrosion is impeded. Both, sleeve and SG tube are fully inspectable by ECT. (author)

  6. Mitigation of corrosion product ingress into SG's

    International Nuclear Information System (INIS)

    Han, S.H.

    1988-01-01

    Design and operation experiences to mitigate corrosion product ingress into SGs in Korea nuclear power plants are briefly reviewed. Maintaining the feedwater pH above 9.6 with morpholine seems to contribute significantly to reduction of iron transport to SGs. Measured iron transport rates were 4.8 g/hr/100 MWe at pH 9.8 and 2.8 g/hr/100 MWe at 9.3, respectively. Removal of corrosion products through SG blowdown is very limited. Its removal efficiency at the higher pH plant was in the neighborhood of 10 %. In one of the Korea Nuclear Units, a large amount of sludge piles were found in the middle of tube bundles especially on the cold leg side. Damaged tubes were identified by the multi-frequency eddy current tests and plugged later during the refueling period. Intermittent blowdown-rate increase was tried to enhance ionic impurity removal through SG blowdown. Even though it was not effective against Na, removal other impurity was improved, resulting in prolonged condensate polisher operation periods by 1 - 2 days. Two-bed polisher design, a cation bed followed by a mixed bed, was chosen for future PWR plants to enhance corrosion product filtering capability of the polishers. Condensate pump discharge polishing and divided hot well polishing methods are currently in consideration. (Nogami, K.)

  7. Pyrotechnic Tubing Connector

    Science.gov (United States)

    Graves, Thomas J.; Yang, Robert A.

    1988-01-01

    Tool forms mechanical seal at joint without levers or hydraulic apparatus. Proposed tool intended for use in outer space used on Earth by heavily garbed workers to join tubing in difficult environments. Called Pyrotool, used with Lokring (or equivalent) fittings. Piston slides in cylinder when pushed by gas from detonating pyrotechnic charge. Impulse of piston compresses fittings, sealing around butting ends of tubes.

  8. Measurement of gas-liquid two-phase flow around horizontal tube bundle using SF6-water. Simulating high-pressure high-temperature gas-liquid two-phase flow of PWR/SG secondary coolant side at normal pressure

    International Nuclear Information System (INIS)

    Ishikawa, Atsushi; Imai, Ryoj; Tanaka, Takahiro

    2014-01-01

    In order to improve prediction accuracy of analysis code used for design and development of industrial products, technology had been developed to create and evaluate constitutive equation incorporated in analysis code. The experimental facility for PWR/SG U tubes part was manufactured to measure local void fraction and gas-liquid interfacial velocity with forming gas-liquid upward two-phase flow simulating high-pressure high-temperature secondary coolant (water-steam) rising vertically around horizontal tube bundle. The experimental facility could reproduce flow field having gas-liquid density ratio equivalent to real system with no heating using SF6 (Sulfur Hexafluoride) gas at normal temperature and pressure less than 1 MPa, because gas-liquid density ratio, surface tension and gas-liquid viscosity ratio were important parameters to determine state of gas-liquid two-phase flow and gas-liquid density ratio was most influential. Void fraction was measured by two different methods of bi-optical probe and conductivity type probe. Test results of gas-liquid interfacial velocity vs. apparent velocity were in good agreement with existing empirical equation within 10% error, which could confirm integrity of experimental facility and appropriateness of measuring method so as to set up original constitutive equation in the future. (T. Tanaka)

  9. 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)

  10. 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

  11. The Hydrogen Detection Technique for SG Protection System

    International Nuclear Information System (INIS)

    Lv Mingyu; Pei Zhiyong; Yu Huajin

    2015-01-01

    SG that is pressure boundary between secondary loop and triple loop is the key equipment of fast reactor, in which heat in secondary loop is transferred to water or steam in triple loop. According to data from IAEA, SG is the highest failure rate equipment in fast reactor, especially because of failure of heat transfer tube. In order to monitor failure of heat transfer tube, Fast Reactor Engineering Department develops diffusion type hydrogen detection system, which is used to detect sodium-water reaction in time. This paper firstly introduces experimental research scheme and results of this hydrogen detection technique; Subsequently, it is described that how this technique can be engineering realized in CEFR; Moreover, through developing a series of calibration tests and hydrogen injection tests, it is obtained that sensitivity, response time and calibration curse for hydrogen detection system of CEFR. (author)

  12. 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)

  13. 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

  14. 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.

  15. Experimental and numerical study of deposit formation in secondary side SG TSP by electrokinetic approach

    International Nuclear Information System (INIS)

    Guillodo, Michael; Foucault, Marc; Ryckelynck, Natacha; Chahma, Farah; Guingo, Mathieu; Mansour, Carine; Alos-Ramos, Olga; Corredera, Geraldine

    2012-09-01

    Corrosion products deposit formation observed in PWR steam generators (SGs) - related to SG free span fouling and SG clogging - is now reported since several years. SG clogging is a localized phenomenon observed between the leading edge of the Tube Support Plate (TSP) and SG tubing materials. Based on visual inspections, it was found that the gaps between SG tubing material and TSP at the lower part of the broached holes were getting progressively blocked. Therefore, for safe operation, most affected PWRs had to be operated at reduced power. TSP blockage was mainly observed for low-pH water chemistry conditioning, which directly depends on the operating water chemistry. The TSP blockage mechanism is complex due to the localized conditions in which flow pattern change, chemistry and electrochemical conditions are not well understood. Electrokinetic considerations could be pointed out to explain the coupling of chemistry, materials and thermohydraulic (T/H) conditions. In this frame AREVA and EDF have launched a long-term R and D program in order to understand the mechanisms driving the formation of SG clogging. This study based on parametric laboratory tests aims to assess the role of secondary water chemistry, material and T/H conditions on deposit formation. The experimental approach focused on electrokinetic measurements of metallic substrates and on the assessment of oxidation properties of materials in secondary side chemistry. An overall analysis of recent results is presented to address SG deposit formation in secondary water chemistry for various conditioning amines - morpholine, ethanolamine and dimethylamine. To complete the study, the experimental results have been correlated to CFD simulations of particle deposition, by means of stochastic Lagrangian models. These calculations have in particular reproduced correctly the location of the most important particle deposit (the leading edge of the test tube), and have stressed the influence of the

  16. 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

  17. 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.

  18. Attachment of iron corrosion products on steam generator tube and feed-water pump in PWRs secondary system

    International Nuclear Information System (INIS)

    Shoda, Y.; Ishihara, N.; Miyata, H.; Ohira, T.; Watanabe, Y.; Nonaka, Y.

    2010-01-01

    Operating experience of the secondary systems in PWRs indicates that scale attachment distinctly have an effect on the performance of water-steam cycle. Attached scale on outer surface of steam generator (SG) tube could induce many problems such as decrease heat efficiency of plant, corrosion of tube by intergranular attack (IGA), and choke of flow channel. Scale attached on rotor blade of feed water pump increases the driving steam consumption to keep the constant flow rate, and results in the thermal efficiency decrease of the plant. In this study, two types of test about scale deposition on equipment were executed in the conditions simulating the secondary system of PWR. One is SG model test, which simulated the circulating boiler composed of single SG tube and blow down line. The deposition rate under AVT condition was equivalent to plants revealed with extended period. High-AVT test provided useful reference, because the deposition rate of power plant is too small to measure in a short period after the beginning of High-AVT operation in Japan. The other is feed water pump model test. The mock-up pump is composed of a rotating stainless steel disk. As a result, it is confirmed that the deposition rate depends mostly on iron concentration in water and the exfoliation rate depends mainly on pH. Applying this information, the scale deposition-growth behavior on the equipment is quantitatively expressed by the model combined of scale deposition behavior and exfoliation behavior couples with the former. These results bring effective estimation for suppressing deposition-growth by the selection of water chemistry management and/or equipment improvement in the PWR secondary system. (author)

  19. Working session 1: Tubing degradation

    International Nuclear Information System (INIS)

    Kharshafdjian, G.; Turluer, G.

    1997-01-01

    A general introductory overview of the purpose of the group and the general subject area of SG tubing degradation was given by the facilitator. The purpose of the session was described as to open-quotes develop conclusions and proposals on regulatory and technical needs required to deal with the issues of SG tubing degradation.close quotes Types, locations and characteristics of tubing degradation in steam generators were briefly reviewed. The well-known synergistic effects of materials, environment, and stress and strain/strain rate, subsequently referred to by the acronym open-quotes MESSclose quotes by some of the group members, were noted. The element of time (i.e., evolution of these variables with time) was emphasized. It was also suggested that the group might want to consider the related topics of inspection capabilities, operational variables, degradation remedies, and validity of test data, and some background information in these areas was provided. The presentation given by Peter Millet during the Plenary Session was reviewed; Specifically, the chemical aspects and the degradation from the secondary side of the steam generator were noted. The main issues discussed during the October 1995 EPRI meeting on secondary side corrosion were reported, and a listing of the potential SG tube degradations was provided and discussed

  20. 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

  1. 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.

  2. 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

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

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

    2016-01-01

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

  4. Study on Temperature Control System Based on SG3525

    Science.gov (United States)

    Cheng, Cong; Zhu, Yifeng; Wu, Junfeng

    2017-12-01

    In this paper, it uses the way of dry bath temperature to heat the microfluidic chip directly by the heating plate and the liquid sample in microfluidic chip is heated through thermal conductivity, thus the liquid sample will maintain at target temperature. In order to improve the reliability of the whole machine, a temperature control system based on SG3525 is designed.SG3525 is the core of the system which uses PWM wave produced by itself to drive power tube to heat the heating plate. The bridge circuit consisted of thermistor and PID regulation ensure that the temperature can be controlled at 37 °C with a correctness of ± 0.2 °C and a fluctuation of ± 0.1 °C.

  5. Heat exchanger tube tool

    International Nuclear Information System (INIS)

    Gugel, G.

    1976-01-01

    Certain types of heat-exchangers have tubes opening through a tube sheet to a manifold having an access opening offset from alignment with the tube ends. A tool for inserting a device, such as for inspection or repair, is provided for use in such instances. The tool is formed by a flexible guide tube insertable through the access opening and having an inner end provided with a connector for connection with the opening of the tube in which the device is to be inserted, and an outer end which remains outside of the chamber, the guide tube having adequate length for this arrangement. A flexible transport hose for internally transporting the device slides inside of the guide tube. This hose is long enough to slide through the guide tube, into the heat-exchanger tube, and through the latter to the extent required for the use of the device. The guide tube must be bent to reach the end of the heat-exchanger tube and the latter may be constructed with a bend, the hose carrying anit-friction elements at interspaced locations along its length to make it possible for the hose to negotiate such bends while sliding to the location where the use of the device is required

  6. Comparative study of water chemistry and surface oxide composition on alloy 600 steam generator tubing

    International Nuclear Information System (INIS)

    Bjoernkvist, L.; Norring, K.; Nyborg, L.

    1993-01-01

    The Ringhals 3 steam generators experience secondary IGSCC on the tubes at support plate locations. Its sister unit Ringhals 4 is so far without IGSCC. Extensive work has been carried out in order to determine the local chemistry in crevices and the composition of deposits and oxide films on the tubes. Hot soaks of the SG:s at zero power has been performed and the water chemistry in occluded crevices of the SGs was predicted to be alkaline, pH 300degreesC = 10. In addition to eddy current testing, a large number of tubes have been pulled and destructively examined. These analysis include SEM/EDS characterization of TSP crevice deposits and Auger electron spectroscopy (AES) with depth profiling to reveal the composition of the tube OD oxide film. The AES analysis show an outer oxide rich in Fe 3 O 4 , mostly deposited. The actual Alloy 600 oxide is found below the magnetite and is 1-2 μm thick. The composition profile of the oxide exhibits a Cr-depletion relative to Ni in the outer part of the oxide, whereas an enrichment is found in depth. In order to correlate the water chemistry to the oxide composition profiles and deposits on pulled tubes, reference samples were prepared in an autoclave. The environments were chosen similar to the predicted Ringhals 3 and 4 crevice chemistry. Exposure both in an alkaline (pH 320degreesC∼ 9.9) and an acidic (pH 320degreesC ∼4.3) environment, containing sodium, chloride and sulphate, was studied. Some samples were also found on the Alloy 600 samples exposed to alkaline environment. Thus the prediction of alkaline chemistry was verified. The enrichment of chromium relative to nickel was shown to be potential and time dependent resulting in an increased Cr/Ni ratio at Cr-max with increasing potential and time

  7. PWSCC in the tube expansion zone - an overview

    International Nuclear Information System (INIS)

    Hernalsteen, P.

    1993-01-01

    Most of the PWR Steam Generators (SG) with tubes in Inconel 600 alloy are affected by Primary Water Stress Corrosion Cracking (PWSCC) in the Expansion Zone (mainly the Roll Transition) of tubes mechanically expanded in the tube sheet. After a description of the defect mechanism and characterization methods, the paper reviews the various measures that can be used to prevent the problem. In-Service Inspection results are presented to illustrate the actual field experience; prediction tools are available to forecast the further SG degradation. Degraded tubes are eventually removed from service; this plugging policy undergoes presently a major evolution towards a mechanism specific approach, taking into account both structural and leakage requirements. The paper reviews various repair techniques that can be used as an alternate to plugging. Ultimately repair has to be weighed against SG replacement with a comprehensive problem management approach. (orig.)

  8. WPEC Subgroup Meetings. Joint SG38/SG39/SG40-CIELO meeting, 20 May 2015

    International Nuclear Information System (INIS)

    Brown, Dave; Palmiotti, G.; Salvatores, M.; Chadwick, M.; Mattoon, Caleb; Yokoyama, K.; Leal, L.; Diez, C.J.; Hill, I.; Ignatyuk, A.

    2015-05-01

    WPEC subgroup 38 (SG38) was formed to develop a new structure for storing nuclear reaction data, that is meant to eventually replace ENDF-6 as the standard way to store and share evaluations. The work of SG38 covers the following tasks: Designing flexible, general-purpose data containers; Determining a logical and easy-to-understand top-level hierarchy for storing evaluated nuclear reaction data; Creating a particle database for storing particles, masses and level schemes; Specifying the infrastructure (plotting, processing, etc.) that must accompany the new structure; Developing an Application Programming Interface or API to allow other codes to access data stored in the new structure; Specifying what tests need to be implemented for quality assurance of the new structure and associated infrastructure; Ensuring documentation and governance of the structure and associated infrastructure. The aim of WPEC subgroup 39 'Methods and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files' is to provide criteria and practical approaches to use effectively the results of sensitivity analyses and cross section adjustments for feedback to evaluators and differential measurement experimentalists in order to improve the knowledge of neutron cross sections, uncertainties, and correlations to be used in a wide range of applications. WPEC subgroup 40-CIELO (Collaborative International Evaluated Library Organization) provides a new working paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts from across the international nuclear reaction data community to identify and document discrepancies among existing evaluated data libraries, measured data, and model calculation interpretations, and aims to make progress in reconciling these discrepancies to create more accurate ENDF-formatted files. SG40-CIELO focusses on 6 important isotopes: "1H, "1"6O, "5"6Fe, "2"3"5","2"3"8U, "2"3"9Pu. This

  9. 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)

  10. 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.

  11. Developments for the outer tracking system of the LHCb experiment

    CERN Document Server

    Bachmann, S; Haas, T; Uwer, U; Walter, M; Wiedner, D

    2004-01-01

    The outer tracking system of the LHCb experiment is discussed. The outer tracking system (OT) is made of three stations and every station is made up of four detecting planes with a double layer of straw tubes. The straw tubes are mounted in detector module boxes made up of sandwich panels. The use of a counting gas with a high drift velocity is suggested to cope with high bunch crossing rate at the LHCb experiment. (Edited abstract) 3 Refs.

  12. Dispersant trial at ANO-2: Results from a short-term trial prior to SG replacement

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Frattini, P.; Robbins, P.; Miller, A.; Varrin, R.; Kreider, M.

    2002-01-01

    Corrosion products in the secondary side of pressurized water reactor (PWR) steam generators (SGs) primarily deposit on the SG tubes. These deposits can inhibit heat transfer, lead to thermal-hydraulic instabilities through blockage of tube supports, and create occluded regions where corrosive species can concentrate along tubes and in tube-to-tube support plate crevices. The performance of the SGs is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. A potential strategy for minimizing deposition of corrosion products on SG internal surfaces is to use an online dispersant to help prevent the corrosion products from adhering to the steam generator surfaces. By inhibiting the deposition of the corrosion products, the dispersant can facilitate more effective removal from the SGs via blowdown. This type of strategy has been employed at fossil boilers for many decades. However, due to the use of inorganic (sulfur and other impurities) polymerization initiators, polymeric dispersants had not been utilized in the nuclear industry. Only recently has a poly-acrylic acid dispersant, developed by BetzDearborn (PAA), been available that meets the criteria for nuclear application. This paper summarizes the results of the short-term PAA dispersant trial in Winter/Spring 2000, lasting approximately 3 months, performed at Arkansas nuclear one unit 2 (ANO-2)-including the chronology of the trial, the increase in blowdown iron removal efficiency with use of the dispersant, and observed effects on SG performance. (authors)

  13. Dispersant trial at ANO-2: Results from a short-term trial prior to SG replacement

    Energy Technology Data Exchange (ETDEWEB)

    Fruzzetti, K.; Frattini, P. [Electric Power Research Inst., Palo Alto, CA (United States); Robbins, P. [Entergy Operations, Arkansas Nuclear One, Russellville, AR (United States); Miller, A. [Pedro Point Technology, Inc., Pacifica, CA (United States); Varrin, R.; Kreider, M. [Dominion Engineering Inc., McLean, VA (United States)

    2002-07-01

    Corrosion products in the secondary side of pressurized water reactor (PWR) steam generators (SGs) primarily deposit on the SG tubes. These deposits can inhibit heat transfer, lead to thermal-hydraulic instabilities through blockage of tube supports, and create occluded regions where corrosive species can concentrate along tubes and in tube-to-tube support plate crevices. The performance of the SGs is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. A potential strategy for minimizing deposition of corrosion products on SG internal surfaces is to use an online dispersant to help prevent the corrosion products from adhering to the steam generator surfaces. By inhibiting the deposition of the corrosion products, the dispersant can facilitate more effective removal from the SGs via blowdown. This type of strategy has been employed at fossil boilers for many decades. However, due to the use of inorganic (sulfur and other impurities) polymerization initiators, polymeric dispersants had not been utilized in the nuclear industry. Only recently has a poly-acrylic acid dispersant, developed by BetzDearborn (PAA), been available that meets the criteria for nuclear application. This paper summarizes the results of the short-term PAA dispersant trial in Winter/Spring 2000, lasting approximately 3 months, performed at Arkansas nuclear one unit 2 (ANO-2)-including the chronology of the trial, the increase in blowdown iron removal efficiency with use of the dispersant, and observed effects on SG performance. (authors)

  14. Predicting the accumulated number of plugged tubes in a steam generator using statistical methodologies

    International Nuclear Information System (INIS)

    Ferng, Y.-M.; Fan, C.N.; Pei, B.S.; Li, H.-N.

    2008-01-01

    A steam generator (SG) plays a significant role not only with respect to the primary-to-secondary heat transfer but also as a fission product barrier to prevent the release of radionuclides. Tube plugging is an efficient way to avoid releasing radionuclides when SG tubes are severely degraded. However, this remedial action may cause the decrease of SG heat transfer capability, especially in transient or accident conditions. It is therefore crucial for the plant staff to understand the trend of plugged tubes for the SG operation and maintenance. Statistical methodologies are proposed in this paper to predict this trend. The accumulated numbers of SG plugged tubes versus the operation time are predicted using the Weibull and log-normal distributions, which correspond well with the plant measured data from a selected pressurized water reactor (PWR). With the help of these predictions, the accumulated number of SG plugged tubes can be reasonably extrapolated to the 40-year operation lifetime (or even longer than 40 years) of a PWR. This information can assist the plant policymakers to determine whether or when a SG must be replaced

  15. 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)

  16. Lathe Attachment Finishes Inner Surface of Tubes

    Science.gov (United States)

    Lancki, A. J.

    1982-01-01

    Extremely smooth finishes are machined on inside surfaces of tubes by new attachment for a lathe. The relatively inexpensive accessory, called a "microhone," holds a honing stone against workpiece by rigid tangs instead of springs as in conventional honing tools. Inner rod permits adjustment of microhoning stone, while outer tube supports assembly. Outer tube is held between split blocks on lathe toolpost. Microhoning can be done with either microhone or workpiece moving and other member stationary.

  17. 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

  18. 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

  19. Fretting wear of steam generator tubes: high-temperature tests on AECL rig

    International Nuclear Information System (INIS)

    Guerout, F.; Zbinden, M.

    1993-07-01

    The R and DD has undertaken the study of fretting-wear of Alloy 600 S.G. tubes which occurs by contact with migrating items. The test series was performed in Canada at AECL Research (Atomic Energy of Canada Limited) as part of an exchange program. Four types of configuration were envisaged: a tube-to-drilled hole support contact which provides reference results and three types of tube-to-support contacts which simulate the tube fretting-wear induced by a welding rod, a threaded rod and a knife-edge rod support. This programme is completed by the study of the contact between a S.G. tube and a neighbouring S.G. tube which has been broken after plugging. (authors). 1 tab., 3 refs

  20. Strategies and actions for the mitigation of the phenomenon DENTING in the tube sheet (TTS Denting) steam generators of the NPP Asco Denting at the top of the SG tube sheet (TTS tube denting) has recently been experienced in new and replacement SGs at several plants; Estrategias y acciones para la mitigacion del fenomeno DENTING en la placa tubular de generadores de vapor (TTS DENTING) de la Central Nuclear de Asco

    Energy Technology Data Exchange (ETDEWEB)

    Espanol Villar, J.

    2013-07-01

    It is highly likely that the accumulation of sludge (deposits) on the tube sheet is clearly associated with the denting occurrence. More specifically, it is commonly believed that an aggressive crevice environment formed within the deposits or in the shallow tight tube sheet to tube crevice below the deposits is at the origin of the denting (tube deformation), and, when present, the consequent stress corrosion cracking (SCC). There are described a set of strategies that have been followed since the emergence of the TTS denting phenomenon on Steam Generator of the Nuclear Power Plants Asco I and II, influenced by the presence of hard sludge in the tube plate of Steam Generators, their results and the evolution of the phenomenon in relation to the various measures taken.

  1. Pressure tube reactor

    International Nuclear Information System (INIS)

    Susuki, Akira; Murata, Shigeto; Minato, Akihiko.

    1993-01-01

    In a pressure tube reactor, a reactor core is constituted by arranging more than two units of a minimum unit combination of a moderator sealing pipe containing a calandria tube having moderators there between and a calandria tube and moderators. The upper header and a lower header of the calandria tank containing moderators are communicated by way of the moderator sealing tube. Further, a gravitationally dropping mechanism is disposed for injecting neutron absorbing liquid to a calandria gas injection portion. A ratio between a moderator volume and a fuel volume is defined as a function of the inner diameter of the moderator sealing tube, the outer diameter of the calandria tube and the diameter of fuel pellets, and has no influence to intervals of a pressure tube lattice. The interval of the pressure tube lattice is enlarged without increasing the size of the pressure tube, to improve production efficiency of the reactor and set a coolant void coefficient more negative, thereby enabling to improve self controllability and safety. Further, the reactor scram can be conducted by injecting neutron absorbing liquid. (N.H.)

  2. Degradation of Alloy 800 steam generator tubing and its long-term behaviour predictions for plant life management

    International Nuclear Information System (INIS)

    Lu, Y.C.; Tapping, R.L.; Pandey, M.D.

    2009-01-01

    Alloy 800 tubing has a good service record in steam generators (SGs) in both German pressurized water reactors and CANDU 6 reactors, however, a recent comprehensive examination of several ex-service SG tubes removed from Darlington Nuclear Generating Station (DNGS) found that these SG tubes (which had experienced shallow pitting in service) were more susceptible to pitting corrosion in laboratory tests than a reference nuclear grade Alloy 800 tubing under SG crevice chemistry conditions. This was an unexpected finding and has raised questions about possible effects of in-service 'aging' on SG tubing. In addition, there has also been recent evidence that a few Alloy 800 tubes have experienced stress corrosion cracking (SCC) in some German pressurized water reactors (PWRs), possibly after many years of degradation-free service, although the inspection history of these tubes is not available to confirm that the reported degradation initiated recently. These findings suggest that Alloy 800 tubing may have some aging degradation susceptibility after many years of service. To provide support for a proactive SG aging management, 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. Experimental work was also performed to investigate the root cause leading to Alloy 800 SG tubing degradation. The results from this study suggested that a combination of negative factors; aggressive chemistry resulting from impurity ingress into the secondary side of the SGs, elevated electrochemical corrosion potential (ECP) during SG transients and surface strain/plastic deformation, might have led to the degradation of the ex-service SG tubing. The studies have shown that each of these conditions in isolation does not cause degradation of Alloy 800 SG tubing; a synergistic combination of factors is required. The OPEX and experimental

  3. 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)

  4. Removal of Secondary Side Deposit and Foreign Objects in SG of Yonggwang Nuclear Power Plant, Unit 2

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Wootae [KHNP-CRI, Daejeon (Korea, Republic of); Yun, Sangjung; Choi, Yongseok; Kim, Taehwa; Seo, Hongchang [Byucksan Digital Valley II, Seoul (Korea, Republic of)

    2014-05-15

    Mock-up test and training before the service were made to minimize operation time and radiation dose. 8 bag filters and 49 cartridge filters were consumed for filtering of soft and hard sludge in SG A, B, and C. Total 21.05 kilograms of sludge was removed. Five foreign objects were found and removed in tube bundle of SG A, in annulus and tube bundle of SG B, and in tube bundle of SG C. During the 20{sup th} OH of Yonggwang NPP unit 2, we removed 21.05 kilograms of sludge and five foreign objects from three steam generators. Gradual increase in weight of sludge removed is assumed to show us that more sludge is created as operation time of steam generator increases. Low level chemical cleaning at 19{sup th} OH removed 517 kilograms of sludge which was dominant of all the sludge removal weight. We could assume, from this fact, that high level chemical cleaning could remove significant amount of sludge compared to mechanical lancing method. Five foreign objects which were removed from inside of SG showed us that more thorough inside cleaning and inspection is necessary during fabrication of steam generator.

  5. Simulation of the VISTA SG heat transfer experiment using MIDAS/SMR

    International Nuclear Information System (INIS)

    Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

    2011-01-01

    As the SMART plant was designed with the helical type tubes in the steam generators, the heat transfer model in that geometry has been implemented in the TASS/SMR-S code and used for the safety analysis. The same correlation was implemented in the MIDAS/SMR, which is being used for the severe accident analyses, to model heat transfer at the steam generators. In this study, the VISTA SG experiment with the helical steam generator tube was simulated with MIDAS/SMR to compare the heat transfer rates through the helical tube

  6. Safety assessment of the potential for foreign object - caused fretting - wear damages on PWR steam generator U-tubes

    International Nuclear Information System (INIS)

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

    2003-01-01

    This study investigates 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 are obtained from three-dimensional SG flow calculation using the ATHOS3 code. Modal analyses are performed for the finite element modelings of U-tubes to get the natural frequency, corresponding mode shape and participation factor. 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. Also, discussed in this study are the effects of flow velocity, internal pressure, tube-to-foreign object contact angle, and vibration of the tube on the remaining life of the tube

  7. 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

  8. 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

  9. SCC analysis of Alloy 600 tubes from a retired steam generator

    Science.gov (United States)

    Hwang, Seong Sik; Kim, Hong Pyo

    2013-09-01

    Steam generators (SG) equipped with Alloy 600 tubes of a Korean nuclear power plants were replaced with a new one having Alloy 690 tubes in 1998 after 20 years of operation. To set up a guide line for an examination of the other SG tubes, a metallographic examination of the defected tubes was carried out. A destructive analysis on 71 tubes was addressed, and a relation among the stress corrosion crack (SCC) defect location, defect depth, and location of the sludge pile was obtained. Tubes extracted from the retired SG were transferred to a hot laboratory. Detailed nondestructive analysis examinations were taken again at the laboratory, and the tubes were then destructively examined. The types and sizes of the cracks were characterized. The location and depth of the SCC were evaluated in terms of the location and height of the sludge. Most axial cracks were in the sludge pile, whereas the circumferential ones were around the top of the tube sheet (TTS) or below the TTS. Average defect depth of the axial cracks was deeper than that of the circumferential ones. Axial cracks at tube support plate (TSP) seem to be related with corrosion/sludge in crevice like at the TTS region. Circumferential cracks at TSP seem to be caused by tube denting at the upper part of the TSP. Tubes not having clear ECT signals for quantifying an ECT data-base. Tubes having no ECT signal. Tubes with a large ECT signal. Tubes with various types and sizes of flaws (primary water stress corrosion cracking (PWSCC), outside diameter stress corrosion cracking (ODSCC), Pit). Tubes with distinct PWSCC or ODSCC. Tubes were extracted from the RSG based on the field ECT with the criteria, and transferred to a hot laboratory at the Korea Atomic Energy Research Institute (KAERI) for destructive examination. A comprehensive ECT inspection was performed again at the hot laboratory to confirm the location of the cracks obtained from a field inspection. These exact locations of the defects were marked on the

  10. Selection of number of SG modules per secondary sodium loop for a Prototype SFR

    International Nuclear Information System (INIS)

    Kim, Eui Kwang; Yoo, Jae-Woon; Joo, Hyung-Kook; Lee, Tae-Ho

    2014-01-01

    Steam generator is a vertical counter flow shell and tube heat exchanger with sodium on shell side and water-steam in tubes. The concept of steam generator modularization is discussed and the advantages of large and small steam generator modules are presented. To estimate the economics of the steam generator modularization for the PGSFR, detailed optimization studies were carried out for 1, 2, 3, 4 and 5 steam generator modules per loop, taking into consideration the effect of capital cost, operating cost, and outage cost. For the economics of the steam generator modularization for a PGSFR, detailed optimization studies were carried out for 1, 2, 3, 4 and 5 steam generator modules per loop. The effect of capital cost, operating cost, and outage cost were taken into consideration. The studies indicated that the design with less SG modules is found to be more economical, if the weld failure rate is below 2.0x1E-5 failure/year. From the economics of the modularization, thermal rate of SG, reliability, operability considerations, appropriate number of SG modules required for PGSFR will be 1 SG module per loop

  11. Measuring of tube expansion

    International Nuclear Information System (INIS)

    Vogeleer, J. P.

    1985-01-01

    The expansion of the primary tubes or sleeves of the steam generator of a nuclear reactor plant are measured while the tubes or sleeves are being expanded. A primary tube or sleeve is expanded by high pressure of water which flows through a channel in an expander body. The water is supplied through an elongated conductor and is introduced through a connector on the shank connected to the conductor at its outer end. A wire extends through the mandrel and through the conductor to the end of the connector. At its inner end the wire is connected to a tapered pin which is subject to counteracting forces produced by the pressure of the water. The force on the side where the wire is connected to the conductor is smaller than on the opposite side. The tapered pin is moved in the direction of the higher force and extrudes the wire outwardly of the conductor. The tapered surface of the tapered pin engages transverse captive plungers which are maintained in engagement with the expanding tube or sleeve as they are moved outwardly by the tapered pin. The wire and the connector extend out of the generator and, at its outer end, the wire is connected to an indicator which measures the extent to which the wire is moved by the tapered pin, thus measuring the expansion of the tube or sleeve as it progresses

  12. Verification of neutron irradiation on S/G tube materials

    International Nuclear Information System (INIS)

    Kang, Byoung Hwi; Lee, S. K.; Jang, D. Y.; Jo, K. H.

    2010-12-01

    The fluence monitors were fabricated with metal wires of the purity ≥ 99.9%, whose dimensions were 0.1mm diameter, about 3mm length, and around 150-200 μg mass range. Three wire samples (Fe, Ni, Ti) were prepared for one irradiation aluminum capsule. Five capsules were irradiated in the OR5 hole of the HANARO reactor at 30 MW power for about 25 days. The reaction rates were calculated by using the measured radiation activity data, and then neutron fluence were obtained from the reaction rates and the weighted neutron cross section with calculated neutron spectrum at the fluence monitor position. The measured neutron fluences were compared to the calculated ones. (Errors ≤ 35%)

  13. Singular deposit formation in PWR due to electrokinetic phenomena - application to SG clogging

    Energy Technology Data Exchange (ETDEWEB)

    Guillodo, M.; Muller, T.; Barale, M.; Foucault, M. [AREVA NP SAS, Technical Centre (France); Clinard, M.-H.; Brun, C.; Chahma, F. [AREVA NP SAS, Chemistry and Radiochemistry Group (France); Corredera, G.; De Bouvier, O. [Electricite de France, Centre d' Expertise de I' inspection dans les domaines de la Realisation et de l' Exploitation (France)

    2009-07-01

    The deposits which cause clogging of the 'foils' of the tube support plates (TSP) in Steam Generators (SG) of PWR present two characteristics which put forward that the mechanism at the origin of their formation is different from the mechanism that drives the formation of homogeneous deposits leading to the fouling of the free spans of SG tubes. Clogging occurs near the leading edge of the TSP and the deposits appear as diaphragms localized between both TSP and SG tubing materials, while the major part of the tube/TSP interstice presents little or no significant clogging. This type of deposit seems rather comparable to the ones which were reproduced in Lab tests to explain the flow rate instabilities observed on a French unit during hot shutdown in the 90's. The deposits which cause TSP clogging are owed to a discontinuity of the streaming currents in the vicinity of a surface singularity (orifices, scratches ...) which, in very low conductivity environment, produce local potential variations and/or current loop in the metallic pipe material due to electrokinetic effects. Deposits can be built by two mechanisms which may or not coexist: (i) accumulation of particles stabilized by an electrostatic attraction due to the local variation of electrokinetic potential, and (ii) crystalline growth of magnetite produced by the oxidation of ferrous ions on the anodic branch of a current loop. Lab investigations carried out by AREVA NP Technical Centre since the end of the 90's showed that this type of deposit occurs when the redox potential is higher than a critical value, and can be gradually dissolved when the potential becomes lower than this value which depends on the 'Material - Chemistry' couple. Special emphasis will be given in this paper to the TSP clogging of SG in PWR secondary coolant dealing particularly with the potential strong effect of electrokinetic phenomena in low conductive environment and in high temperature conditions

  14. The Outer Space Treaty

    Science.gov (United States)

    Johnson, Christopher Daniel

    2018-01-01

    Negotiated at the United Nations and in force since 1967, the Outer Space Treaty has been ratified by over 100 countries and is the most important and foundational source of space law. The treaty, whose full title is "Treaty on Principles Governing the Activities of States in the Exploration and Use of Outer Space, Including the Moon and Other Celestial Bodies," governs all of humankind's activities in outer space, including activities on other celestial bodies and many activities on Earth related to outer space. All space exploration and human spaceflight, planetary sciences, and commercial uses of space—such as the global telecommunications industry and the use of space technologies such as position, navigation, and timing (PNT), take place against the backdrop of the general regulatory framework established in the Outer Space Treaty. A treaty is an international legal instrument which balances rights and obligations between states, and exists as a kind of mutual contract of shared understandings, rights, and responsibilities between them. Negotiated and drafted during the Cold War era of heightened political tensions, the Outer Space Treaty is largely the product of efforts by the United States and the USSR to agree on certain minimum standards and obligations to govern their competition in "conquering" space. Additionally, the Outer Space Treaty is similar to other treaties, including treaties governing the high seas, international airspace, and the Antarctic, all of which govern the behavior of states outside of their national borders. The treaty is brief in nature and only contains 17 articles, and is not comprehensive in addressing and regulating every possible scenario. The negotiating states knew that the Outer Space Treaty could only establish certain foundational concepts such as freedom of access, state responsibility and liability, non-weaponization of space, the treatment of astronauts in distress, and the prohibition of non-appropriation of

  15. 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

  16. 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

  17. Heat treated tube for cladding nuclear fuel element

    International Nuclear Information System (INIS)

    Eddens, F.C.; White, D.W.; Harmon, J.L.

    1983-01-01

    The zirconium alloy tube comprises a metallurgical gradient across the width of the tube wall wherein the tube has a more corrosion-resistant metallurgical condition at the outer circumference and a less corrosion-resistant metallurgical condition at the inner circumference. The metallurgical gradient can be generated by heating an outer circumferential portion of the tube to the high alpha or mixed alpha plus beta range while maintaining the inner surface at a lower temperature, followed by cooling of the tube. Preferably the tube is made of Zircaloy. (author)

  18. 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)

  19. Saturn's outer magnetosphere

    Science.gov (United States)

    Schardt, A. W.; Behannon, K. W.; Carbary, J. F.; Eviatar, A.; Lepping, R. P.; Siscoe, G. L.

    1983-01-01

    Similarities between the Saturnian and terrestrial outer magnetosphere are examined. Saturn, like Earth, has a fully developed magnetic tail, 80 to 100 RS in diameter. One major difference between the two outer magnetospheres is the hydrogen and nitrogen torus produced by Titan. This plasma is, in general, convected in the corotation direction at nearly the rigid corotation speed. Energies of magnetospheric particles extend to above 500 keV. In contrast, interplanetary protons and ions above 2 MeV have free access to the outer magnetosphere to distances well below the Stormer cutoff. This access presumably occurs through the magnetotail. In addition to the H+, H2+, and H3+ ions primarily of local origin, energetic He, C, N, and O ions are found with solar composition. Their flux can be substantially enhanced over that of interplanetary ions at energies of 0.2 to 0.4 MeV/nuc.

  20. State-of-the-art review of OPG steam generator tubing degradation mechanisms

    International Nuclear Information System (INIS)

    Brennenstuhl, A.M.; Ramamurthy, S.; Good, G.M.

    2009-01-01

    Steam generator (SG) degradation has been a major cause of pressurized water reactor (PWR) incapability world-wide and has limited the useful life of SGs at some utilities. The vast majority of the degradation has been the result of SCC of the thin walled nickel alloy SG tubes and has been most prevalent in mill annealed (MA) Alloy 600. Fortunately, Ontario Power Generation (OPG) SG tubes are manufactured from alloys that have much better resistance to this form of localized corrosion than Alloy 600MA and as a consequence have not encountered SCC to date. Other forms of degradation nevertheless have been experienced; some units at Pickering - B in particular have had many Alloy 400 SG tubes removed from service due to severe underdeposit corrosion (UDC) and costly modifications have been made to Darlington SGs to prevent leaks as a result of SG tube fretting-wear at tube supports. Degradation other than UDC and fretting-wear which could pose a threat to the future reliable operation of OPG's nuclear fleet has also been observed. Important activities in effectively managing SG degradation include determining the mode of degradation and arriving at an understanding of the contributing factors. This is done by a combination of non-destructive examination (NDE) of SG tubing in-situ, SG tube removals for metallurgical examination and research and development. SG tube metallurgical examinations provide information that can be used in the timely development of a strategy dealing with the degradation in the short to intermediate timeframe. Determining the main causative factors at a mechanistic level helps to improve the predictive capability and increases the probability of dealing with the problem in the most cost-effective way. OPG has used this approach together with in-situ NDE inspections during planned outages of its nuclear reactors to minimize the possibility of unscheduled outages and provide the best possible fitness-for-service assessments. Many metallurgical

  1. Tube plug

    International Nuclear Information System (INIS)

    Zafred, P. R.

    1985-01-01

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

  2. 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

  3. Pressure tube reactor

    International Nuclear Information System (INIS)

    Seki, Osamu; Kumasaka, Katsuyuki.

    1988-01-01

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

  4. Gastrostomy Tube (G-Tube)

    Science.gov (United States)

    ... any of these problems: a dislodged tube a blocked or clogged tube any signs of infection (including redness, swelling, or warmth at the tube site; discharge that's yellow, green, or foul-smelling; fever) excessive bleeding or drainage from the tube site severe abdominal pain lasting ...

  5. Sludge Removal and Retrieval of Foreign Object in SG of Kori Nuclear Power Plant, Unit 4

    International Nuclear Information System (INIS)

    Jeong, Wootae; Kim, Sangtae; Kim, Youngkug; Kang, Seokchul

    2014-01-01

    Sludge deposit was removed and foreign objects were inspected and retrieved on secondary side tube sheet of the SG during January 23 and February 22, April 15 and 27 in 2013. FOLAS-I lancing system, video probe and retrieval tools were used for lancing and foreign object removal respectively. Operators of the lancing system participated in mock-up training before doing the service to minimize operation time and radiation dose. Foreign objects were searched on top of 7 th TSP (tube support plate), on annulus and in tube bundle. Four objects were found and removed on annulus and in tube bundle. During the 21 st OH of Kori NPP unit 4, we removed 345.9 kilo gram of sludge and four foreign objects from three steam generators. Foreign objects which were removed from inside of SG showed us that relatively large object such as the hooked bolt might exists in steam generators. We can conclude that identifying and removing foreign object is very important to avoid possible tube failure. Removing circular metal of 152.4 gram also was successfully removed

  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. Development of sputter ion pump based SG leak detection system for Fast Breeder Test Reactor

    International Nuclear Information System (INIS)

    Babu, B.; Sureshkumar, K.V.; Srinivasan, G.

    2013-01-01

    Highlights: ► Development and commissioning of SG leak detection system for FBTR. ► Development of Robust method of using sputter ion pump based system. ► Modifications for improving reliability and availability. ► On line injection of hydrogen in sodium during reactor operation. ► Triplication of the SG leak detection system. - Abstract: The Fast Breeder Test Reactor (FBTR) is a 40 MWt, loop type sodium cooled fast reactor built at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam as a fore-runner to the second stage of Indian nuclear power programme. The reactor design is based on the French reactor Rapsodie with several modifications which include the provision of a steam-water circuit and turbo-generator. FBTR uses sodium as the coolant in the main heat transport medium to transfer heat from the reactor core to the feed water in the tertiary loop for producing superheated steam, which drives the turbo-generator. Sodium and water flow in shell and tube side respectively, separated by thin-walls of the ferritic steel tubes of the once-through steam generator (SG). Material defects in these tubes can lead to leakage of water into sodium, resulting in sodium water reactions leading to undesirable consequences. Early detection of water or steam leaks into sodium in the steam generator units of liquid metal fast breeder reactors (LMFBR) is an important requirement from safety and economic considerations. The SG leak in FBTR is detected by Sputter Ion Pump (SIP) based Steam Generator Leak Detection (SGLD) system and Thermal Conductivity Detector (TCD) based Hydrogen in Argon Detection (HAD) system. Many modifications were carried out in the SGLD system for the reactor operation to improve the reliability and availability. This paper details the development and the acquired experience of SIP based SGLD system instrumentation for real time hydrogen detection in sodium for FBTR.

  10. Tube-in-shell heat exchangers

    International Nuclear Information System (INIS)

    Richardson, J.

    1976-01-01

    Tube-in-shell heat exchangers normally comprise a bundle of parallel tubes within a shell container, with a fluid arranged to flow through the tubes in heat exchange with a second fluid flowing through the shell. The tubes are usually end supported by the tube plates that separate the two fluids, and in use the tube attachments to the tube plates and the tube plates can be subject to severe stress by thermal shock and frequent inspection and servicing are required. Where the heat exchangers are immersed in a coolant such as liquid Na such inspection is difficult. In the arrangement described a longitudinally extending central tube is provided incorporating axially spaced cylindrical tube plates to which the opposite ends of the tubes are attached. Within this tube there is a tubular baffle that slidably seals against the wall of the tube between the cylindrical tube plates to define two co-axial flow ducts. These ducts are interconnected at the closed end of the tube by the heat exchange tubes and the baffle comprises inner and outer spaced walls with the interspace containing Ar. The baffle is easily removable and can be withdrawn to enable insertion of equipment for inspecting the wall of the tube and tube attachments and to facilitate plugging of defective tubes. Cylindrical tube plates are believed to be superior for carrying pressure loads and resisting the effects of thermal shock. Some protection against thermal shock can be effected by arranging that the secondary heat exchange fluid is on the tube side, and by providing a thermal baffle to prevent direct impingement of hot primary fluid on to the cylindrical tube plates. The inner wall of the tubular baffle may have flexible expansible region. Some nuclear reactor constructions incorporating such an arrangement are described, including liquid metal reactors. (U.K.)

  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. Performance of the LHCb Outer Tracker

    CERN Document Server

    Arink, R; Bachmann, S.; Bagaturia, Y.; Band, H.; Bauer, Th.; Berkien, A.; Farber, Ch.; Bien, A.; Blouw, J.; Ceelie, L.; Coco, V.; Deckenhoff, M.; Deng, Z.; Dettori, F.; van Eijk, D.; Ekelhof, R.; Gersabeck, E.; Grillo, L.; Hulsbergen, W.D.; Karbach, T.M.; Koopman, R.; Kozlinskiy, A.; Langenbruch, Ch.; Lavrentyev, V.; Linn, Ch.; Merk, M.; Merkel, J.; Meissner, M.; Michalowski, J.; Morawski, P.; Nawrot, A.; Nedos, M.; Pellegrino, A.; Polok, G.; van Petten, O.; Rovekamp, J.; Schimmel, F.; Schuylenburg, H.; Schwemmer, R.; Seyfert, P.; Serra, N.; Sluijk, T.; Spaan, B.; Spelt, J.; Storaci, B.; Szczekowski, M.; Swientek, S.; Tolk, S.; Tuning, N.; Uwer, U.; Wiedner, D.; Witek, M.; Zeng, M.; Zwart, A.

    2014-01-01

    The LHCb Outer Tracker is a gaseous detector covering an area of 5x6 m2 with 12 double layers of straw tubes. The detector with its services are described together with the commissioning and calibration procedures. Based on data of the first LHC running period from 2010 to 2012, the performance of the readout electronics and the single hit resolution and efficiency are presented. The efficiency to detect a hit in the central half of the straw is estimated to be 99.2%, and the position resolution is determined to be approximately 200 um. The Outer Tracker received a dose in the hottest region corresponding to 0.12 C/cm, and no signs of gain deterioration or other ageing effects are observed.

  13. Test Station for Measuring Aluminum Tube Geometrical Parameters

    CERN Document Server

    Oansea, D; Gongadze, A L; Gostkin, M I; Dedovich, D V; Evtoukhovitch, P G; Comanescu, B; Kotov, S A; Necsoiu, T; Potrap, I N; Rogalev, E V; Tskhadadze, E G; Chelkov, G A

    2001-01-01

    A test station for quality control of aluminum tube outer diameter and wall thickness is presented. The tested tubes are used for drift detector assembly of ATLAS (LHC, CERN) muon system. The outer diameter and wall thickness of aluminium tubes are measured by means of noncontact optical and ultrasonic methods respectively with the accuracy of 3 {\\mu}m. The testing process is automatic and interacts with the production data base.

  14. TPS for Outer Planets

    Science.gov (United States)

    Venkatapathy, Ethiraj; Ellerby, D.; Gage, P.; Gasch, M.; Hwang, H.; Prabhu, D.; Stackpoole, M.; Wercinski, Paul

    2018-01-01

    This invited talk will provide an assessment of the TPS needs for Outer Planet In-situ missions to destinations with atmosphere. The talk will outline the drivers for TPS from destination, science, mission architecture and entry environment. An assessment of the readiness of the TPS, both currently available and under development, for Saturn, Titan, Uranus and Neptune are provided. The challenges related to sustainability of the TPS for future missions are discussed.

  15. Designing sgRNAs with CRISPy web

    DEFF Research Database (Denmark)

    Blin, Kai; Lee, Sang Yup; Weber, Tilmann

    2017-01-01

    Tilmann Weber’s group at the Novo Nordisk Foundation Center for Biosustainability developed a user-friendly, web server implementation of the sgRNA prediction software, CRISPy, for non-computer scientists.......Tilmann Weber’s group at the Novo Nordisk Foundation Center for Biosustainability developed a user-friendly, web server implementation of the sgRNA prediction software, CRISPy, for non-computer scientists....

  16. 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)

  17. Outer Continental Shelf Lands Act

    Data.gov (United States)

    National Oceanic and Atmospheric Administration, Department of Commerce — This data represents geographic terms used within the Outer Continental Shelf Lands Act (OCSLA or Act). The Act defines the United States outer continental shelf...

  18. 30 CFR 250.1626 - Tubing and wellhead equipment.

    Science.gov (United States)

    2010-07-01

    ... 30 Mineral Resources 2 2010-07-01 2010-07-01 false Tubing and wellhead equipment. 250.1626 Section... GAS AND SULPHUR OPERATIONS IN THE OUTER CONTINENTAL SHELF Sulphur Operations § 250.1626 Tubing and wellhead equipment. (a) No tubing string shall be placed into service or continue to be used unless such...

  19. Evaluation of critical current density and residual resistance ratio limits in powder in tube Nb$_{3}$Sn conductors

    CERN Document Server

    Segal, Christopher; Sung, Zu Hawn; Lee, Peter J; Sailer, Bernd; Thoener, Manfred; Schlenga, Klaud; Ballarino, Amalia; Bottura, Luca; Bordini, Bernardo; Scheuerlein, Christian; Larnalestier, David C

    2016-01-01

    High critical current density ( Jc) Nb$_{3}$Sn A15 multifilamentary wires require a large volume fraction of small grain (SG), superconducting A15 phase, as well as Cu stabilizer with high Residual Resistance Ratio (RRR) to provide electromagnetic stabilization and protection. In powder-in-tube (PIT) wires the unreacted Nb7.5 wt%Ta outer layer of the tubular filaments acts as a diffusion barrier and protects the interfilamentary Cu stabilizer from Sn contamination. A high RRR requirement generally imposes a restricted A15 reaction heat treatment to prevent localized full reaction of the filament that could allow Sn to reach the Cu. In this study we investigate recent high quality PIT wires that achieve a Jc (12 T, 4.2 K) up to ∼2500 A mm−2 and find that the minimum diffusion barrier thickness decreases as the filament aspect ratio increases from ∼1 in the inner rings of filaments to 1.3 in the outer filament rings. We found that just 2–3 diffusion barrier breaches can degrade RRR from 300 to 150 or le...

  20. 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)

  1. 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)

  2. 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.

  3. Management of outer space

    Science.gov (United States)

    Perek, Lubos

    1993-10-01

    Various aspects of space-environment management are discussed. Attention is called to the fact that, while space radio communications are already under an adequate management by the International Communications Union, the use of nuclear power sources is regulated by the recently adopted set of principles, and space debris will be discussed in the near future at the UN COPUOS, other aspects of management of outer space received little or no attention of the international community. These include the competency of crews and technical equipment of spacecraft launched by newcomers to space exploration; monitoring of locations and motions of space objects (now in national hands), with relevant data made accessible through a computer network; and the requirement to use space only for beneficial purposes and not for promoting narrow and debatable interests damaging the outer space environment and impeding on astronomical observations. It is suggested that some of these tasks would be best performed by an international space agency within the UN system of organizations.

  4. 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.

  5. 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.

  6. 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

  7. Outer atmospheric research

    International Nuclear Information System (INIS)

    Anderson, J.L.

    1988-01-01

    The region above the earth from about 90 km to 150 km is a major part of the upper or outer atmosphere. It is relatively unexplored, being too high for balloons or aircraft and too low for persistent orbiting spacecraft. However, the concept of a tethered subsatellite, deployed downward from an orbiting, more massive craft such as the Space Shuttle, opens the possibility of a research capability that could provide global mapping of this region. The need for research in this thick spherical shell above the earth falls into two major categories: (1) scientific data for understanding and modeling the global atmosphere and thereby determining its role in the earth system, and (2) engineering data for the design of future aerospace vehicles that will operate there. This paper presents an overview and synthesis of the currently perceived research needs and the state-of-the-art of the proposed tethered research capability. 16 references

  8. Tube-AVB gap measurements using an eddy current rotating probe

    International Nuclear Information System (INIS)

    Badson, F.; Chiron, D.; Trumpff, B.

    1988-01-01

    The wears of tubes due to flow induced vibrations have been observed after a few years of operating PWR steam generators (SG). The vibration and wear are intimately related to the gap between tubes and anti-vibration bars (AVB's) located in the bundle. The authors report the development of an eddy current (EC) method for the measurement of this gap. The method is based on using an EC probe rotating in the tube. Since for each measurement zone the tube is interacting with two AVB's the use of a rotating EC probe is necessary to perform separate and accurate measurements of each tube-AVB gap

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  10. Precision heat forming of tetrafluoroethylene tubing

    Science.gov (United States)

    Ruiz, W. V.; Thatcher, C. S. (Inventor)

    1981-01-01

    An invention that provides a method of altering the size of tetrafluoroethylene tubing which is only available in limited combination of wall thicknesses and diameter are discussed. The method includes the steps of sliding the tetrafluoroethylene tubing onto an aluminum mandrel and clamping the ends of the tubing to the mandrel by means of clamps. The tetrafluorethylene tubing and mandrel are then placed in a supporting coil which with the mandrel and tetrafluorethylene tubing are then positioned in a insulated steel pipe which is normally covered with a fiber glass insulator to smooth out temperature distribution therein. The entire structure is then placed in an event which heats the tetrafluorethylene tubing which is then shrunk by the heat to the outer dimension of the aluminum mandrel. After cooling the aluminum mandrel is removed from the newly sized tetrafluorethylene tubing by a conventional chemical milling process.

  11. More Thoughts on AG-SG Comparisons and SG Scale Factor Determinations

    Science.gov (United States)

    Crossley, David; Calvo, Marta; Rosat, Severine; Hinderer, Jacques

    2018-03-01

    We revisit a number of details that arise when doing joint AG-SG (absolute gravimeter-superconducting gravimeter) calibrations, focusing on the scale factor determination and the AG mean value that derives from the offset. When fitting SG data to AG data, the choice of which time span to use for the SG data can make a difference, as well as the inclusion of a trend that might be present in the fitting. The SG time delay has only a small effect. We review a number of options discussed recently in the literature on whether drops or sets provide the most accurate scale factor, and how to reject drops and sets to get the most consistent result. Two effects are clearly indicated by our tests, one being to smooth the raw SG 1 s (or similar sampling interval) data for times that coincide with AG drops, the other being a second pass in processing to reject residual outliers after the initial fit. Although drops can usefully provide smaller SG calibration errors compared to using set data, set values are more robust to data problems but one has to use the standard error to avoid large uncertainties. When combining scale factor determinations for the same SG at the same station, the expected gradual reduction of the error with each new experiment is consistent with the method of conflation. This is valid even when the SG data acquisition system is changed, or different AG's are used. We also find a relationship between the AG mean values obtained from SG to AG fits with the traditional short-term AG (`site') measurements usually done with shorter datasets. This involves different zero levels and corrections in the AG versus SG processing. Without using the Micro-g FG5 software it is possible to use the SG-derived corrections for tides, barometric pressure, and polar motion to convert an AG-SG calibration experiment into a site measurement (and vice versa). Finally, we provide a simple method for AG users who do not have the FG5-software to find an internal FG5 parameter that

  12. More Thoughts on AG-SG Comparisons and SG Scale Factor Determinations

    Science.gov (United States)

    Crossley, David; Calvo, Marta; Rosat, Severine; Hinderer, Jacques

    2018-05-01

    We revisit a number of details that arise when doing joint AG-SG (absolute gravimeter-superconducting gravimeter) calibrations, focusing on the scale factor determination and the AG mean value that derives from the offset. When fitting SG data to AG data, the choice of which time span to use for the SG data can make a difference, as well as the inclusion of a trend that might be present in the fitting. The SG time delay has only a small effect. We review a number of options discussed recently in the literature on whether drops or sets provide the most accurate scale factor, and how to reject drops and sets to get the most consistent result. Two effects are clearly indicated by our tests, one being to smooth the raw SG 1 s (or similar sampling interval) data for times that coincide with AG drops, the other being a second pass in processing to reject residual outliers after the initial fit. Although drops can usefully provide smaller SG calibration errors compared to using set data, set values are more robust to data problems but one has to use the standard error to avoid large uncertainties. When combining scale factor determinations for the same SG at the same station, the expected gradual reduction of the error with each new experiment is consistent with the method of conflation. This is valid even when the SG data acquisition system is changed, or different AG's are used. We also find a relationship between the AG mean values obtained from SG to AG fits with the traditional short-term AG (`site') measurements usually done with shorter datasets. This involves different zero levels and corrections in the AG versus SG processing. Without using the Micro-g FG5 software it is possible to use the SG-derived corrections for tides, barometric pressure, and polar motion to convert an AG-SG calibration experiment into a site measurement (and vice versa). Finally, we provide a simple method for AG users who do not have the FG5-software to find an internal FG5 parameter that

  13. 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)

  14. Improved management of SG BD demineralizer for reduced generation of low-level radioactive spent resin in Korean nuclear power plants

    International Nuclear Information System (INIS)

    Rhee, I.; Cho, D.; Yeon, J.

    2003-01-01

    Most nuclear power plants in Korea have adopted Ethanolamine(ETA) as a secondary pH control agent to increase the pH at the liquid phase, which may reduce the corrosion in steam generator tubes and moisture separator/reheat system. Along with its beneficial effect of SG protection from corrosion and degradation, the replacement of ammonia with ETA causes the increased generation of spent resin and the reduced run time of demineralizer in steam generator blowdown(SG BD) system. The composition ratio of cation- to anion- exchange resin in SG BD mixed bed should be increased in the ETA chemistry environment to meet the ratio of cation to anion in the aqueous solution, which results in the simultaneous exhaustion of cation and anion exchange resins. The utilization rate of mixed bed is greatest at the cation-to-anion ratio of 95:1 on the theoretical equivalent basis in the solution, but practically highest at that of 22:1 due to the possible inhomogeneous distribution of cation and anion exchange resins in SG BD bed. The run time of the bed could be extended by 30% such that, at that much, the purchase cost of new resin is saved and the production rate of spent resin is reduced. The guideline on the replacement of resin in SG BD bed is not necessary to secure the removal of radioactive particles without the leakage of the primary coolant into the secondary side since all the radioactive ions can be eliminated by SG BD bed with the sufficient time. They are retained during more than one month after their ingress into the SG BD bed without leakage. With the reduced replacement, thus, the SG BD spent resin that comprises 65% of low-level radioactive solid waste can be much cut down

  15. Inclusive search for b → sg

    International Nuclear Information System (INIS)

    1998-07-01

    The authors describe an inclusive search for flavor changing neutral current decays of the type b → s gluon in the SLD experiment. Models of b → sg indicate that the production of high momentum kaons is enhanced over background from standard B decays. If the branching ratio for b → sg is ∼ 10%, then such an enhancement should have a good signal to background ratio. The analysis makes use of the particle identification and high precision vertexing capabilities of SLD to search for such an enhancement. The data sample consists of 300K hadronic Z 0 decays collected between 1993 and 1997

  16. Control rod guide tube assemblies

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    A nuclear fuel assembly including sleeves telescoped over end portions of control rod guide tubes which bear against internal shoulders of the sleeves. Upper ends of the sleeves protrude beyond a control rod guide tube spider and are locked in place by means of a resilient cellular lattice or lock that is seated in mating grooves in the outer surfaces of the sleeves. A grapple is provided for disengaging the entire lock structure spider and associated washers, springs and a grill from the end of the fuel assembly in order to enable these components to be removed and subsequently replaced on the fuel assembly after inspection and repair. (UK)

  17. Dispersant trial at ANO-2: Qualification for a short-term trial prior to SG replacement

    International Nuclear Information System (INIS)

    Fruzzetti, K.; Frattini, P.; Robbins, P.; Miller, A.; Varrin, R.; Kreider, M.

    2002-01-01

    Corrosion products in the secondary side of pressurized water reactor (PWR) steam generators (SGs) primarily deposit on the SG tubes. These deposits can inhibit heat transfer, lead to thermal-hydraulic instabilities through blockage of tube supports, and create occluded regions where corrosive species can concentrate along tubes and in tube-to-tube support plate crevices. The performance of the SGs is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. Currently, there are two strategies employed by utilities for minimizing deposit formation on steam generator internal surfaces. The first is to minimize the source term, i.e., reduce the amount of corrosion products in the feedwater. Two methods are commonly used to accomplish this goal: chemistry optimization and plant modifications. The first method uses alternate amines to control the at-temperature pH (pH T ) in specific locations of the secondary system, thereby minimizing the corrosion of balance of plant (BOP) metals. The second method requires removal of metals from the secondary system that are a significant source of corrosion products (e.g., replace 90/10 Cu/Ni condenser tubes with titanium). The second strategy for lowering deposit loads utilizes chemical or mechanical means to remove existing deposits from the SGs (e.g., chemical cleaning or sludge lancing). Many utilities have opted for a combination of these two strategies. A third potential strategy for minimizing deposition of corrosion products on SG internal surfaces is to use online dispersant addition to help prevent the corrosion products from adhering to the steam generator surfaces. By inhibiting the deposition of the corrosion products, the dispersant can facilitate more effective removal from the SGs via blowdown. This type of strategy has been employed at fossil boilers for many decades. However, due to the use of inorganic (sulfur and other impurities) polymerization

  18. Dispersant trial at ANO-2: Qualification for a short-term trial prior to SG replacement

    Energy Technology Data Exchange (ETDEWEB)

    Fruzzetti, K.; Frattini, P. [Electric Power Research Inst., Palo Alto, CA (United States); Robbins, P. [Entergy Operations, Arkansas Nuclear One, Russellville, AR (United States); Miller, A. [Pedro Point Technology, Inc., Pacifica, CA (United States); Varrin, R.; Kreider, M. [Dominion Engineering Inc., McLean, VA (United States)

    2002-07-01

    Corrosion products in the secondary side of pressurized water reactor (PWR) steam generators (SGs) primarily deposit on the SG tubes. These deposits can inhibit heat transfer, lead to thermal-hydraulic instabilities through blockage of tube supports, and create occluded regions where corrosive species can concentrate along tubes and in tube-to-tube support plate crevices. The performance of the SGs is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. Currently, there are two strategies employed by utilities for minimizing deposit formation on steam generator internal surfaces. The first is to minimize the source term, i.e., reduce the amount of corrosion products in the feedwater. Two methods are commonly used to accomplish this goal: chemistry optimization and plant modifications. The first method uses alternate amines to control the at-temperature pH (pH{sub T}) in specific locations of the secondary system, thereby minimizing the corrosion of balance of plant (BOP) metals. The second method requires removal of metals from the secondary system that are a significant source of corrosion products (e.g., replace 90/10 Cu/Ni condenser tubes with titanium). The second strategy for lowering deposit loads utilizes chemical or mechanical means to remove existing deposits from the SGs (e.g., chemical cleaning or sludge lancing). Many utilities have opted for a combination of these two strategies. A third potential strategy for minimizing deposition of corrosion products on SG internal surfaces is to use online dispersant addition to help prevent the corrosion products from adhering to the steam generator surfaces. By inhibiting the deposition of the corrosion products, the dispersant can facilitate more effective removal from the SGs via blowdown. This type of strategy has been employed at fossil boilers for many decades. However, due to the use of inorganic (sulfur and other impurities) polymerization

  19. Vacuum Outer-Gap Structure in Pulsar Outer Magnetospheres

    International Nuclear Information System (INIS)

    Gui-Fang, Lin; Li, Zhang

    2009-01-01

    We study the vacuum outer-gap structure in the outer magnetosphere of rotation-powered pulsars by considering the limit of trans-field height through a pair production process. In this case, the trans-field height is limited by the photon-photon pair production process and the outer boundary of the outer gap can be extended outside the light cylinder. By solving self-consistently the Poisson equation for electrical potential and the Boltzmann equations of electrons/positrons and γ-rays in a vacuum outer gap for the parameters of Vela pulsar, we obtain an approximate geometry of the outer gap, i.e. the trans-field height is limited by the pair-production process and increases with the radial distance to the star and the width of the outer gap starts at the inner boundary (near the null charge surface) and ends at the outer boundary which locates inside or outside the light cylinder depending on the inclination angle. (geophysics, astronomy, and astrophysics)

  20. Fuel assembly and fuel cladding tube

    International Nuclear Information System (INIS)

    Tsutsumi, Shinro; Ito, Ken-ichi; Inagaki, Masatoshi; Nakajima, Junjiro.

    1996-01-01

    A fuel cladding tube is a zirconium liner tube formed by lining a pure zirconium layer on the inner side of a zirconium alloy tube. The fuel cladding tube is formed by extrusion molding of a composite billet formed by inserting a pure zirconium billet into a zirconium alloy billet. Accordingly, the pure zirconium layer and the zirconium alloy tube are strongly joined by metal bond. The fuel cladding tube has an external oxide film on the outer surface of the zirconium alloy tube and an internal oxide film on the inner side of the pure zirconium layer. The external oxide film has a thickness preferably of about 1μm. The internal oxide film has a thickness of not more than 10μm, preferably, from 1 to 5μm. With such a constitution, flaws to be formed on both inner and outer surfaces of the cladding tube upon assembling a fuel assembly can be reduced thereby enabling to reduce the amount of hydrogen absorbed to the cladding tube. (I.N.)

  1. Spring/dimple instrument tube restraint

    International Nuclear Information System (INIS)

    DeMario, E.E.; Lawson, C.N.

    1993-01-01

    A nuclear fuel assembly for a pressurized water nuclear reactor has a spring and dimple structure formed in a non-radioactive insert tube placed in the top of a sensor receiving instrumentation tube thimble disposed in the fuel assembly and attached at a top nozzle, a bottom nozzle, and intermediate grids. The instrumentation tube thimble is open at the top, where the sensor or its connection extends through the cooling water for coupling to a sensor signal processor. The spring and dimple insert tube is mounted within the instrumentation tube thimble and extends downwardly adjacent the top. The springs and dimples restrain the sensor and its connections against lateral displacement causing impact with the instrumentation tube thimble due to the strong axial flow of cooling water. The instrumentation tube has a stainless steel outer sleeve and a zirconium alloy inner sleeve below the insert tube adjacent the top. The insert tube is relatively non-radioactivated inconel alloy. The opposed springs and dimples are formed on diametrically opposite inner walls of the insert tube, the springs being formed as spaced axial cuts in the insert tube, with a web of the insert tube between the cuts bowed radially inwardly for forming the spring, and the dimples being formed as radially inward protrusions opposed to the springs. 7 figures

  2. Porphyromonas gingivalis Outer Membrane Vesicles Mediate Coaggregation and Piggybacking of Treponema denticola and Lachnoanaerobaculum saburreum

    Directory of Open Access Journals (Sweden)

    Daniel Grenier

    2013-01-01

    Full Text Available Porphyromonas gingivalis sheds outer membrane vesicles that contain several virulence factors, including adhesins. In this study, we investigated the ability of P. gingivalis outer membrane vesicles to mediate the coaggregation and piggybacking of Treponema denticola and Lachnoanaerobaculum saburreum. Marked coaggregation between T. denticola and L. saburreum occurred in the presence of P. gingivalis outer membrane vesicles. Sucrose was an effective chemoattractant for the motile species T. denticola. The addition of outer membrane vesicles to a mixture of T. denticola and L. saburreum significantly increased the number of nonmotile bacteria that migrated into a sucrose-filled capillary tube immersed in the bacterial mixture. Under optimal conditions, the number of nonmotile L. saburreum in the capillary tube increased approximately 5-fold, whereas no increase occurred when boiled vesicles were used. This study showed that P. gingivalis outer membrane vesicles mediate coaggregation between T. denticola and L. saburreum and that nonmotile bacteria can be translocated by piggybacking on spirochetes.

  3. 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)

  4. Feeding Tubes

    Science.gov (United States)

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

  5. Assessment of corrosion failure in copper tube of refrigerator unit

    International Nuclear Information System (INIS)

    Mohd Harun; Hafizal Yazid; Zaiton Selamat; Mohd Shariff Sattar; Muhamamd Jalil

    2007-01-01

    The copper tubes of the refrigerator unit have been coated with red and white color paints. According to the date of purchase and complaint recorded, the tube leaked after operation about one year. It was observed that the tubes became black and green in color at U-bend of the tube. No corrosion occurred on the internal surface of the tube. The leaking started at outer surface of the tube. The leaking started at outer surface and propagated to the internal surface of the tubes. The leaking damage was caused by corrosive species either from atmospheric corrosion or the paint contained chloride and sulfur elements. The corrosive species of sulfur and chlorine were a main factor in pitting corrosion. (author)

  6. 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.

  7. The fate of the outer plasmasphere

    International Nuclear Information System (INIS)

    Elphic, R.C.; Thomsen, M.F.; Borovsky, J.E.

    1997-01-01

    Both the solar wind and the ionosphere contribute to Earth close-quote s magnetospheric plasma environment. However, it is not widely appreciated that the plasmasphere is a large reservoir of ionospheric ions that can be tapped to populate the plasma sheet. We employ empirical models of high-latitude ionospheric convection and the geomagnetic field to describe the transport of outer plasmasphere flux tubes from the dayside, over the polar cap and into the magnetotail during the early phases of a geomagnetic storm. We calculate that this process can give rise to high densities of cold plasma in the magnetotail lobes and in the near-Earth plasma sheet during times of enhanced geomagnetic activity, and especially during storms. This model can help explain both polar cap ionization patches and the presence of cold flowing ions downtail.copyright 1997 American Geophysical Union

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  9. 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

  10. EDF steam generators fleet: In-operation monitoring of TSP blockage and tube fouling

    Energy Technology Data Exchange (ETDEWEB)

    Bertrand, P.; Gay, N.; Crinon, R. [Electricite De France (France)

    2012-07-01

    EDF operates 58 Pressurized Water Reactors in France. In the mid 2000‟s some of them have been affected by Steam Generators (SG) Tube Support Plates (TSP) blockage and U-tubes external surface fouling with iron oxides deposits due to corrosion of secondary-side components. These issues have been tackled by a global maintenance strategy of chemical cleanings and a method for in-operation monitoring of fouling and TSP blockage has been developed and is implemented since mid 2009. This monitoring is aimed at giving information for SG maintenance planning as regards non destructive examinations and chemical cleaning. This paper will first remind of the physical reasons of fouling and TSP blockage and identify the resulting stakes regarding safety and availability along with the action levers available to control both phenomena. Then details will be given on how in-operation monitoring of fouling and TSP blockage is carried out, using measurements of Wide Range water Level (WRL) and SG steam pressure during thermally stabilized periods. Information will also be given on how those data are analyzed and shared as well at a local as at a corporate level to participate in the planning of SG inspection and maintenance operations. Finally, possible refinements will be discussed, notably regarding the issue of WRL measurements reliability and the possibility to use the analysis of SG dynamic behavior during power transients to assess the TSP blockage ratio. In terms of „issues requiring discussion‟, the following are operational issues currently being investigated by EDF: 1. SG pressure can have quite large variations during one operating cycle (notably after a plant trip) and from one cycle to the other and generally pressure tends to decrease on a long-term basis. How can such variations be explained? What are the solutions to moderate/stop the pressure loss? 2. On some of the SG-models operated by EDF, hard curative Chemical Cleaning of the U-tubes didn't bring

  11. Turbine airfoil with outer wall thickness indicators

    Science.gov (United States)

    Marra, John J; James, Allister W; Merrill, Gary B

    2013-08-06

    A turbine airfoil usable in a turbine engine and including a depth indicator for determining outer wall blade thickness. The airfoil may include an outer wall having a plurality of grooves in the outer surface of the outer wall. The grooves may have a depth that represents a desired outer surface and wall thickness of the outer wall. The material forming an outer surface of the outer wall may be removed to be flush with an innermost point in each groove, thereby reducing the wall thickness and increasing efficiency. The plurality of grooves may be positioned in a radially outer region of the airfoil proximate to the tip.

  12. [In silico CRISPR-based sgRNA design].

    Science.gov (United States)

    Wang, Yuanli; Chuai, Guohui; Yan, Jifang; Shi, Lei; Liu, Qi

    2017-10-25

    CRISPR-based genome editing has been widely implemented in various cell types. In-silico single guide RNA (sgRNA) design is a key step for successful gene editing using CRISPR system. Continuing efforts are made to refine in-silico sgRNA design with high on-target efficacy and reduced off-target effects. In this paper, we summarize the present sgRNA design tools, and show that efficient in-silico models can be built that integrate current heterogeneous genome-editing data to derive unbiased sgRNA design rules and identify key features for improving sgRNA design. Our review shows that systematic comparisons and evaluation of on-target and off-target effects of sgRNA will allow more precise genome editing and gene therapies using the CRISPR system.

  13. Properties and application study of Inconel alloy tube made in China

    International Nuclear Information System (INIS)

    Yang Xiang; Su Xingwan; Wen Yan

    1997-01-01

    The mech-physical properties and the corrosion resistance properties of the SG tube of Inconel alloy made in China under any conditions are briefly presented, and the test and research for bending and expending the tubes have been performed. In the process of corrosion experiments the Inconel alloy tubes were compared with that of the same kind of materials made in foreign countries. The Inconel alloy tubes have better stress corrosion resistance cracking prosperities than Inconel 600 and Incoloy 800 when they were in the solutions which contained high concentrated chlorine ion and alkali at high temperature

  14. 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)

  15. 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

  16. Recent progress in SG level control in French PWR plants

    International Nuclear Information System (INIS)

    Parry, A.; Petetrot, J.F.; Vivier, M.J.

    1985-10-01

    Controlling the steam generator (SG) level is of major importance in a large PWR plant. This has led to extensive work on SG computer models. This paper presents results of the comparison between calculations and tests on the first four-loop plant in France. Four-loop plants started up after 1985 will be equipped with digital instead of analog controllers. A new SG level control has been designed and then optimised using the validated SG model. A prototype of this new system has been successfully tested on a three-loop plant. 4 refs

  17. 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.

  18. 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)

  19. Automatic integrated testing bench for tubes in translation

    International Nuclear Information System (INIS)

    Dufayet, J.P.; Perdijon, J.

    1976-01-01

    All the nondestructive tests required for receiving the cladding tubes intended for fast nuclear reactor are integrated on this bench: quality control by eddy currents and ultra-sounds, thickness and (inner and outer) diameter measurement. The linear displacement of the tube allows very high rates to be attained [fr

  20. Circumferential buckling instability of a growing cylindrical tube

    KAUST Repository

    Moulton, D.E.; Goriely, A.

    2011-01-01

    A cylindrical elastic tube under uniform radial external pressure will buckle circumferentially to a non-circular cross-section at a critical pressure. The buckling represents an instability of the inner or outer edge of the tube. This is a common

  1. A pulsed eddy current probe for inspection of support plates from within Alloy-800 steam generator tubes

    International Nuclear Information System (INIS)

    Krause, T. W.; Babbar, V. K.; Underhill, P. R.

    2014-01-01

    Support plate degradation and fouling in nuclear steam generators (SGs) can lead to SG tube corrosion and loss of efficiency. Inspection and monitoring of these conditions can be integrated with preventive maintenance programs, thereby advancing station-life management processes. A prototype pulsed eddy current (PEC) probe, targeting inspection issues associated with SG tubes in SS410 tube support plate structures, has been developed using commercial finite element (FE) software. FE modeling was used to identify appropriate driver and pickup coil configurations for optimum sensitivity to changes in gap and offset for Alloy-800 SG tubes passing through 25 mm thick SS410 support plates. Experimental measurements using a probe that was manufactured based on the modeled configuration, were used to confirm the sensitivity of differential PEC signals to changes in relative position of the tube within the tube support plate holes. Models investigated the effect of shift and tilt of tube with respect to hole centers. Near hole centers and for small shifts, modeled signal amplitudes from the differentially connected coil pairs were observed to change linearly with tube shift. This was in agreement with experimentally measured TEC coil response. The work paves the way for development of a system targeting the inspection and evaluation of support plate structures in steam generators

  2. A pulsed eddy current probe for inspection of support plates from within Alloy-800 steam generator tubes

    Science.gov (United States)

    Krause, T. W.; Babbar, V. K.; Underhill, P. R.

    2014-02-01

    Support plate degradation and fouling in nuclear steam generators (SGs) can lead to SG tube corrosion and loss of efficiency. Inspection and monitoring of these conditions can be integrated with preventive maintenance programs, thereby advancing station-life management processes. A prototype pulsed eddy current (PEC) probe, targeting inspection issues associated with SG tubes in SS410 tube support plate structures, has been developed using commercial finite element (FE) software. FE modeling was used to identify appropriate driver and pickup coil configurations for optimum sensitivity to changes in gap and offset for Alloy-800 SG tubes passing through 25 mm thick SS410 support plates. Experimental measurements using a probe that was manufactured based on the modeled configuration, were used to confirm the sensitivity of differential PEC signals to changes in relative position of the tube within the tube support plate holes. Models investigated the effect of shift and tilt of tube with respect to hole centers. Near hole centers and for small shifts, modeled signal amplitudes from the differentially connected coil pairs were observed to change linearly with tube shift. This was in agreement with experimentally measured TEC coil response. The work paves the way for development of a system targeting the inspection and evaluation of support plate structures in steam generators.

  3. Ear Tubes

    Science.gov (United States)

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

  4. 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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  7. Development of remote field ECT sensor for high temperature steam generator tubes

    International Nuclear Information System (INIS)

    Onoue, Akira; Yamada, Fumiaki; Imai, Yoshiyuki; Watanabe, Tomoo; Ozawa, Kazumasa

    2005-02-01

    Commercialized Fast Breeder Reactor (FBR)s have to achieve competitive unit price in electricity generation with other energy sources by reducing not only construction and fuel cost but also operation and maintenance cost, in order to be introduced in line with market principles. Operation and maintenance cost cannot be reduced until plant utilization factor is enhanced by shortening duration times of periodical inspections and expanding continuous operation periods. Critical paths in periodical inspections should be shortened to reduce entire duration time of a periodical inspections should be shortened to reduce entire duration time of a periodical inspection. and reduction of the inspection time is desired. Reflecting this background, as a research activity within the Feasibility Study for Future Commercialized FBRs, technology for volumetric inspection of SG heat transfer tubes in high temperature is being developed, in order to reduce the inspection time by skipping cooling down process. This report describes a series of experiments of heat-resistant remote field (RF) ECT probe to evaluate its defect detection performance on outer surface of heat transfer tubes. The results are summarized as listed below: (1) Defects can be detected in high temperature if sodium is drained, but cannot be detected if tube is submerged in liquid sodium. (2) The goal detection performance against round wall thinning is thought to be possibly achieved, because the measured S/N ratio exceeds 9.3 in detecting artificial round wall thinning with 10 mm width and depth beyond 10%. (3) Round wall slits can possibly detected because the S/N ratio exceeded 3.7 in detecting artificial round wall slits with 0.5 mm width and depth beyond 15%. (4) Defects of partial wall thinning are difficult to be detected, because the S/N ratio was less than 2.0 in detecting partial wall thinning with 10 mm width and 10 mm axial length and depth up to 20%. (5) In detecting defects of 12Cr steel tubes by

  8. 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

  9. The CMS Outer Hadron Calorimeter

    CERN Document Server

    Acharya, Bannaje Sripathi; Banerjee, Sunanda; Banerjee, Sudeshna; Bawa, Harinder Singh; Beri, Suman Bala; Bhandari, Virender; Bhatnagar, Vipin; Chendvankar, Sanjay; Deshpande, Pandurang Vishnu; Dugad, Shashikant; Ganguli, Som N; Guchait, Monoranjan; Gurtu, Atul; Kalmani, Suresh Devendrappa; Kaur, Manjit; Kohli, Jatinder Mohan; Krishnaswamy, Marthi Ramaswamy; Kumar, Arun; Maity, Manas; Majumder, Gobinda; Mazumdar, Kajari; Mondal, Naba Kumar; Nagaraj, P; Narasimham, Vemuri Syamala; Patil, Mandakini Ravindra; Reddy, L V; Satyanarayana, B; Sharma, Seema; Singh, B; Singh, Jas Bir; Sudhakar, Katta; Tonwar, Suresh C; Verma, Piyush

    2006-01-01

    The CMS hadron calorimeter is a sampling calorimeter with brass absorber and plastic scintillator tiles with wavelength shifting fibres for carrying the light to the readout device. The barrel hadron calorimeter is complemented with a outer calorimeter to ensure high energy shower containment in CMS and thus working as a tail catcher. Fabrication, testing and calibrations of the outer hadron calorimeter are carried out keeping in mind its importance in the energy measurement of jets in view of linearity and resolution. It will provide a net improvement in missing $\\et$ measurements at LHC energies. The outer hadron calorimeter has a very good signal to background ratio even for a minimum ionising particle and can hence be used in coincidence with the Resistive Plate Chambers of the CMS detector for the muon trigger.

  10. Origin of Outer Solar System

    Science.gov (United States)

    Holman, Matthew J.; Lindstrom, David (Technical Monitor)

    2005-01-01

    Our ongoing research program combines extensive deep and wide-field observations using a variety of observational platforms with numerical studies of the dynamics of small bodies in the outer solar system in order to advance the main scientific goals of the community studying the Kuiper belt and the outer solar system. These include: (1) determining the relative populations of the known classes of KBOs as well as other possible classes; ( 2 ) determining the size distributions or luminosity function of the individual populations or the Kuiper belt as a whole; (3) determining the inclinations distributions of these populations; (4) establishing the radial extent of the Kuiper belt; ( 5 ) measuring and relating the physical properties of different types of KBOs to those of other solar system bodies; and, (6) completing our systematic inventory of the satellites of the outer planets.

  11. Electron tube

    Science.gov (United States)

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

    2011-12-20

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

  12. Nuclear fuel grid outer strap

    International Nuclear Information System (INIS)

    Duncan, R.; Craver, J.E.

    1989-01-01

    This patent describes a nuclear reactor fuel assembly grid. It comprises a first outer grip strap segment end. The first end having a first tab arranged in substantially the same plane as the plane defined by the first end; a second outer grip strap end. The second end having a second slot arranged in substantially the same plane as the plane defined by the second end, with the tab being substantially disposed in the slot, defining a socket therebetween; and a fort tine interposed substantially perpendicularly in the socket

  13. Stochastic modeling of inspection uncertainties and applications to pitting flaws in steam generator tubes

    International Nuclear Information System (INIS)

    Mao, D.; Yuan, X.-X.; Pandey, M.D.

    2009-01-01

    Steam generators (SG) are a major pressure retaining component of great safety significance in nuclear power plants. Due to various manufacturing, operation and maintenance activities, as well as material interaction with the surrounding chemical environment, the SG tubes have been subject to a number of degradation modes. Among them, the under-deposit pitting corrosion at outside surfaces of the SG tubes just on top of the tubesheet support plates has had a serious impact on the integrity of the SG tubes. This paper presents an advanced probabilistic model of pitting corrosion characterizing the inherent randomness of the pitting process and measurement uncertainties of the in-service inspection (ISI) data obtained from eddy current (EC) inspections. A Bayesian method based on Markov Chain Monte Carlo (MCMC) simulation is developed for estimating the model parameters. The proposed model is able to predict the actual pit number, the actual pit depth as well as the maximum pit depth, which is the main interest of the pitting corrosion model. (author)

  14. Chest tube insertion

    Science.gov (United States)

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

  15. A novel investigation of heat transfer characteristics in rifled tubes

    Science.gov (United States)

    Jegan, C. Dhayananth; Azhagesan, N.

    2018-05-01

    The experimental investigation of heat transfer of water flowing in a rifled tube was explored at different pressures and at various operating conditions in a rifled tube heat exchanger. The specifications for the inner and outer diameters of the inner tube are 25.8 and 50.6 mm, respectively. The working fluids used in shell side and tube side are cold and hot water. The rifled tube was made of the stainless steel with 4 ribs, 50.6 mm outer diameter, 0.775 mm rib height, 58o helix angle and the length 1500 mm. The effect of pressure, wall heat flux and friction factor were discussed. The results confirm that even at low pressures the rifled tubes has an obvious enhancement in heat transfer compared with smooth tube. Results depicts that the Nusselt number increases with Reynolds number and the friction factor decreases with increase in Reynolds number and the heat transfer rate is higher for the rifled tube when compared to smooth tube, because of strong swirl flow due to centrifugal action. It also confirms that, the friction factor obtained from the rifled tube is significantly higher than that of smooth tube.

  16. 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)

  17. Getting Sloshed in Outer Space

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 12; Issue 12. Getting Sloshed in Outer Space - Liquid Behavior in Microgravity. N Ananthkrishnan. General Article Volume 12 Issue 12 December 2007 pp 40-45. Fulltext. Click here to view fulltext PDF. Permanent link:

  18. Outer space structure and development

    International Nuclear Information System (INIS)

    Zeldovich, J.; Novikov, I.

    1975-01-01

    A brief account is presented answering the question of what in fact the outer space we observe consists of. The principle of spatial homogeneity of the universe and the idea of non-stationary cosmology are discussed. The origin and the future development of the universe are explained using the two above mentioned and some other hypotheses. (J.K.)

  19. Outer space structure and development

    Energy Technology Data Exchange (ETDEWEB)

    Zeldovich, J; Novikov, I

    1975-10-01

    A brief account is presented answering the question of what in fact the outer space we observe consists of. The principle of spatial homogeneity of the universe and the idea of non-stationary cosmology are discussed. The origin and the future development of the universe are explained using the two above mentioned and some other hypotheses.

  20. Plasmas in the outer heliosphere

    Science.gov (United States)

    Belcher, J. W.; Richardson, J. D.; Lazarus, A. J.; Gazis, P. R.; Barnes, A.

    1995-01-01

    We review the observed properties of the solar wind in the outer heliosphere, including observations from Voyager and the Pioneers, as well as from inner heliospheric probes as appropriate. These observations are crucial to modeling of the heliosphere and its interactions with the interstellar medium, since the wind ram pressure and its temporal variations are important in understanding the distance to the termination shock and heliopause and how those boundaries might vary in time. We focus on results since Solar Wind 7. Among the issues we will discuss are: (1) the time scales for and statistical properties of variations in the ram pressure in the outer heliosphere, and how those variations might affect the morphology of the heliospheric/interstellar medium interface; (2) the question of possible solar wind slowing in the outer heliosphere due to the pick-up of interstellar ions; (3) the issue of whether there is bulk heating of the solar wind associated either with interstellar ion pick-up or with continued heating due to stream-stream interactions; (4) evidence for latitudinal variations in solar wind properties; and (5) the 1.3 year periodicities apparent in the outer heliosphere, and the close correspondence with similar variations seen with inner heliospheric probes.

  1. 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)

  2. Outer Synchronization of Complex Networks by Impulse

    International Nuclear Information System (INIS)

    Sun Wen; Yan Zizong; Chen Shihua; Lü Jinhu

    2011-01-01

    This paper investigates outer synchronization of complex networks, especially, outer complete synchronization and outer anti-synchronization between the driving network and the response network. Employing the impulsive control method which is uncontinuous, simple, efficient, low-cost and easy to implement in practical applications, we obtain some sufficient conditions of outer complete synchronization and outer anti-synchronization between two complex networks. Numerical simulations demonstrate the effectiveness of the proposed impulsive control scheme. (general)

  3. Reactor Power Meter type SG-8

    Energy Technology Data Exchange (ETDEWEB)

    Glowacki, S W

    1981-01-01

    The report describes the principle and electronic circuits of the Reactor Power Meter type SG-8. The gamma radiation caused by the activity of the reactor first cooling circuit affectes the ionization chamber being the detector of the instrument. The output detector signal direct current is converted into the frequency of electric pulses by means of the current-to-frequency converter. The output converter frequency is measured by the digital frequency meter: the number of measured digits in time unit is proportional to the reactor power.

  4. 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.

  5. Active acoustic leak detection for LMFBR steam generator. Pt. 5. Experiment for detection of bubbles using the SG full sector model

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi

    1997-01-01

    In order to prevent the expansion of tube damages and to maintain structural safety in steam generators (SG) of fast breeder reactors (FBR), it is necessary to detect precisely and immediately the leakage of water from tubes of heat exchangers. Therefore, an active acoustic method, which detects the sound attenuation due to bubbles generated in the sodium-water reactions, it being developed. In this paper, the attenuation characteristics of sound attenuated by bubbles and influence of background noise are investigated experimentally by using an SG full sector model (diameter ratio about 1/1, height ratio about 1/7) simulating the actual SG. As an experimental result, the received sound attenuation for ten seconds was more than 10 dB from air bubble injection when injected bubble of 10 l/s (equivalence water leak rate about 10 g/s). The attenuation of sound are least affected by bubble injection position of heat exchanger tube bunch department. And the time was about 25 seconds till the sound attenuation became 10 dB in case of quantity of air bubble 1 l/s (equivalent water leak rate about 1 g/s). It is clarified that the background noise hardly influenced water leak detection performance as a result of having examined influence of background noise. (author)

  6. 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.

  7. 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

  8. 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

  9. photomultiplier tubes

    CERN Multimedia

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

  10. photomultiplier tube

    CERN Multimedia

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

  11. Control rod guide tube assembly

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1982-01-01

    An improved fuel assembly is described as consisting of a sleeve that engages one end of a control rod guide tube essentially fixing the guide tube to one of the fuel assembly end structures. The end of the sleeve protrudes above the surface of the end fitting. The outer surface of the sleeve has a peripheral groove that engages the resilient sides of a cellular grid or lattice shaped lock. This lock fixes the sleeve in position between the various elements that comprise the end fitting, thereby eliminating a profusion of costly and potentially troublesome nuts, threaded studs and the like that are frequently employed in the fuel assemblies that are presently in use

  12. 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

  13. 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

  14. Detailed Performance of the Outer Tracker at LHCb

    CERN Document Server

    Tuning, N

    2014-01-01

    The LHCb Outer Tracker is a gaseous detector covering an area of 5x6m2 with 12 double layers of straw tubes. Based on data of the first LHC running period from 2010 to 2012, the performance in terms of the single hit resolution and efficiency are presented. Details on the ionization length and subtle effects regarding signal reflections and the subsequent time-walk correction are given. The efficiency to detect a hit in the central half of the straw is estimated to be 99.2%, and the position resolution is determined to be approximately 200 um, depending on the detailed implementation of the internal alignment of individual detector modules. The Outer Tracker received a dose in the hottest region corresponding to 0.12 C/cm, and no signs of gain deterioration or other ageing effects are observed.

  15. SG39 Deliverables. Comments on Covariance Data

    International Nuclear Information System (INIS)

    Yokoyama, Kenji

    2015-01-01

    The covariance matrix of a scattered data set, x_i (i=1,n), must be symmetric and positive-definite. As one of WPEC/SG39 contributions to the SG40/CIELO project, several comments or recommendations on the covariance data are described here from the viewpoint of nuclear-data users. To make the comments concrete and useful for nuclear-data evaluators, the covariance data of the latest evaluated nuclear data library, JENDL-4.0 and ENDF/B-VII.1 are treated here as the representative materials. The surveyed nuclides are five isotopes that are most important for fast reactor application. The nuclides, reactions and energy regions dealt with are followings: Pu-239: fission (2.5∼10 keV) and capture (2.5∼10 keV), U-235: fission (500 eV∼10 keV) and capture (500 eV∼30 keV), U-238: fission (1∼10 MeV), capture (below 20 keV, 20∼150 keV), inelastic (above 100 keV) and elastic (above 20 keV), Fe-56: elastic (below 850 keV) and average scattering cosine (above 10 keV), and, Na-23: capture (600 eV∼600 keV), inelastic (above 1 MeV) and elastic (around 2 keV)

  16. Development of Probability Evaluation Methodology for High Pressure/Temperature Gas Induced RCS Boundary Failure and SG Creep Rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Byung Chul; Hong, Soon Joon; Lee, Jin Yong; Lee, Kyung Jin; Lee, Kuh Hyung [FNC Tech. Co., Seoul (Korea, Republic of)

    2008-04-15

    Existing MELCOR 1.8.5 model was improved in view of severe accident natural circulation and MELCOR 1.8.6 input model was developed and calculation sheets for detailed MELCOR 1.8.6 model were produced. Effects of natural circulation modeling were found by simulating SBO accident by comparing existing model with detailed model. Major phenomenon and system operations which affect on natural circulation by high temperature and high pressure gas were investigated and representative accident sequences for creep rupture model of RCS pipeline and SG tube were selected.

  17. Turbine airfoil with a compliant outer wall

    Science.gov (United States)

    Campbell, Christian X [Oviedo, FL; Morrison, Jay A [Oviedo, FL

    2012-04-03

    A turbine airfoil usable in a turbine engine with a cooling system and a compliant dual wall configuration configured to enable thermal expansion between inner and outer layers while eliminating stress formation in the outer layer is disclosed. The compliant dual wall configuration may be formed a dual wall formed from inner and outer layers separated by a support structure. The outer layer may be a compliant layer configured such that the outer layer may thermally expand and thereby reduce the stress within the outer layer. The outer layer may be formed from a nonplanar surface configured to thermally expand. In another embodiment, the outer layer may be planar and include a plurality of slots enabling unrestricted thermal expansion in a direction aligned with the outer layer.

  18. Advanced evaluation method of SG TSP BEC hole blockage rate

    International Nuclear Information System (INIS)

    Izumida, Hiroyuki; Nagata, Yasuyuki; Harada, Yutaka; Murakami, Ryuji

    2003-01-01

    In spite of the control of the water chemistry of SG secondary feed-water in PWR-SG, SG TSP BEC holes, which are the flow path of secondary water, are often clogged. In the past, the trending of BEC hole blockage rate has conducted by evaluating ECT original signals and visual inspections. However, the ECT original signals of deposits are diversified, it becomes difficult to analyze them with the existing evaluation method using the ECT original signals. In this regard, we have developed the secondary side visual inspection system, which enables the high-accuracy evaluation of BEC hole blockage rate, and new ECT signal evaluation method. (author)

  19. 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

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

    International Nuclear Information System (INIS)

    Joo, Kyung Mun; Hong, Jun Hee

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-15

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

  2. 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

  3. In-service examination of IHX tubing with eddy current NDT equipment

    International Nuclear Information System (INIS)

    Brown, R.L.

    1972-01-01

    Single and multiple frequency eddy current (ET) nondestructive testing (NDT) techniques and equipment were investigated for in-service inspection of sodium-contaminated intermediate heat exchanger (IHX) tubing. A four frequency technique, demonstrated in the laboratory, was relatively insensitive to signals caused by probe motion, tube support plates, and residual sodium on the outer surface of the tubes. No method was found to avoid the signals from residual sodium on the inside surfaces of the tube. (U.S.)

  4. Mitochondrial dysfunction underlying outer retinal diseases

    DEFF Research Database (Denmark)

    Lefevere, Evy; Toft-Kehler, Anne Katrine; Vohra, Rupali

    2017-01-01

    Dysfunction of photoreceptors, retinal pigment epithelium (RPE) or both contribute to the initiation and progression of several outer retinal disorders. Disrupted Müller glia function might additionally subsidize to these diseases. Mitochondrial malfunctioning is importantly associated with outer...

  5. Dismantling Experiment of Mock-up Tube Bundle of Steam Generator

    International Nuclear Information System (INIS)

    Kim, Sung Kyun; Lee, Kune Woo

    2010-01-01

    A SG (steam generator) is one of the biggest decommissioning components in nuclear power plants and one has been replaced 2∼6 times during the whole operation of a nuclear power plant. The old SG should be decommissioned for the purpose of the volume reduction of radioactive waste. Among the components of SG, the tube bundle is one of the most difficult items to be dismantled due to the fact that it is very hard to cut since it is made of Inconel 600 which has high resistance of corrosion and abrasion. Moreover, All cutting process should be performed by remotely since radioactive contamination of the internal surface of SG tubes is very high (about 150,000∼300,000 Bq/cm 2 ). Therefore, it is necessary to choose the appropriate cutting methods by the pros and cons analysis for candidate dismantling technologies and to do experiment study for the validation. In this study, the results of cutting experiment for a mock-up bundle by using band saw cutting method are described herein

  6. Eustachian tube patency

    Science.gov (United States)

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

  7. Feeding tube - infants

    Science.gov (United States)

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

  8. WORKSHOP: Inner space - outer space

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    During the first week of May, the Fermilab theoretical astrophysics group hosted an international conference on science at the interface of particle physics and cosmology/astrophysics. The conference (Inner Space-Outer Space) was attended by a very diverse group of more than 200 physical scientists, including astronomers, astrophysicists, cosmologists, low-temperature physicists, and elementary particle theorists and experimentalists. The common interest which brought this diverse group to gether is the connection between physics on the smallest scale probed by man - the realm of elementary particle physics - and physics on the largest scale imaginable (the entire Universe) - the realm of cosmology

  9. WORKSHOP: Inner space - outer space

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1984-09-15

    During the first week of May, the Fermilab theoretical astrophysics group hosted an international conference on science at the interface of particle physics and cosmology/astrophysics. The conference (Inner Space-Outer Space) was attended by a very diverse group of more than 200 physical scientists, including astronomers, astrophysicists, cosmologists, low-temperature physicists, and elementary particle theorists and experimentalists. The common interest which brought this diverse group to gether is the connection between physics on the smallest scale probed by man - the realm of elementary particle physics - and physics on the largest scale imaginable (the entire Universe) - the realm of cosmology.

  10. Outer scale of atmospheric turbulence

    Science.gov (United States)

    Lukin, Vladimir P.

    2005-10-01

    In the early 70's, the scientists in Italy (A.Consortini, M.Bertolotti, L.Ronchi), USA (R.Buser, Ochs, S.Clifford) and USSR (V.Pokasov, V.Lukin) almost simultaneously discovered the phenomenon of deviation from the power law and the effect of saturation for the structure phase function. During a period of 35 years we have performed successively the investigations of the effect of low-frequency spectral range of atmospheric turbulence on the optical characteristics. The influence of the turbulence models as well as a outer scale of turbulence on the characteristics of telescopes and systems of laser beam formations has been determined too.

  11. Tube holding system

    International Nuclear Information System (INIS)

    Cunningham, R.C.

    1978-01-01

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

  12. Quantitative experiments on thermal hydraulic characteristics of an annular tube with twisted fins

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Dairaku, Masayuki; Taniguchi, Masaki; Sato, Kazuyoshi; Suzuki, Satoshi; Akiba, Masato

    2003-11-01

    Thermal hydraulic experiments measuring critical heat flux (CHF) and pressure drop of an annular tube with twisted fins, ''annular swirl tube'', has been performed to examine its applicability to the ITER divertor cooling structure. The annular swirl tube consists of two concentric circular tubes, the outer and inner tubes. The outer tube with outer and inner diameters (OD and ID) of 21 mm and 15 mm is made of Cu-alloy that is CuCrZr and oe of candidate materials of the ITER divertor cooling tube. The inner tube with OD of 11 mm and ID of 9 mm is made of stainless steal. It has an external swirl fin with twist ratio (y) of three to enhance its heat transfer performance. In this tube, cooling water flows inside of the inner tube first, and then returns into an annulus between the outer and inner tubes with a swirl flow at an end-return of the cooling tube. The CHF experiments show that no degradation of CHF of the annular swirl tube in comparison with the conventional swirl tube whose dimensions are similar to those of the outer tube of the annular swirl tube. A minimum axial velocity of 7.1 m/s is required to remove the incident heat flux of 28MW/m 2 , the ITER design value. Applicability of the JAERI's correlation for the heat transfer to the annular swirl tube is also demonstrated by comparing the experimental results with those of the numerical analysis. The friction factor correlation for the annular flow with the twisted fins is also proposed for the hydrodynamic design of the ITER vertical target. The least pressure drop at the end-return is obtained by using the hemispherical end-plug. Its radius is the same as that of ID of the outer cooling tube. These results show that thermal-hydraulic performance of the annular swirl tube is promising in application to the cooling structure for the ITER vertical target. (author)

  13. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    International Nuclear Information System (INIS)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong

    2012-01-01

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced

  14. Manufacturing process for the metal ceramic hybrid fuel cladding tube

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Yang Il; Kim, Sun Han; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    For application in LWRs with suppressed hydrogen release, a metal-ceramic hybrid cladding tube has been proposed. The cladding consists of an inner zirconium tube and outer SiC fiber matrix SiC ceramic composite. The inner zirconium allows the matrix to remain fully sealed even if the ceramic matrix cracks through. The outer SiC composite can increase the safety margin by taking the merits of the SiC itself. However, it is a challenging task to fabricate the metal-ceramic hybrid tube. Processes such as filament winding, matrix impregnation, and surface costing are additionally required for the existing Zr based fuel cladding tubes. In the current paper, the development of the manufacturing process will be introduced.

  15. Can we simulate the development of ODSCC defects in steam generator tubes?

    International Nuclear Information System (INIS)

    Cizelj, L.; Kovac, M.; Dvorsek, T.

    1999-01-01

    The qualitative and quantitative aspects of degradation mechanism causing early retirement of SG tubing are not yet explained to the level allowing for accurate predictions of future behavior. On the other hand, a large amount of data related to tube degradation, inspection, repair, and plant operation have been collected during recent years. It allows for reasonably accurate quantitative predictions, based on statistical analysis of past events and assumption of reasonably constant operating conditions. A computational algorithm was developed to simulate life cycle of ODSCC defects: initiation, growth, measurement, and repair. The main feature of the algorithm is the possibility to address some important changes in the operating parameters, especially those related to the conditions during the plant shutdown. The algorithm can be used to get better insight into the background of SG aging and to predict the future populations of defects, as shown in a realistic numerical example.(author)

  16. Thermal optimization of primary side in double-tube OTSG

    International Nuclear Information System (INIS)

    Wei Xinyu; Dai Chunhui; Hou Suxia; Tai Yun; Zhao Fuyu

    2011-01-01

    Once-through steam generator (OTSG) is usually used in the integrated nuclear power plants which require smaller volume and better effect of heat transfer. The double-tube OTSG component which is composed of straight tube outside and helical tube inside is presented in this paper. The primary fluid is divided into two parts, one is in the inner tube and the other is in the gap among outer tubes. The flow distribution ratio of the primary fluid obviously affects the heat transfer. Thus, the problem of optimization emerges, i.e. how to find an optimal flow distribution ratio with a maximum heat exchange. Analyzed the effects of the distribution ratio on heat transfer, the optimal distribution ratio is obtained by the constrained nonlinear optimization method. Subsequently, the optimal distribution ratio is achieved by a throttling set in the entrance of the inner tube. The result is in substantial agreement with the literature. (author)

  17. Circumferential buckling instability of a growing cylindrical tube

    KAUST Repository

    Moulton, D.E.

    2011-03-01

    A cylindrical elastic tube under uniform radial external pressure will buckle circumferentially to a non-circular cross-section at a critical pressure. The buckling represents an instability of the inner or outer edge of the tube. This is a common phenomenon in biological tissues, where it is referred to as mucosal folding. Here, we investigate this buckling instability in a growing elastic tube. A change in thickness due to growth can have a dramatic impact on circumferential buckling, both in the critical pressure and the buckling pattern. We consider both single- and bi-layer tubes and multiple boundary conditions. We highlight the competition between geometric effects, i.e. the change in tube dimensions, and mechanical effects, i.e. the effect of residual stress, due to differential growth. This competition can lead to non-intuitive results, such as a tube growing to be thinner and yet buckle at a higher pressure. © 2011 Elsevier Ltd. All rights reserved.

  18. Tracking chamber made of 15-mm mylar drift tubes

    Science.gov (United States)

    Kozhin, A.; Borisov, A.; Bozhko, N.; Fakhrutdinov, R.; Plotnikov, I.

    2017-05-01

    We are presenting a drift chamber composed from three layers of mylar drift tubes with outer diameter 15 mm. The pipe is made of strip of mylar film 125 micrometers thick covered with aluminium from the both sides. A strip of mylar is wrapped around the mandrel. Pipe is created by ultrasonic welding. A single drift tube is self-supported structure withstanding 350 g wire tension without supports and internal overpressure. About 400 such tubes were assembled. Design, quality control procedures of the drift tubes are described. Seven chambers were glued from these tubes of 560 mm length. Each chamber consists of 3 layers, 16 tubes per layer. Several chambers were tested with cosmic rays. Results of the tests, counting rate plateau and coordinate resolution are presented.

  19. Tracking chamber made of 15-mm mylar drift tubes

    International Nuclear Information System (INIS)

    Kozhin, A.; Borisov, A.; Bozhko, N.; Fakhrutdinov, R.; Plotnikov, I.

    2017-01-01

    We are presenting a drift chamber composed from three layers of mylar drift tubes with outer diameter 15 mm. The pipe is made of strip of mylar film 125 micrometers thick covered with aluminium from the both sides. A strip of mylar is wrapped around the mandrel. Pipe is created by ultrasonic welding. A single drift tube is self-supported structure withstanding 350 g wire tension without supports and internal overpressure. About 400 such tubes were assembled. Design, quality control procedures of the drift tubes are described. Seven chambers were glued from these tubes of 560 mm length. Each chamber consists of 3 layers, 16 tubes per layer. Several chambers were tested with cosmic rays. Results of the tests, counting rate plateau and coordinate resolution are presented.

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  1. Ultrasonic inspection of steam generator tubes in Superphenix F.B.R. Power plant

    International Nuclear Information System (INIS)

    Gondard, C.

    1991-01-01

    An ultrasonic method has been developed to test of the S.G's tubes of SPX fast breeder reactor. A new type of rotating probes for cracks and wall thickness measurements have been built up and successfully tested. The data acquisition and processing system SPARTACUS was used; it allows high frequency digitalization and powerful signal processings using frequency representations. The actual performances were tested on natural defects under representative operating conditions

  2. WPEC Subgroup Meetings. Joint SG39/SG40-CIELO meeting, 14 May 2014

    International Nuclear Information System (INIS)

    Chadwick, M.; Yokoyama, Kenji; Ishikawa, M.; Archier, P.; Palmiotti, G.; Salvatores, Massimo; Dupont, E.; ); Leal, L.

    2014-05-01

    The aim of WPEC subgroup 39 'Methods and approaches to provide feedback from nuclear and covariance data adjustment for improvement of nuclear data files' is to provide criteria and practical approaches to use effectively the results of sensitivity analyses and cross section adjustments for feedback to evaluators and differential measurement experimentalists in order to improve the knowledge of neutron cross sections, uncertainties, and correlations to be used in a wide range of applications. WPEC subgroup 40-CIELO (Collaborative International Evaluated Library Organization) provides a new working paradigm to facilitate evaluated nuclear reaction data advances. It brings together experts from across the international nuclear reaction data community to identify and document discrepancies among existing evaluated data libraries, measured data, and model calculation interpretations, and aims to make progress in reconciling these discrepancies to create more accurate ENDF-formatted files. SG40-CIELO focusses on 6 important isotopes: "1H, "1"6O, "5"6Fe, "2"3"5","2"3"8U, "2"3"9Pu. This document is the proceedings of the Joint SG39/SG40-CIELO meeting, held at the NEA, Issy-les-Moulineaux, France, on 19-20 May 2015. It comprises all the available presentations (slides) given by the participants: 1 - CIELO Progress (M. Chadwick); 2 - Revised Recommendations from ADJ2010 Adjustment (K. Yokoyama); 3 - Comparisons and Discussions on Adjustment trends from JEFF (CEA) (P. Archier); 4 - Feedback on CIELO Isotopes from ENDF/B-VII.0 Adjustment (G. Palmiotti); 5 - Demonstration - Plot comparisons of participants' results (E. Dupont); 6 - Some very preliminary indications from recent adjustment studies intercomparison (M. Salvatores); 7 - Resonance Region of "5"6Fe for the CIELO Project (L. Leal); 8 - A-priori and a-posteriori covariance data in nuclear cross section adjustments: issues and challenges (G. Palmiotti); 9 - Comments on Covariance Data of JENDL-4.0 and ENDF/B-VII.1 (K

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

    Science.gov (United States)

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

    2015-02-01

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

  4. Bender/Coiler for Tubing

    Science.gov (United States)

    Stoltzfus, J. M.

    1983-01-01

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

  5. Preliminary Study on Biosynthesis of Bacterial Nanocellulose Tubes in a Novel Double-Silicone-Tube Bioreactor for Potential Vascular Prosthesis

    Directory of Open Access Journals (Sweden)

    Feng Hong

    2015-01-01

    Full Text Available Bacterial nanocellulose (BNC has demonstrated a tempting prospect for applications in substitute of small blood vessels. However, present technology is inefficient in production and BNC tubes have a layered structure that may bring danger after implanting. Double oxygen-permeable silicone tubes in different diameters were therefore used as a tube-shape mold and also as oxygenated supports to construct a novel bioreactor for production of the tubular BNC materials. Double cannula technology was used to produce tubular BNC via cultivations with Acetobacter xylinum, and Kombucha, a symbiosis of acetic acid bacteria and yeasts. The results indicated that Kombucha gave higher yield and productivity of BNC than A. xylinum. Bacterial nanocellulose was simultaneously synthesized both on the inner surface of the outer silicone tube and on the outer surface of the inner silicone tube. Finally, the nano BNC fibrils from two directions formed a BNC tube with good structural integrity. Scanning electron microscopy inspection showed that the tubular BNC had a multilayer structure in the beginning but finally it disappeared and an intact BNC tube formed. The mechanical properties of BNC tubes were comparable with the reported value in literatures, demonstrating a great potential in vascular implants or in functional substitutes in biomedicine.

  6. Preliminary Study on Biosynthesis of Bacterial Nanocellulose Tubes in a Novel Double-Silicone-Tube Bioreactor for Potential Vascular Prosthesis.

    Science.gov (United States)

    Hong, Feng; Wei, Bin; Chen, Lin

    2015-01-01

    Bacterial nanocellulose (BNC) has demonstrated a tempting prospect for applications in substitute of small blood vessels. However, present technology is inefficient in production and BNC tubes have a layered structure that may bring danger after implanting. Double oxygen-permeable silicone tubes in different diameters were therefore used as a tube-shape mold and also as oxygenated supports to construct a novel bioreactor for production of the tubular BNC materials. Double cannula technology was used to produce tubular BNC via cultivations with Acetobacter xylinum, and Kombucha, a symbiosis of acetic acid bacteria and yeasts. The results indicated that Kombucha gave higher yield and productivity of BNC than A. xylinum. Bacterial nanocellulose was simultaneously synthesized both on the inner surface of the outer silicone tube and on the outer surface of the inner silicone tube. Finally, the nano BNC fibrils from two directions formed a BNC tube with good structural integrity. Scanning electron microscopy inspection showed that the tubular BNC had a multilayer structure in the beginning but finally it disappeared and an intact BNC tube formed. The mechanical properties of BNC tubes were comparable with the reported value in literatures, demonstrating a great potential in vascular implants or in functional substitutes in biomedicine.

  7. Preliminary Study on Biosynthesis of Bacterial Nanocellulose Tubes in a Novel Double-Silicone-Tube Bioreactor for Potential Vascular Prosthesis

    Science.gov (United States)

    Wei, Bin; Chen, Lin

    2015-01-01

    Bacterial nanocellulose (BNC) has demonstrated a tempting prospect for applications in substitute of small blood vessels. However, present technology is inefficient in production and BNC tubes have a layered structure that may bring danger after implanting. Double oxygen-permeable silicone tubes in different diameters were therefore used as a tube-shape mold and also as oxygenated supports to construct a novel bioreactor for production of the tubular BNC materials. Double cannula technology was used to produce tubular BNC via cultivations with Acetobacter xylinum, and Kombucha, a symbiosis of acetic acid bacteria and yeasts. The results indicated that Kombucha gave higher yield and productivity of BNC than A. xylinum. Bacterial nanocellulose was simultaneously synthesized both on the inner surface of the outer silicone tube and on the outer surface of the inner silicone tube. Finally, the nano BNC fibrils from two directions formed a BNC tube with good structural integrity. Scanning electron microscopy inspection showed that the tubular BNC had a multilayer structure in the beginning but finally it disappeared and an intact BNC tube formed. The mechanical properties of BNC tubes were comparable with the reported value in literatures, demonstrating a great potential in vascular implants or in functional substitutes in biomedicine. PMID:26090420

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  10. 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

  11. Method and apparatus for testing closed-end tubes in heat exchangers of nuclear reactors and the like

    International Nuclear Information System (INIS)

    Seyd, G.; Bergbauer, A.; Paulsen, U.

    1975-01-01

    A description is given of a test stopper which is insertable into a tube closed at one end for testing the tightness of the tube with a fluid under pressure, the tube being in a heat exchanger of a nuclear reactor or the like. The test stopper includes a tubular outer jacket that is expandable outwardly to tightly seat the stopper in the tube. The stopper also has front and back end-face members joined to the ends of the outer jacket to define a closed space within the jacket. With the stopper inserted into the tube, the front end-face member and the closed end portion of the tube define a closed inner region of the tube. An inner tubular member, disposed within the outer jacket, partitions the closed space within the jacket into an annular outer chamber and a cylindrical inner chamber. A pressure-fluid supply selectively supplies fluid to the chambers. The outer jacket expands in response to fluid admitted to the annular chamber and the front end-face member has a through bore to admit fluid under pressure to the inner region of the tube. A method of testing of such a tube with a fluid under pressure includes inserting the test stopper into the tube and then expanding the outer jacket of the stopper to seat the stopper firmly in the tube. A fluid under pressure is directed through the stopper and into the closed region defined by the front end-face member of the stopper and the closed end portion of the tube. The pressure of the fluid introduced into this closed region is monitored for detecting a leak in the closed-end tube

  12. Benchmarking CRISPR on-target sgRNA design.

    Science.gov (United States)

    Yan, Jifang; Chuai, Guohui; Zhou, Chi; Zhu, Chenyu; Yang, Jing; Zhang, Chao; Gu, Feng; Xu, Han; Wei, Jia; Liu, Qi

    2017-02-15

    CRISPR (Clustered Regularly Interspaced Short Palindromic Repeats)-based gene editing has been widely implemented in various cell types and organisms. A major challenge in the effective application of the CRISPR system is the need to design highly efficient single-guide RNA (sgRNA) with minimal off-target cleavage. Several tools are available for sgRNA design, while limited tools were compared. In our opinion, benchmarking the performance of the available tools and indicating their applicable scenarios are important issues. Moreover, whether the reported sgRNA design rules are reproducible across different sgRNA libraries, cell types and organisms remains unclear. In our study, a systematic and unbiased benchmark of the sgRNA predicting efficacy was performed on nine representative on-target design tools, based on six benchmark data sets covering five different cell types. The benchmark study presented here provides novel quantitative insights into the available CRISPR tools. © The Author 2017. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  13. Staranje cevi uparjalnikov v Jedrski elektrarni Krško: Aging of tubes in the Krško nuclear power plant's steam generators:

    OpenAIRE

    Androjna, Ferdo; Cizelj, Leon

    2000-01-01

    The paper reviews the domestic efforts devoted to the safe and reliable operation of the Krško nuclear power plant (NPP) at full power, close to the design limit of the steam generators (18% of plugged tubes) for a full decade. This includes an overview of the recent status and history of the degradation processes, discussion of repair criteria, defining the acceptable size of defects and selected results from safety analyses supporting the operation of degraded steam generator (SG) tubes. It...

  14. Methodology for demonstrating the integrity of Steam Generator Tubes NPP Almaraz; Metodologia para la demostracion de la integridad de los tubos de Generador de Vapor de C. N. Almaraz

    Energy Technology Data Exchange (ETDEWEB)

    Campana Martin, J.; Cueto-Felgueroso Garcia, C.

    2013-07-01

    Steam Generator Program requires the performance of a Degradation Assessment prior to each refueling outage. The overall purpose of DA is to ensure that appropriate inspections are performed during the upcoming outage, and that the requisite information for integrity assessment is provided. Integrity assessment is performed after each SG tube inspection and includes two stages. The first one, Condition Monitoring is an assessment which confirms that SG tubes have met Performance Criteria during previous inspection interval. The second stage, Operational Assessment is an assessment which demonstrates that Performance Criteria will be met during the next inspection interval.

  15. Outer grid strap protruding spring repair apparatus

    International Nuclear Information System (INIS)

    Widener, W.H.

    1987-01-01

    This patent describes a nuclear fuel assembly grid spring repair apparatus for repairing a spring formed on an outer strap of a fuel assembly grid and having a portion protruding outwardly beyond the strap, the apparatus comprising: (a) a support frame defining an opening and having means defining a guide channel extending along the opening in a first direction; (b) means mounted on the frame and being adjustable for attaching the frame to the outer strap of the support grid so that the frame opening is aligned with the outwardly protruding spring on the outer strap; (c) an outer slide having a passageway defined therethrough and being mounted in the guide channel for reciprocable movement along the frame opening in the first direction for aligning the passageway with the outwardly protruding portion of the spring on the outer strap. The outer slide also has means defining a guide way extending along the passageway in a second direction generally orthogonal to the first direction; (d) a spring reset mechanism being operable for resetting the protruding spring to a nonprotruding position relative to the outer strap when the mechanism is aligned with the protruding portion of the spring; and (e) an inner slide supporting the spring reset mechanism and being mounted to the guide way for reciprocable movement along the passageway of the outer slide in the second direction for aligning the spring reset mechanism with the protruding portion of the spring on the outer strap

  16. 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

  17. 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

  18. Experimental facility design for a gap heat transfer in a double wall tube

    International Nuclear Information System (INIS)

    Nam, Ho Yun; Hong, Jong Gan; Kim, Jong Man; Kim, Jong Bum; Jeong, Ji Young

    2012-01-01

    A reliable steam generator design is one of the most critical issues in developing a sodium cooled fast reactor (SFR), and various efforts to avoid potential sodium water reaction (SWR) have been made. For this reason, SFR steam generators have been developed to improve its reliability using a double wall tube (DWT), which has two barriers between the sodium and water. Most steam generators for SFRs are the shell and tube type. Steam at high pressure and low temperature flows inside the inner tubes, which are heated by the shell side sodium at low pressure and high temperature. Since the inner and outer tubes of conventional DWTs are made of identical materials, the degree of thermal expansion is somewhat different between the two concentric tubes owing to their temperature difference. Therefore, a greater temperature difference results in less contact pressures between the inner and outer tubes. This feature results in a deterioration of the heat transfer capability of DWTs. Current developments are focused on an improvement of heat transfer capability by investigating the gap conductance between the two concentric tubes. To improve the heat transfer capability of DWTs, it is preferable to use different tube materials (Fig. 1). It is recommended to choose the inner tube material whose thermal expansion coefficient is greater than that of the outer tube by 10 to 15%

  19. Dimensional Measurements of Three Tubes by Computed Tomography

    International Nuclear Information System (INIS)

    Schneberk, D.J.; Martz, H.E. Jr.; Brown, W.D.

    2004-01-01

    Low density polyethylene (LDPE), copper (Cu), and gold (Au) tubes were scanned on KCAT to identify and evaluate the impact of phase effects on quantitative object recovery. These tubes are phantoms for high energy density capsules.[Logan, et al. 2004] Digital radiographs for each tube are shown in Figure 1. The LDPE tube was scanned at 60 kV, while the Cu and the Au tubes were scanned at 140 kV. All tubes were scanned at a magnification of 3, with approximately 100-mm distance between the exit plane of the tube and the scintillator. Notice the prominence of the outer bright and inner dark edges for the LDPE tube DR, and their absence from the Cu and Au tube DRs. The bright and dark edges are a result of change in phase of the x-rays. The x-ray fluence is partly attenuated and partly refracted. The location near the outer edge of the tube appears to be more attenuating since those x-rays have refracted to locations just outside the tube. Alternatively, the added counts from the refraction result in intensities that are greater than the incident intensity effectively representing a ''negative attenuation''. This results in more counts in that location than in the incident intensity image violating the ''positive-definite'' requirement for standard CT reconstruction methodologies. One aspect of our CT processing techniques remove some of this signal on the outside of the object. The goal of this paper is to evaluate the accuracy of our dimensional measurement methods for mesoscale object inspection

  20. Gas laser tube and method of fabricating same

    International Nuclear Information System (INIS)

    Garman, L.E.

    1975-01-01

    An improved gas laser tube is fabricated by counter boring the ends of a tubular aluminum extrusion having an inner tubular portion supported from an outer tubular portion via the intermediary of a plurality of radially directed support vanes or legs. Metallic transverse walls are sealed across the ends of the tubular extrusion to define the ends of a gas tight metallic envelope. An electrically insulative glow discharge tube is axially disposed within and supported by the inner tubular portion of the extrusion in axial alignment with an optical resonator of the laser tube. (U.S.)

  1. FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

    Science.gov (United States)

    Loeb, E.; Nicklas, J.H.

    1959-02-01

    A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

  2. Trash to Supply Gas (TtSG) Project Overview

    Science.gov (United States)

    Hintze, Paul; Santiago-Maldonado, Edgardo; Kulis, Michael J.; Lytle, John K.; Fisher, John W.; Vaccaro, Helen; Ewert, Michael K.; Broyan, James L.

    2012-01-01

    Technologies that reduce logistical needs are a key to long term space missions. Currently, trash and waste generated during a mission is carried during the entire roundtrip mission or stored inside a logistic module which is de-orbited into Earth's atmosphere for destruction. The goal of the Trash to Supply Gas (TtSG) project is to develop space technology alternatives for converting trash and other waste materials from human spaceflight into high-value products that might include propellants or power system fuels in addition to life support oxygen and water. In addition to producing a useful product from waste, TtSG will decrease the volume needed to store waste on long term space missions. This paper presents an overview of the TtSG technologies and future plans for the project.

  3. Formability of Micro-Tubes in Hydroforming

    International Nuclear Information System (INIS)

    Hartl, Christoph; Anyasodor, Gerald; Lungershausen, Joern

    2011-01-01

    Micro-hydroforming is a down-scaled metal forming process, based on the expansion of micro-tubes by internal pressurization within a die cavity. The objective of micro-hydroforming is to provide a technology for the economic mass production of complex shaped hollow micro-components. Influence of size effects in metal forming processes increases with scaling down of metal parts. Investigations into the change in formability of micro-tubes due to metal part scaling down constituted an important subject within the conducted fundamental research work. Experimental results are presented, concerning the analysis of the formability of micro-tubes made from stainless steel AISI 304 with an outer diameter of 800 μm and a wall thickness of 40 μm. An average ratio of tube wall thickness to grain size of 1.54 of up to 2.56 was analyzed. Miniaturised mechanical standard methods as well as bulge tests with internal hydrostatic pressurization of the tubular specimens were applied to analyze the influence of size-dependent effects. A test device was developed for the bulge experiments which enabled the pressurization of micro-tubes with internal pressures up to 4000 bar. To determine the attainable maximum achievable expansion ratio the tubes were pressurized in the bulge tests with increasing internal pressure until instability due to necking and subsequent bursting occurred. Comparisons with corresponding tests of macro-tubes, made from the here investigated material, showed a change in formability of micro-tubes which was attributed to the scaling down of the hydroforming process. In addition, a restricted applicability of existing theoretical correlations for the determination of the maximum pressure at bursting was observed for down-scaled micro-hydroforming.

  4. 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)

  5. Applicability of Alignment and Combination Rules to Burst Pressure Prediction of Multiple-flawed Steam Generator Tube

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Myeong Woo; Kim, Ji Seok; Kim, Yun Jae [Korea University, Seoul (Korea, Republic of); Jeon, Jun Young [Doosan Heavy Industries and Consruction, Seoul (Korea, Republic of); Lee, Dong Min [Korea Plant Service and Engineering, Technical Research and Development Institute, Naju (Korea, Republic of)

    2016-05-15

    Alignment and combination rules are provided by various codes and standards. These rules are used to determine whether multiple flaws should be treated as non-aligned or as coplanar, and independent or combined flaws. Experimental results on steam generator (SG) tube specimens containing multiple axial part-through-wall (PTW) flaws at room temperature (RT) are compared with assessment results based on the alignment and combination rules of the codes and standards. In case of axial collinear flaws, ASME, JSME, and BS7910 treated multiple flaws as independent flaws and API 579, A16, and FKM treated multiple flaws as combined single flaw. Assessment results of combined flaws were conservative. In case of axial non-aligned flaws, almost flaws were aligned and assessment results well correlate with experimental data. In case of axial parallel flaws, both effective flaw lengths of aligned flaws and separated flaws was are same because of each flaw length were same. This study investigates the applicability of alignment and combination rules for multiple flaws on the failure behavior of Alloy 690TT steam generator (SG) tubes that widely used in the nuclear power plan. Experimental data of burst tests on Alloy 690TT tubes with single and multiple flaws that conducted at room temperature (RT) by Kim el al. compared with the alignment rules of these codes and standards. Burst pressure of SG tubes with flaws are predicted using limit load solutions that provide by EPRI Handbook.

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

    International Nuclear Information System (INIS)

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

    2000-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-09-01

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

  8. Design of 3-D Printed Concentric Tube Robots.

    Science.gov (United States)

    Morimoto, Tania K; Okamura, Allison M

    2016-12-01

    Concentric tube surgical robots are minimally invasive devices with the advantages of snake-like reconfigurability, long and thin form factor, and placement of actuation outside the patient's body. These robots can also be designed and manufactured to acquire targets in specific patients for treating specific diseases in a manner that minimizes invasiveness. We propose that concentric tube robots can be manufactured using 3-D printing technology on a patient- and procedure-specific basis. In this paper, we define the design requirements and manufacturing constraints for 3-D printed concentric tube robots and experimentally demonstrate the capabilities of these robots. While numerous 3-D printing technologies and materials can be used to create such robots, one successful example uses selective laser sintering to make an outer tube with a polyether block amide and uses stereolithography to make an inner tube with a polypropylene-like material. This enables a tube pair with precurvatures of 0.0775 and 0.0455 mm -1 , which can withstand strains of 20% and 5.5% for the outer and inner tubes, respectively.

  9. Wall thickness measurements of tubes by Internal Rotary Inspection System (IRIS)- a comparative study with metallography

    International Nuclear Information System (INIS)

    Subramanian, C.V.; Joseph, A.; Ramesh, A.S.; Jayakumar, T.; Kalyanasundaram, P.; Baldev Raj

    1996-01-01

    Internal Rotary Inspection System (IRIS) is a relatively new ultrasonic system of heat exchanger/ steam condenser tubes and pipelines for measurement of wall thinning and pitting due to corrosion. The wall thickness measurements made during a scan around the circumference of the tube are displayed as a stationary rectilinear display of circumferential cross section (Bscan) of the tube. The paper describes the results obtained on tubes of various materials used in process industries having corrosion on inner and outer surfaces of the tube. (author)

  10. 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

  11. 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.

  12. 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

  13. Experimental study of flow patterns near tube support structures

    International Nuclear Information System (INIS)

    Rummens, H.E.C.; Turner, C.W.

    1994-07-01

    Extensive blockage of broached support plates in steam generators has occurred at the Bruce A Nuclear Generating Station (NGS), forcing unit derating in 1988 March. Blockage has also been found on the lower broached plates of the Pickering B and Point Lepreau NGSs. Water chemistry and operating conditions are known to influence fouling directly. We suspect that flow patterns also play a role, that these patterns are influenced by the geometry of steam generator (SG) components, and that particularly the broached plate design actively creates an environment favorable to deposition. Experiments are in progress to examine the flow patterns near various tube supports: the broached plate, two types of lattice bars, and the formed bars. Preliminary tests in an air/water loop with 1/2- and 7-tube SG mockups containing the tube supports have been completed. Flow patterns were visualized using injected air bubbles. Local velocities and turbulence levels were measured using a laser technique, which confirmed observations of flow recirculation and stagnation. Axial pressure profiles were measured to determine overall resistance coefficients, and to identify local pressure extremes. Some visualization tests were also carried out on an artificially fouled broached plate. Based on results to date, several deposition mechanisms are proposed: deposition of particles in stagnant regions, deposition of solubles due to flashing in low-pressure regions, and deposition in smaller channels due to steam migration toward larger channels. A qualitative assessment of the tube support designs based on these mechanisms implies that the relative resistances to fouling are: (WORST) broach plate << lattice bars << formed bars (BEST). As the air/water simulation shows only hydraulic flow patterns, further tests will be done in a simple liquid/vapor Freon loop to examine thermal effects. (author). 3 refs., 10 figs

  14. Improved method of measurement for outer leak

    International Nuclear Information System (INIS)

    Xu Guang

    2012-01-01

    Pneumatic pipeline is installed for the airborne radioactivity measurement equipment, air tightness and outer leak rate are essential for the testing of the characteristics, both in the national criteria and ISO standards, an improved practical method is available for the measurement of the outer air leak rate based on the engineering experiences for the equipment acceptance and testing procedure. (authors)

  15. Heat transfer in laminar flow for a finned double - tube

    International Nuclear Information System (INIS)

    Colle, S.

    1977-01-01

    An analitical study of the steady-state heat transfer in laminar flow in finned double-tube heat exchangers is presented. The fins are plane, straight and continous, equally spaced and are fixed over the external surface of the inner tube. A constant peripheral temperature distribution is assumed to apply over the inner tube surface and each fin, and a constant peripheral heat flux is assumed to apply over the outer tube surface, while the overall heat flux is suposed to be uniform in the longitudinal direction of the duct. The prediction of the thermal performance of the finned double-tube is made by means of the relationship between the Nusselt number, the boundary conditions and the geometric characteristcs of the duct. (author) [pt

  16. 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

  17. X-ray tubes

    International Nuclear Information System (INIS)

    Young, R.W.

    1979-01-01

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

  18. Pressure tube type reactors

    International Nuclear Information System (INIS)

    Komada, Masaoki.

    1981-01-01

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

  19. Gastrostomy feeding tube - bolus

    Science.gov (United States)

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

  20. Feeding tube insertion - gastrostomy

    Science.gov (United States)

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

  1. Neural Tube Defects

    Science.gov (United States)

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

  2. Proposed examination of defect detection of magnetic tube where alternating probe in type of insertion is used

    International Nuclear Information System (INIS)

    Kiya, Atsushi; Gotoh, Yuji; Sakurai, Kenta

    2008-01-01

    In various plants in the thermal power plant and the nuclear plant, a lot of steel tubes are used for various places such as heat exchangers, and these steel tubes should inspect regular for a healthy securing of the plant. Then, the outer side defect inspection of the magnetic substance steel tube using an electromagnetic phenomenon was examined in this research. It is shown that the inspection of the outer side defect on a steel tube with baffle is possible using the proposed method. (author)

  3. 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

  4. 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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  6. 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

  7. Development of a graphical user interface for sgRNAcas9 and its application.

    Science.gov (United States)

    Zhao, Chang-zhi; Zhang, Yi; Li, Guang-lei; Chen, Ji-liang; Li, Jing-Jin; Ren, Rui-min; Ni, Pan; Zhao, Shu-hong; Xie, Sheng-song

    2015-10-01

    The CRISPR/Cas9 genome editing technique is a powerful tool for researchers. However, off-target effects of the Cas9 nuclease activity is a recurrent concern of the CRISPR system. Thus, designing sgRNA (single guide RNA) with minimal off-target effects is very important. sgRNAcas9 is a software package, which can be used to design sgRNA and to evaluate potential off-target cleavage sites. In this study, a graphical user interface for sgRNAcas9 was developed using the Java programming language. In addition, off-target effect for sgRNAs was evaluated according to mismatched number and "seed sequence" specification. Moreover, sgRNAcas9 software was used to design 34 124 sgRNAs, which can target 4691 microRNA (miRNA) precursors from human, mouse, rat, pig, and chicken. In particular, the off-target effect of a sgRNA targeting to human miR-206 precursor was analyzed, and the on/off-target activity of this sgRNA was validated by T7E1 assay in vitro. Taken together, these data showed that the interface can simplify the usage of the sgRNAcas9 program, which can be used to design sgRNAs for the majority of miRNA precursors. We also found that the GC% of those sgRNAs ranged from 40% to 60%. In summary, the sgRNAcas9 software can be easily used to design sgRNA with minimal off-target effects for any species. The software can be downloaded from BiooTools website (http://www.biootools.com/).

  8. Identification of leaky steam generators by iodine mapping technique and development of tools for cutting of tubes of steam generators of Indian PHWRS

    International Nuclear Information System (INIS)

    Subba Rao, D.

    2006-01-01

    Kakrapar Atomic Power Station (2X220 MWe) located in Mandvi Taluka of Surat District in the state of Gujarat is the fifth Nuclear Power Station of the country. It has got an excellent record in the field of operation, safety, public awareness and emergency preparedness. KAPS Unit -1 achieved first criticality in Sep-1992 and was declared for commercial operation in may-1993. KAPS Unit -2 achieved first criticality in Jan-1995 and was declared for commercial operation in Sep-1995. So far station has generated about 30 billion units.Unit-1 achieved 98.4% and was graded as a world's No.1 in year 2002 amongst all CANDU type reactors. KAPS Unit -1 has made another record of operating continuously for more than 300 days in Indian PHWR s operating history. This paper mainly deals with the Indian PHWRs Steam Generators (SG) tube leaks, leaky steam generator identification by Iodine mapping, and development of special tool for cutting, removal and plugging of leaky tubes. These Steam Generators are designed by M/s Kraft Werke Union (KWU) of Siemens Group, West Germany, and Manufactured by M/s ENSA, SPAIN for Unit- 1 and by M/s MAN-GHH, Germany for Unit- 2. First time in October-2002 one of the Steam Generators of Unit-1 developed tube leak. To identify leaky Steam Generator, KAPS has established a method of Iodine mapping. With that the leaky SG was identified in very short time and corrective actions were taken immediately. Total three tube leaks (two in SG-4 of Unit-1 and one in SG-1 of unit-2) were experienced in both Units'. Following observations were made on SG tubes failure: All failures were in cold leg side; All Failures / deterioration locations were in front of main feed water nozzle; All Failures / deterioration locations were observed to be just above tube support plate (TSP) number 4 or 5; Deterioration ( i.e. wall thinning) observed from OD side and these tubes were adjacent to failed tubes; In all the three incidents, failed / deteriorated tubes were

  9. Incorporation of squalene into rod outer segments

    International Nuclear Information System (INIS)

    Keller, R.K.; Fliesler, S.J.

    1990-01-01

    We have reported previously that squalene is the major radiolabeled nonsaponifiable lipid product derived from [ 3 H]acetate in short term incubations of frog retinas. In the present study, we demonstrate that newly synthesized squalene is incorporated into rod outer segments under similar in vitro conditions. We show further that squalene is an endogenous constituent of frog rod outer segment membranes; its concentration is approximately 9.5 nmol/mumol of phospholipid or about 9% of the level of cholesterol. Pulse-chase experiments with radiolabeled precursors revealed no metabolism of outer segment squalene to sterols in up to 20 h of chase. Taken together with our previous absolute rate studies, these results suggest that most, if not all, of the squalene synthesized by the frog retina is transported to rod outer segments. Synthesis of protein is not required for squalene transport since puromycin had no effect on squalene incorporation into outer segments. Conversely, inhibition of isoprenoid synthesis with mevinolin had no effect on the incorporation of opsin into the outer segment. These latter results support the conclusion that the de novo synthesis and subsequent intracellular trafficking of opsin and isoprenoid lipids destined for the outer segment occur via independent mechanisms

  10. 21 CFR 868.5800 - Tracheostomy tube and tube cuff.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Tracheostomy tube and tube cuff. 868.5800 Section... (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Therapeutic Devices § 868.5800 Tracheostomy tube and tube cuff. (a) Identification. A tracheostomy tube and tube cuff is a device intended to be placed into a...

  11. Device and process for controlling the shoot peening efficiency, of a steam generator tube inner surface

    International Nuclear Information System (INIS)

    Isnardon, G.; Jacquier, P.; Voisembert, S.

    1988-01-01

    This device comprises an outer envelope of tubular shape applied on the face of the tubular plate around one end of the tube to be peened. A tool comprising a nozzle for the projection of the peening particles is axially mounted in the outer envelope. The controlling device comprises at least one piezoelectric sensor arranged to be in contact with the wall of the outer envelope and measuring means for the electrical signal generated by the sensor. The projection nozzle is brought into the outer envelope at the level of the sensor after each peening operation and the electrical voltage of the signal produced by the sensor is measured [fr

  12. A comparison of R-22, R-134a, R-410a, and R-407c condensation performance in smooth and enhanced tubes: Part 2, Pressure drop

    Energy Technology Data Exchange (ETDEWEB)

    Eckels, S J; Tesene, B A

    1999-07-01

    This paper reports pressure drops during condensation for R-22, R-134a, R-410a, and R-407c in three enhanced tubes and one smooth tube. The test tubes were a 3/8 inch outer diameter smooth tube, a 3/8 inch outer diameter microfin tube, a 5/16 inch outer diameter microfin tube, and a 5/8 inch outer diameter microfin tube. Pressure drops are reported at four mass fluxes, at two saturation temperatures, and over a range of average qualities in the test tubes. The pressure drops for R-410a were approximately 40% lower than those of R-22 in both tubes. R-407c had 10% to 20% lower pressure drops than R-22, while 134-a had slightly larger pressure drops than R-22. The microfin tube pressure drops were, on average, 40% to 80% higher than those for the smooth tube for all refrigerants. The pressure drop penalty of the microfin tube was shown to decrease with increased quality.

  13. Investigation of the Geometry of Metal Tube Walls after Necking in Uniaxial Tension

    Directory of Open Access Journals (Sweden)

    Chong Li

    2017-03-01

    Full Text Available Abstract: In order to characterize the deformation and true stress–strain relation of metal tubes, the geometry of tube walls after necking in uniaxial tension need to be determined. The paper investigated the necking process of metal tube. A large number of tensile tests and finite element analysis of 1Cr18Ni9Ti tubes with different sizes were conducted. It was found that the geometry of outer tube wall in the necking region can be described using a logistic regression model. The final geometry of the tube is determined by original tube diameter and wall thickness. The offset of tube walls are affected by two competing factors: volume constancy and necking. The offset distances of outer and inner walls are mainly affected by original wall thickness. The length of the necking zone is more influenced by original tube diameter. Tube elongation at fracture increases slightly as tube diameter gets larger, while the wall thickness has almost no impact on the elongation.

  14. Small RNAs controlling outer membrane porins

    DEFF Research Database (Denmark)

    Valentin-Hansen, Poul; Johansen, Jesper; Rasmussen, Anders A

    2007-01-01

    are key regulators of environmental stress. Recent work has revealed an intimate interplay between small RNA regulation of outer membrane proteins and the stress-induced sigmaE-signalling system, which has an essential role in the maintenance of the integrity of the outer membrane.......Gene regulation by small non-coding RNAs has been recognized as an important post-transcriptional regulatory mechanism for several years. In Gram-negative bacteria such as Escherichia coli and Salmonella, these RNAs control stress response and translation of outer membrane proteins and therefore...

  15. Intercostal drainage tube or intracardiac drainage tube?

    Directory of Open Access Journals (Sweden)

    N Anitha

    2016-01-01

    Full Text Available Although insertion of chest drain tubes is a common medical practice, there are risks associated with this procedure, especially when inexperienced physicians perform it. Wrong insertion of the tube has been known to cause morbidity and occasional mortality. We report a case where the left ventricle was accidentally punctured leading to near-exsanguination. This report is to highlight the need for experienced physicians to supervise the procedure and train the younger physician in the safe performance of the procedure.

  16. Intercostal drainage tube or intracardiac drainage tube?

    Science.gov (United States)

    Anitha, N; Kamath, S Ganesh; Khymdeit, Edison; Prabhu, Manjunath

    2016-01-01

    Although insertion of chest drain tubes is a common medical practice, there are risks associated with this procedure, especially when inexperienced physicians perform it. Wrong insertion of the tube has been known to cause morbidity and occasional mortality. We report a case where the left ventricle was accidentally punctured leading to near-exsanguination. This report is to highlight the need for experienced physicians to supervise the procedure and train the younger physician in the safe performance of the procedure.

  17. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... YouTube Videos » NEI YouTube Videos: Amblyopia Listen NEI YouTube Videos YouTube Videos Home Age-Related Macular Degeneration ... Retinopathy of Prematurity Science Spanish Videos Webinars NEI YouTube Videos: Amblyopia Embedded video for NEI YouTube Videos: ...

  18. Pediatric cuffed endotracheal tubes

    Directory of Open Access Journals (Sweden)

    Neerja Bhardwaj

    2013-01-01

    Full Text Available Endotracheal intubation in children is usually performed utilizing uncuffed endotracheal tubes for conduct of anesthesia as well as for prolonged ventilation in critical care units. However, uncuffed tubes may require multiple changes to avoid excessive air leak, with subsequent environmental pollution making the technique uneconomical. In addition, monitoring of ventilatory parameters, exhaled volumes, and end-expiratory gases may be unreliable. All these problems can be avoided by use of cuffed endotracheal tubes. Besides, cuffed endotracheal tubes may be of advantage in special situations like laparoscopic surgery and in surgical conditions at risk of aspiration. Magnetic resonance imaging (MRI scans in children have found the narrowest portion of larynx at rima glottides. Cuffed endotracheal tubes, therefore, will form a complete seal with low cuff pressure of <15 cm H 2 O without any increase in airway complications. Till recently, the use of cuffed endotracheal tubes was limited by variations in the tube design marketed by different manufacturers. The introduction of a new cuffed endotracheal tube in the market with improved tracheal sealing characteristics may encourage increased safe use of these tubes in clinical practice. A literature search using search words "cuffed endotracheal tube" and "children" from 1980 to January 2012 in PUBMED was conducted. Based on the search, the advantages and potential benefits of cuffed ETT are reviewed in this article.

  19. 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)

  20. Synthesis and characterization of a novel tube-in-tube nanostructured PPy/MnO2/CNTs composite for supercapacitor

    International Nuclear Information System (INIS)

    Li, Juan; Que, Tingli; Huang, Jianbin

    2013-01-01

    Graphical abstract: A novel tube-in-tube nanostructured PPy/MnO 2 /CNTs composite have been successfully fabricated. Its inner tubules are CNTs and the outer tubules are template-synthesized PPy. Most MnO 2 nanoparticles are sandwiched between the inner and outer wall, some relatively large particles are also latched onto the outside wall of the PPy tube. The composite yields a good electrochemical reversibility through 1000 cycles’ cyclic voltammogram (CV) test and galvanostatic charge–discharge experiments at different current densities. Display Omitted Highlights: ► We fabricate a ternary organic–inorganic complex of PPy/MnO 2 /CNTs composite. ► We characterize its morphological structures and properties by several techniques. ► The composite possesses the typical tube-in-tube nanostructures. ► Most MnO 2 nanoparticles are sandwiched between the inner CNTs and outer PPy wall. ► The composite has good electrochemical reversibility for supercapacitor. -- Abstract: Ternary organic–inorganic complex of polypyrrole/manganese dioxide/carbon nanotubes (PPy/MnO 2 /CNTs) composite was prepared by in situ chemical oxidation polymerization of pyrrole in the host of inorganic matrix of MnO 2 and CNTs, using complex of methyl orange (MO)/FeCl 3 was used as a reactive self-degraded soft-template. The morphological structures of the composite were characterized by scanning electron microscopy (SEM), transmission electron microscopy (TEM), high-resolution transmission electron microscopic (HRTEM), Fourier transform infrared spectroscopy (FT-IR) and X-ray diffraction (XRD), respectively. All the results indicate that the PPy/MnO 2 /CNTs composite possesses the typical tube-in-tube nanostructures: the inner tubules are CNTs and the outer tubules are template-synthesized PPy. MnO 2 nanoparticles may either sandwich the space between the inner and outer tubules or directly latch onto the wall of the PPy tubes. The composite yields a good electrochemical

  1. 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

  2. An experimental study on the impact collapse characteristics of CF/Epoxy circular tubes

    International Nuclear Information System (INIS)

    Kim, Y.N.; Im, K.H.; Park, J.W.; Yang, I.Y.

    2003-01-01

    This study is to investigate the energy absorption characteristics of CF/Epoxy (Carbon-Fiber/Epoxy Resin) circular tubes in static and impact tests. The experimental results varied significantly as a function of interlaminar number, orientation angle of outer and trigger. When a CFRP composite tube is crushed, static/impact energy is consumed by friction between the loading plate and the splayed fronds of the tube, by fracture of the fibers, matrix and their interface, and the response is complex and depends on the interaction among the different mechanisms, such as transverse shearing, laminar bending and local buckling. The collapse mode depended upon orientation angle of outer of CFRP tubes and loading status(static/impact). Typical collapse modes of CFRP tubes are wedge collapse mode, splaying collapse mode and fragmentation collapse mode

  3. 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)

  4. Lunar Lava Tube Sensing

    Science.gov (United States)

    York, Cheryl Lynn; Walden, Bryce; Billings, Thomas L.; Reeder, P. Douglas

    1992-01-01

    Large (greater than 300 m diameter) lava tube caverns appear to exist on the Moon and could provide substantial safety and cost benefits for lunar bases. Over 40 m of basalt and regolith constitute the lava tube roof and would protect both construction and operations. Constant temperatures of -20 C reduce thermal stress on structures and machines. Base designs need not incorporate heavy shielding, so lightweight materials can be used and construction can be expedited. Identification and characterization of lava tube caverns can be incorporated into current precursor lunar mission plans. Some searches can even be done from Earth. Specific recommendations for lunar lava tube search and exploration are (1) an Earth-based radar interferometer, (2) an Earth-penetrating radar (EPR) orbiter, (3) kinetic penetrators for lunar lava tube confirmation, (4) a 'Moon Bat' hovering rocket vehicle, and (5) the use of other proposed landers and orbiters to help find lunar lava tubes.

  5. Categorising YouTube

    DEFF Research Database (Denmark)

    Simonsen, Thomas Mosebo

    2011-01-01

    This article provides a genre analytical approach to creating a typology of the User Generated Content (UGC) of YouTube. The article investigates the construction of navigation processes on the YouTube website. It suggests a pragmatic genre approach that is expanded through a focus on YouTube......’s technological affordances. Through an analysis of the different pragmatic contexts of YouTube, it is argued that a taxonomic understanding of YouTube must be analysed in regards to the vacillation of a user-driven bottom-up folksonomy and a hierarchical browsing system that emphasises a culture of competition...... and which favours the already popular content of YouTube. With this taxonomic approach, the UGC videos are registered and analysed in terms of empirically based observations. The article identifies various UGC categories and their principal characteristics. Furthermore, general tendencies of the UGC within...

  6. 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

  7. 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.

  8. 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

  9. 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.

  10. In Silico Meets In Vivo: Towards Computational CRISPR-Based sgRNA Design.

    Science.gov (United States)

    Chuai, Guo-Hui; Wang, Qi-Long; Liu, Qi

    2017-01-01

    CRISPR-based genome editing has been widely implemented in various cell types. In silico single guide RNA (sgRNA) design is a key step for successful gene editing using the CRISPR system, and continuing efforts are aimed at refining in silico sgRNA design with high on-target efficacy and reduced off-target effects. Many sgRNA design tools are available, but careful assessments of their application scenarios and performance benchmarks across different types of genome-editing data are needed. Efficient in silico models can be built that integrate current heterogeneous genome-editing data to derive unbiased sgRNA design rules and identify key features for improving sgRNA design. Comprehensive evaluation of on-target and off-target effects of sgRNA will allow more precise genome editing and gene therapies using the CRISPR system. Copyright © 2016 Elsevier Ltd. All rights reserved.

  11. Laser performance upgrade for precise ICF experiment in SG-Ⅲ laser facility

    Directory of Open Access Journals (Sweden)

    Wanguo Zheng

    2017-09-01

    Full Text Available The SG-Ⅲ laser facility (SG-Ⅲ is the largest laser driver for inertial confinement fusion (ICF researches in China, which has 48 beamlines and can deliver 180 kJ ultraviolet laser energy in 3 ns. In order to meet the requirements of precise physics experiments, some new functionalities need to be added to SG-Ⅲ and some intrinsic laser performances need upgrade. So at the end of SG-Ⅲ's engineering construction, the 2-year laser performance upgrade project started. This paper will introduce the newly added functionalities and the latest laser performance of SG-Ⅲ. With these function extensions and performance upgrade, SG-Ⅲ is now fully prepared for precise ICF experiments and solidly paves the way towards fusion ignition.

  12. Rectangular drift tube characteristics

    International Nuclear Information System (INIS)

    Denisov, D.S.; Musienko, Yu.V.

    1985-01-01

    Results on the study of the characteristics of a 50 x 100 mm aluminium drift tube are presented. The tube was filled with argon-methane and argon-isobutane mixtures. With 16 per cent methane concentration the largest deviation from a linear relation between the drift time and the drift path over 50 mm is less than 2 mm. The tube filled with argon-isobutane mixture is capable of operating in a limited streamer mode

  13. Oscillations of the Outer Boundary of the Outer Radiation Belt During Sawtooth Oscillations

    Directory of Open Access Journals (Sweden)

    Jae-Hun Kim

    2006-09-01

    Full Text Available We report three sawtooth oscillation events observed at geosynchronous orbit where we find quasi-periodic (every 2-3 hours sudden flux increases followed by slow flux decreases at the energy levels of ˜50-400 keV. For these three sawtooth events, we have examined variations of the outer boundary of the outer radiation belt. In order to determine L values of the outer boundary, we have used data of relativistic electron flux observed by the SAMPEX satellite. We find that the outer boundary of the outer radiation belt oscillates periodically being consistent with sawtooth oscillation phases. Specifically, the outer boundary of the outer radiation belt expands (namely, the boundary L value increases following the sawtooth particle flux enhancement of each tooth, and then contracts (namely, the boundary L value decreases while the sawtooth flux decreases gradually until the next flux enhancement. On the other hand, it is repeatedly seen that the asymmetry of the magnetic field intensity between dayside and nightside decreases (increases due to the dipolarization (the stretching on the nightside as the sawtooth flux increases (decreases. This implies that the periodic magnetic field variations during the sawtooth oscillations are likely responsible for the expansion-contraction oscillations of the outer boundary of the outer radiation belt.

  14. Categorising YouTube

    OpenAIRE

    Simonsen, Thomas Mosebo

    2011-01-01

    This article provides a genre analytical approach to creating a typology of the User Generated Content (UGC) of YouTube. The article investigates the construction of navigation processes on the YouTube website. It suggests a pragmatic genre approach that is expanded through a focus on YouTube’s technological affordances. Through an analysis of the different pragmatic contexts of YouTube, it is argued that a taxonomic understanding of YouTube must be analysed in regards to the vacillation of a...

  15. Heated Tube Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Heated Tube Facility at NASA GRC investigates cooling issues by simulating conditions characteristic of rocket engine thrust chambers and high speed airbreathing...

  16. 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

  17. 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

  18. Modelling of pressure tube Quench using PDETWO

    International Nuclear Information System (INIS)

    Parlatan, Y.; Lei, Q.M.; Kwee, M.

    2004-01-01

    Transient two-dimensional heat conduction calculations have been carried out to determine the time-dependent temperature distribution in an overheated pressure tube during quenching with water. The purpose of the calculations is to provide input for evaluation of thermal (secondary) stresses in the pressure tube due to quench. The quench phenomenon in pressure tubes could occur in several hypothetical accident scenarios, including incidents involving intermittent buoyancy-induced flow during outages. In these scenarios, there will be two (radial and axial) or three dimensional temperature gradients, resulting in thermal stresses in the pressure tube, as the water front reaches and starts to cool down the hot pressure tube. The transient, two-dimensional heat conduction equation in the pressure tube during quench is solved using a FORTRAN package called PDETWO, available in the open literature for solving time-dependent coupled systems of non-linear partial differential equations over a two-dimensional rectangular region. This routine is based on finite difference solution of coupled, non-linear partial differential equations. Temperature gradient in the circumferential gradient is neglected for conservatism and convenience. The advancing water front is not modelled explicitly, and assumed to be at a uniform temperature and moving at a constant velocity inferred from experimental data. For outer surface and both ends of the pressure tube in the axial direction, a zero-heat flux boundary condition is assumed, while for the inner surface a moving water-quench front is assumed by appropriately varying the fluid temperature and the heat transfer coefficient. The pressure tube is assumed to be at a uniform temperature of 400 o C initially, to represent conditions expected during an intermittent buoyancy-influenced flow scenario. The results confirm the expectations that axial temperature gradients and associated heat fluxes are small in comparison with those in the

  19. Stop Smoking—Tube-In-Tube Helical System for Flameless Calcination of Minerals

    Directory of Open Access Journals (Sweden)

    Nils Haneklaus

    2017-11-01

    Full Text Available Mineral calcination worldwide accounts for some 5–10% of all anthropogenic carbon dioxide (CO2 emissions per year. Roughly half of the CO2 released results from burning fossil fuels for heat generation, while the other half is a product of the calcination reaction itself. Traditionally, the fuel combustion process and the calcination reaction take place together to enhance heat transfer. Systems have been proposed that separate fuel combustion and calcination to allow for the sequestration of pure CO2 from the calcination reaction for later storage/use and capture of the combustion gases. This work presents a new tube-in-tube helical system for the calcination of minerals that can use different heat transfer fluids (HTFs, employed or foreseen in concentrated solar power (CSP plants. The system is labeled ‘flameless’ since the HTF can be heated by other means than burning fossil fuels. If CSP or high-temperature nuclear reactors are used, direct CO2 emissions can be divided in half. The technical feasibility of the system has been accessed with a brief parametric study here. The results suggest that the introduced system is technically feasible given the parameters (total heat transfer coefficients, mass- and volume flows, outer tube friction factors, and –Nusselt numbers that are examined. Further experimental work will be required to better understand the performance of the tube-in-tube helical system for the flameless calcination of minerals.

  20. Template synthesis of test tube nanoparticles using non-destructive replication.

    Science.gov (United States)

    Wagner, Jonathan; Yao, Jingyuan; Rodgers, David; Hinds, Bruce

    2013-03-01

    Nano test tubes are a promising delivery vehicle for a range of therapeutics, including small molecule drugs and biologics. However, current template synthesis methods of producing nano test tubes are prohibitively expensive and time consuming. Here, non-destructive template replication was used to increase nano test tube yield from porous alumina by more than a hundredfold. We demonstrate how to produce nano test tubes of several sizes and compositions, including hybrid tubes with different inner and outer surfaces for targeted surface chemistry. Nano test tubes were readily suspended and stored in aqueous solutions without the need for chemical treatment. These nano test tubes should find application as delivery vehicles for therapeutics, particularly for processive 'bionanoreactors' loaded with enzymes.

  1. Template synthesis of test tube nanoparticles using non-destructive replication

    International Nuclear Information System (INIS)

    Wagner, Jonathan; Rodgers, David; Yao Jingyuan; Hinds, Bruce

    2013-01-01

    Nano test tubes are a promising delivery vehicle for a range of therapeutics, including small molecule drugs and biologics. However, current template synthesis methods of producing nano test tubes are prohibitively expensive and time consuming. Here, non-destructive template replication was used to increase nano test tube yield from porous alumina by more than a hundredfold. We demonstrate how to produce nano test tubes of several sizes and compositions, including hybrid tubes with different inner and outer surfaces for targeted surface chemistry. Nano test tubes were readily suspended and stored in aqueous solutions without the need for chemical treatment. These nano test tubes should find application as delivery vehicles for therapeutics, particularly for processive ‘bionanoreactors’ loaded with enzymes. (paper)

  2. 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

  3. An Intraoral Miniature X-ray Tube Based on Carbon Nanotubes for Dental Radiography

    OpenAIRE

    Hyun Jin Kim; Hyun Nam Kim; Hamid Saeed Raza; Han Beom Park; Sung Oh Cho

    2016-01-01

    A miniature X-ray tube based on a carbon-nanotube electron emitter has been employed for the application to a dental radiography. The miniature X-ray tube has an outer diameter of 7 mm and a length of 47 mm. The miniature X-ray tube is operated in a negative high-voltage mode in which the X-ray target is electrically grounded. In addition, X-rays are generated only to the teeth directions using a collimator while X-rays generated to other directions are shielded. Hence, the X-ray tube can be ...

  4. Doppler method leak detection for LMFBR steam generators. Pt. 2. Detection characteristics of bubble in-water using large scale SG model

    International Nuclear Information System (INIS)

    Kumagai, Hiromichi

    2000-01-01

    To prevent the expansion of tube damage and to maintain structural integrity in the steam generators (SGs) of a fast breeder reactor (FBR), it is necessary to detect precisely and immediately the leakage of water from heat transfer tubes. Therefore, an active acoustic method was developed. Previous studies have revealed that, in practical steam generators, the active acoustic method can detect bubbles of 10 l/s within 10 seconds. However to prevent the expansion of damage to neighboring tubes, it is necessary to detect smaller leakages of water from the heat transfer tubes. The Doppler method is designed to detect small leakages and to find the source of a leak before damage spreads to neighboring tubes. The detection sensitivity of the Doppler method and the influence of background noise were investigated experimentally. In-water experiments were performed using an SG full-sector model that simulates actual SGs. The results show that the Doppler method can detect bubbles of 0.1 l/s (equivalent to a water leak rate of about 0.1 g/s) within a few seconds and that the background noise has little effect on water leak detection performance. The Doppler method thus has great potential for the detection of water leakage in SGs. (author)

  5. Dispersant Application during SG Wet Layup at SK Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Hyukchul; Lee, Dooho; Sung, Kibang [KHNP Central Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The corrosion products in the feedwater are deposited onto the steam generators (SGs) despite the effort to control them within limit of impurity. This deposit is one of causes for occurrence of SCC (Stress Corrosion Cracking), water level fluctuation and further corrosion of SGs. To minimize corrosion and remove deposit, the nuclear power plants apply high pH to the secondary system and SG chemical cleaning, respectively. But these methods can be costly and carry risks of extended outages or incomplete cleaning. Another method is an on-line dispersant application. The role of dispersant is to make deposit suspended in the SGs. Then, the suspended deposit is discharged to the blowdown system. The iron removal is increased in the blowdown system during the dispersant application. Additional significant benefit in the form of reduced corrosion product transport may be obtained through applying dispersant in the SGs wet lay operational mode. This method helps to reduce the total SGs loading without affecting critical outage activities and with minimal additional effort on the part of the utilities. This study provides the results of the dispersant application trial during the SG wet layup at SK Unit 1. As the PAA concentrations were increased, the corrosion rates of Alloy 690 and SA 106 Gr.B were increased. The corrosion rate of Alloy 690 was 2 times less than that of SA 106 Gr.B at 100 ppm of PAA based on the electrochemical experimental. There were no significant feasibility problems with application of PAA during the SG wet layup. The reasonable estimation of the additional mass removed by the presence of PAA during SGs wet layup is 460 g. The iron removal depended on PAA concentration injected based on the comparative results of the SK Unit 1 and TMI-1. It is expected that injection of PAA into the SG result in a significant decrease in the amount of iron transported to the SGs during the startup.

  6. Secondary circuit water chemistry and related problems with SG

    Energy Technology Data Exchange (ETDEWEB)

    Ignatov, V; Ivanov, V [Balakovo Nuclear Power Plant (Russian Federation)

    2001-07-01

    Necessity for SG feed water and blowdown systems modernization Balakovo NPP steam generators PGV-1000M was identified at Units with VVER-1000 during commissioning separational, thermo-hydraulic and thermo-chemical testings. It was discovered, that in zone of 'hot' header coolant salt concentration (concentration of dissolved salts) was almost 2 times more, than salt concentration in blowdown water. A number of chemical testings was performed to investigate and optimize salts distribution in water volume of PGV-1000. (R.P.)

  7. Secondary circuit water chemistry and related problems with SG

    International Nuclear Information System (INIS)

    Ignatov, V.; Ivanov, V.

    2001-01-01

    Necessity for SG feed water and blowdown systems modernization Balakovo NPP steam generators PGV-1000M was identified at Units with VVER-1000 during commissioning separational, thermo-hydraulic and thermo-chemical testings. It was discovered, that in zone of 'hot' header coolant salt concentration (concentration of dissolved salts) was almost 2 times more, than salt concentration in blowdown water. A number of chemical testings was performed to investigate and optimize salts distribution in water volume of PGV-1000. (R.P.)

  8. Method for shaping polyethylene tubing

    Science.gov (United States)

    Kramer, R. C.

    1981-01-01

    Method forms polyethylene plastic tubing into configurations previously only possible with metal tubing. By using polyethylene in place of copper or stain less steel tubing inlow pressure systems, fabrication costs are significantly reduced. Polyethylene tubing can be used whenever low pressure tubing is needed in oil operations, aircraft and space applications, powerplants, and testing laboratories.

  9. Molybdenum Tube Characterization report

    Energy Technology Data Exchange (ETDEWEB)

    Beaux II, Miles Frank [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Usov, Igor Olegovich [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-07

    Chemical vapor deposition (CVD) techniques have been utilized to produce free-standing molybdenum tubes with the end goal of nuclear fuel clad applications. In order to produce tubes with properties desirable for this application, deposition rates were lowered requiring long deposition durations on the order of 50 hours. Standard CVD methods as well as fluidized-bed CVD (FBCVD) methods were applied towards these objectives. Characterization of the tubes produced in this manner revealed material suitable for fuel clad applications, but lacking necessary uniformity across the length of the tubes. The production of freestanding Mo tubes that possess the desired properties across their entire length represents an engineering challenge that can be overcome in a next iteration of the deposition system.

  10. 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

  11. Outer planet probe cost estimates: First impressions

    Science.gov (United States)

    Niehoff, J.

    1974-01-01

    An examination was made of early estimates of outer planetary atmospheric probe cost by comparing the estimates with past planetary projects. Of particular interest is identification of project elements which are likely cost drivers for future probe missions. Data are divided into two parts: first, the description of a cost model developed by SAI for the Planetary Programs Office of NASA, and second, use of this model and its data base to evaluate estimates of probe costs. Several observations are offered in conclusion regarding the credibility of current estimates and specific areas of the outer planet probe concept most vulnerable to cost escalation.

  12. 27 CFR 9.207 - Outer Coastal Plain.

    Science.gov (United States)

    2010-04-01

    ... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Outer Coastal Plain. 9.207... Outer Coastal Plain. (a) Name. The name of the viticultural area described in this section is “Outer Coastal Plain”. For purposes of part 4 of this chapter, “Outer Coastal Plain” is a term of viticultural...

  13. Influence of the S/N ratio on the corrosion release of Alloy 690 tubes in a primary coolant

    International Nuclear Information System (INIS)

    Shim, Hee-Sang; Choi, Myung Sik; Kim, Kyung Mo; Seo, Myung Ji; Hur, Do Haeng; Choi, Tack-Sang; Yoo, One

    2014-01-01

    Alloy 690TT is a promising steam generator (SG) tube material of a pressurized water reactor due to its excellent resistance to stress corrosion cracking (SCC) that has caused problems in Alloy 600 as an old SG tube material. The qualities of this material have been managed thoroughly from manufacturing step under various specification regulations as well as in in-service step. For examples, the surface roughness are prescribed as the values less than 1.6 μm for the tube outside and 0.5 μm for the inside, respectively. In addition, the surface state and defect must be qualified through the eddy current test (ECT) and the ultrasonic test (UT) according to the ASME Section III, NB2550. Then, the signal-to-noise (S/N) ratio, which is measured using ECT bobbin probe, is the important criteria to determine the material and it shall be 15 to 1 or higher at the standard frequency for any fixed 0.5 m length of any tube. The corrosion behaviours of the Alloy 690TT under high-temperature pressurized primary water have been studied widely in a point of the SCC but discussed narrowly in a point of the corrosion release. In particular, the effect of the S/N ratio on the corrosion release of this material surface has been rarely investigated. In this work, we evaluate the influence of the S/N ratio on the corrosion release of Alloy 690 SG tubes. The specimens with different S/N ratio were selected through ECT bobbin inspection and a corrosion release test was conducted using a simulated primary circulation loop. The material properties and oxidation behaviours were investigated by surface profiler, scanning electron microscopy, transmission electron microscopy, grazing incidence X-ray diffraction and etc. As a result, the corrosion rate was matched preferably with the MRPC characteristics showing macroscopic surface state rather than with the bobbin S/N ratio results. (author)

  14. The outer tracker detector of the HERA-B experiment. Pt. 3. Operation and performance

    International Nuclear Information System (INIS)

    Albrecht, H.; Bauer, T.S.; Utrecht Univ.; Beck, M.

    2006-12-01

    In this paper we describe the operation and performance of the HERA-B Outer Tracker, a 112674 channel system of planar drift tube layers. The performance of the HERA-B Outer Tracker system fullfilled all requirements for stable and efficient operation in a hadronic environment, thus confirming the adequacy of the honeycomb drift tube technology and of the front-end readout system. The detector was stably operated with a gas gain of 3 . 10 4 in an Ar/CF 4 /CO 2 (65:35:5) gas mixture, yielding a good efficiency for triggering and track reconstruction, larger than 95 % for tracks with momenta above 5 GeV/c. The hit resolution of the drift cells was 300 to 320 μm and the relative momentum resolution can be described as: σ(p)/p(%) = (1.61 ± 0.02) + (0.0051 ± 0.0006) . p. At the end of the HERA-B running no aging effects in the Outer Tracker cells were observed. (orig.)

  15. Heating device for thermal treatment of curred small diameter tubes and utilization of this device

    International Nuclear Information System (INIS)

    Jacquier, P.

    1988-01-01

    The heating device is made by a helical winding constituted from a resistance heating wire. The heating wire constituted the central core of a coaxial cable comprising an outer tubular metal envelope and an insulating layer interpolated between the central core and the outer envelope. The coaxial cable is wound in order to form a helical winding that forms the flexible element for introduction to the tube to be treated [fr

  16. 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

  17. Outer-2-independent domination in graphs

    Indian Academy of Sciences (India)

    independent dominating set of a graph is a set of vertices of such that every vertex of ()\\ has a neighbor in and the maximum vertex degree of the subgraph induced by ()\\ is at most one. The outer-2-independent domination ...

  18. Intershell correlations in photoionization of outer shells

    Energy Technology Data Exchange (ETDEWEB)

    Amusia, M.Ya. [The Racah Institute of Physics, The Hebrew University of Jerusalem, Jerusalem 91904 (Israel); A.F. Ioffe Physical-Technical Institute, St. Petersburg 194021 (Russian Federation); Chernysheva, L.V. [A.F. Ioffe Physical-Technical Institute, St. Petersburg 194021 (Russian Federation); Drukarev, E.G. [National Research Center “Kurchatov Institute”, Konstantinov Petersburg Nuclear Physics Institute, St. Petersburg 188300 (Russian Federation)

    2016-02-15

    We demonstrate that the cross sections for photoionization of the outer shells are noticeably modified at the photon energies close to the thresholds of ionization of the inner shells due to correlations with the latter. The correlations may lead to increase or to decrease of the cross sections just above the ionization thresholds.

  19. Outer-2-independent domination in graphs

    Indian Academy of Sciences (India)

    Outer-2-independent domination in graphs. MARCIN KRZYWKOWSKI1,2,∗, DOOST ALI MOJDEH3 and MARYEM RAOOFI4. 1Department of Pure and Applied Mathematics, University of Johannesburg,. Johannesburg, South Africa. 2Faculty of Electronics, Telecommunications and Informatics, Gdansk University.

  20. Intershell correlations in photoionization of outer shells

    International Nuclear Information System (INIS)

    Amusia, M.Ya.; Chernysheva, L.V.; Drukarev, E.G.

    2016-01-01

    We demonstrate that the cross sections for photoionization of the outer shells are noticeably modified at the photon energies close to the thresholds of ionization of the inner shells due to correlations with the latter. The correlations may lead to increase or to decrease of the cross sections just above the ionization thresholds.

  1. EDF field experience of 182 J-Groove welds on CRDMs and SG channel head nozzles

    International Nuclear Information System (INIS)

    Duisabeau, L.; Deforge, D.; Thebault, Y.; Stindel, M.; Lemaire, E.

    2011-01-01

    The Reactor Pressure Vessel Head (RPVH) replacement program, which began after a leak occurrence in a vessel head nozzle in Alloy 600 at Bugey Unit 3, was a unique opportunity to perform an extended inspection program on the welds from the decommissioned RPV heads. This paper presents the actual results of this program. More than 800 CRDM J groove welds from 18 decommissioned RPV heads were inspected by automatic dye penetrant testing. Detected indications were characterized by viewing tools specifically developed and in some specific cases, by destructive investigations in hot lab. Some welding defects were observed but no indication corresponding to stress corrosion cracking (SCC) was detected at the welds wet surface nor propagation from welding manufacturing defects, including the weld with the longest operating time on EDF power plants (170 000 h). Very few cases of SCC propagation from Alloy 600 to Alloy 182 are reported. One case of initiation at the weld root pass was observed. From design, the weld root pass (mechanically loaded) of CRDM (Control Rod Drive Mechanism) nozzles is not in contact with primary water and the cracking observed occurred after a through wall cracking of the Alloy 600 tube, enabling primary water to wet the root pass. Concerning the steam generator (SG) drain nozzle, the alloy 182 weld root is directly in contact with primary water. In June 2008, a primary water leakage was suspected on a steam generator bowl drain while conducting a bare metal visual examination during the plant's outage. Dye penetrant testing of the weld and metallographic replica were implemented during the 2008 and 2009 refuelling outages to confirm a leakage by SCC. Manufacturing reports analyses revealed that the drain nozzle weld was repaired and had not been stress relieved during manufacturing. EDF has decided to plug this nozzle and to enforce the maintenance policy for similar components with the same manufacturing specificity. Regarding national and

  2. Automation in tube finishing bay

    International Nuclear Information System (INIS)

    Bhatnagar, Prateek; Satyadev, B.; Raghuraman, S.; Syama Sundara Rao, B.

    1997-01-01

    Automation concept in tube finishing bay, introduced after the final pass annealing of PHWR tubes resulted in integration of number of sub-systems in synchronisation with each other to produce final cut fuel tubes of specified length, tube finish etc. The tube finishing bay which was physically segregated into four distinct areas: 1. tube spreader and stacking area, 2. I.D. sand blasting area, 3. end conditioning, wad blowing, end capping and O.D. wet grinding area, 4. tube inspection, tube cutting and stacking area has been studied

  3. Electrohydrodynamic enhancement of in-tube convective condensation heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Sadek, H.; Robinson, A.J.; Ching, C.Y.; Shoukri, M. [McMaster University, Department of Mechanical Engineering, Hamilton, Ont. (Canada); Cotton, J.S. [Dana Corporation, Long Manufacturing Division, Oakville, Ont. (Canada)

    2006-05-15

    An experimental investigation of electrohydrodynamic (EHD) augmentation of heat transfer for in-tube condensation of flowing refrigerant HFC-134a has been performed in a horizontal, single-pass, counter-current heat exchanger with a rod electrode placed in the centre of the tube. The effects of varying the mass flux (55kg/m{sup 2}s==<263kg/m{sup 2}s), inlet quality (0.2=tube. (author)

  4. 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

  5. Estimation of S/G ratio in woods using 1064 nm FT-Raman spectroscopy

    Science.gov (United States)

    Umesh P. Agarwal; Sally A. Ralph; Dharshana Padmakshan; Sarah Liu; Steven D. Karlen; Cliff Foster; John Ralph

    2015-01-01

    Two simple methods based on the 370 cm-1 Raman band intensity were developed for estimation of syringyl-to-guaiacyl (S/G) ratio in woods. The methods, in principle, are representative of the whole cell wall lignin and not just the portion of lignin that gets cleaved to release monomers, for example, during certain S/G chemical analyses. As such,...

  6. CRISPRscan: designing highly efficient sgRNAs for CRISPR/Cas9 targeting in vivo

    Science.gov (United States)

    Moreno-Mateos, Miguel A.; Vejnar, Charles E.; Beaudoin, Jean-Denis; Fernandez, Juan P.; Mis, Emily K.; Khokha, Mustafa K.; Giraldez, Antonio J.

    2015-01-01

    CRISPR/Cas9 technology provides a powerful system for genome engineering. However, variable activity across different single guide RNAs (sgRNAs) remains a significant limitation. We have analyzed the molecular features that influence sgRNA stability, activity and loading into Cas9 in vivo. We observe that guanine enrichment and adenine depletion increase sgRNA stability and activity, while loading, nucleosome positioning and Cas9 off-target binding are not major determinants. We additionally identified truncated and 5′ mismatch-containing sgRNAs as efficient alternatives to canonical sgRNAs. Based on these results, we created a predictive sgRNA-scoring algorithm (CRISPRscan.org) that effectively captures the sequence features affecting Cas9/sgRNA activity in vivo. Finally, we show that targeting Cas9 to the germ line using a Cas9-nanos-3′-UTR fusion can generate maternal-zygotic mutants, increase viability and reduce somatic mutations. Together, these results provide novel insights into the determinants that influence Cas9 activity and a framework to identify highly efficient sgRNAs for genome targeting in vivo. PMID:26322839

  7. Thermal-hydraulic behavior on break simulation of steam generator U-tube

    International Nuclear Information System (INIS)

    Seul, Kwang Won; Bang, Young Seok; Lee, Sukho; Kim, Hho Jung

    1995-01-01

    The thermal-hydraulic behavior depending on the break simulation in a steam generator U-tube was investigated and identified the code predictability on plant responses during SGTR accident. The calculated results were compared and assessed with LSTF SB-SG-06 test data. The RELAP5/MOD3.1 code well predicted the sequence of events and the significant phenomena, such as the asymmetric loop behavior, the RCS cooldown and heat transfer by the natural circulation, and system depressurization, even though there were some differences from the experimental data. The break flowrate was found to be sensitive to the break model and affected the system behavior

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  9. Helically coiled tube heat exchanger

    International Nuclear Information System (INIS)

    Harris, A.M.

    1981-01-01

    In a heat exchanger such as a steam generator for a nuclear reactor, two or more bundles of helically coiled tubes are arranged in series with the tubes in each bundle integrally continuing through the tube bundles arranged in series therewith. Pitch values for the tubing in any pair of tube bundles, taken transverse to the path of the reactor coolant flow about the tubes, are selected as a ratio of two unequal integers to permit efficient operation of each tube bundle while maintaining the various tube bundles of the heat exchanger within a compact envelope. Preferably, the helix angle and tube pitch parallel to the path of coolant flow are constant for all tubes in a single bundle so that the tubes are of approximately the same length within each bundle

  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. X-ray tube

    International Nuclear Information System (INIS)

    Webley, R.S.

    1975-01-01

    The object of the invention described is to provide an X-ray tube providing a scanned X-ray output which does not require a scanned electron beam. This is obtained by an X-ray tube including an anode which is rotatable about an axis, and a source of a beam of energy, for example an electron beam, arranged to impinge on a surface of the anode to generate X-radiation substantially at the region of incidence on the anode surface. The anode is rotatable about the axis to move the region of incidence over the surface. The anode is so shaped that the rotation causes the region of incidence to move in a predetermined manner relative to fixed parts of the tube so that the generated X-radiation is scanned in a predetermined manner relative to the tube. (UK)

  12. Fuel assembly guide tube

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1979-01-01

    This invention is directed toward a nuclear fuel assembly guide tube arrangement which restrains spacer grid movement due to coolant flow and which offers secondary means for supporting a fuel assembly during handling and transfer operations

  13. Bull Moose Tube Company

    Science.gov (United States)

    The EPA is providing notice of a proposed Administrative Penalty Assessment against the Bull Moose Tube Company, a business located at 1819 Clarkson Road, Chesterfield, MO, 63017, for alleged violations at the facility located at 406 East Industrial Drive,

  14. Tracheostomy tube - eating

    Science.gov (United States)

    Trach - eating ... take your first bites. Certain factors may make eating or swallowing harder, such as: Changes in the ... easier to swallow. Suction the tracheostomy tube before eating. This will keep you from coughing while eating, ...

  15. Streak tube development

    International Nuclear Information System (INIS)

    Hinrichs, C.K.; Estrella, R.M.

    1979-01-01

    A research program for the development of a high-speed, high-resolution streak image tube is described. This is one task in the development of a streak camera system with digital electronic readout, whose primary application is for diagnostics in underground nuclear testing. This program is concerned with the development of a high-resolution streak image tube compatible with x-ray input and electronic digital output. The tube must be capable of time resolution down to 100 psec and spatial resolution to provide greater than 1000 resolution elements across the cathode (much greater than presently available). Another objective is to develop the capability to make design changes in tube configurations to meet different experimental requirements. A demountable prototype streak tube was constructed, mounted on an optical bench, and placed in a vacuum system. Initial measurements of the tube resolution with an undeflected image show a resolution of 32 line pairs per millimeter over a cathode diameter of one inch, which is consistent with the predictions of the computer simulations. With the initial set of unoptmized deflection plates, the resolution pattern appeared to remain unchanged for static deflections of +- 1/2-inch, a total streak length of one inch, also consistent with the computer simulations. A passively mode-locked frequency-doubled dye laser is being developed as an ultraviolet pulsed light source to measure dynamic tube resolution during streaking. A sweep circuit to provide the deflection voltage in the prototype tube has been designed and constructed and provides a relatively linear ramp voltage with ramp durations adjustable between 10 and 1000 nsec

  16. Researching YouTube

    OpenAIRE

    Arthurs, Jane; Drakopoulou, Sophia; Gandini, Alessandro

    2018-01-01

    ‘Researching YouTube’ introduces the special issue of Convergence which arose out of an international academic conference on YouTube that was held in London at Middlesex University in September 2016. The conference aimed to generate a robust overview of YouTube’s changing character and significance after its first ten years of development by creating a productive dialogue between speakers from different disciplines and cultures, and between YouTube-specific research and wider debates in media...

  17. Tubing crimping pliers

    Science.gov (United States)

    Lindholm, G.T.

    1981-02-27

    The disclosure relates to pliers and more particularly to pliers for crimping two or more pieces of copper tubing together prior to their being permanently joined by brazing, soldering or the like. A die containing spring-loaded pins rotates within a cammed ring in the head of the pliers. As the die rotates, the pins force a crimp on tubing held within the pliers.

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  19. Emergency transfer tube closure and process for sealing transfer tube under emergency conditions

    International Nuclear Information System (INIS)

    Hardin, R.T. Jr.; Marshall, J.R.

    1987-01-01

    In a nuclear fuel reactor well that includes a transfer tube projecting outwardly from wall thereof, the transfer tube is described having a first closure assembly. The transfer tube has a circumferential flange extending outwardly laterally therefrom, an emergency transfer tube closure therefor comprising; a pair of elongated, vertically-extending U-shaped guides, one U-shaped guide disposed laterally on each side of the transfer tube, each of the U-shaped guides comprising a base and laterally extending flanges thereon, the U-shaped guides having their open ends facing each other, a closure plate, having a surface facing the circumferential flange greater in area than the area circumscribed by the outer circumference of the circumferential flange, vertically disposed the U-shaped guides, the closure plate normally being disposed in a vertical plane just slightly in front of the vertical plane of the circumferential flange, two pairs of rollers, one pair of which is rotatably mounted on each side of the closure plate adjacent the U-shaped guides, riding on the inner portion of each of the flanges of each of the U-shaped guides. Each of the U-shaped guides is provided with a pair of spatially disposed openings on a flange thereof adjacent the wall of the nuclear fuel reactor well, each of the pairs of openings being disposed on each of the U-shaped guides a distance equal to the distance between the center lines of the corresponding pair of rollers riding within the U-shaped guides, each of the openings being sufficiently large to receive a corresponding roller of the pairs of rollers in the U-shaped guides. The openings is shaped on the flanges of the U-shaped guides so that when the pairs of rollers are disposed therein, the face of the closure plate will be in sealing engagement with the circumferential flange of the transfer tube

  20. Transport comparison of multiwall carbon nanotubes by contacting outer shell and all shells.

    Science.gov (United States)

    Luo, Qiang; Cui, A-Juan; Zhang, Yi-Guang; Lu, Chao; Jin, Ai-Zi; Yang, Hai-Fang; Gu, Chang-Zhi

    2010-11-01

    Carbon nanotubes, particularly multiwall carbon nanotubes (MWCNTs) can serve as interconnects in nanoelectronic devices and integrated circuits because of their extremely large current-carrying capacity. Many experimental results about the transport properties of individual MWCNTs by contacting outer shell or all shells have been reported. In this work, a compatible method with integrated circuit manufacturing process was presented to compare the transport property of an individual multiwall carbon nanotube (MWCNT) by contacting outer shell only and all shells successively. First of the Ti/Au electrodes contacting outer shell only were fabricated onto the nanotube through the sequence of electron beam lithography (EBL) patterning, metal deposition and lift-off process. After the characterization of its transport property, focused ion beam (FIB) was used to drill holes through the same nanotube at the as-deposited electrodes. Then new contact to the holes and electrodes were made by ion-induced deposition of tungsten from W(CO)6 precursor gas. The transport results indicated that the new contact to all shells can clear up the intershell resistance and the electrical conductance of the tube can be improved about 8 times compared to that of by contacting outer shell only.

  1. Acute Zonal Cone Photoreceptor Outer Segment Loss.

    Science.gov (United States)

    Aleman, Tomas S; Sandhu, Harpal S; Serrano, Leona W; Traband, Anastasia; Lau, Marisa K; Adamus, Grazyna; Avery, Robert A

    2017-05-01

    The diagnostic path presented narrows down the cause of acute vision loss to the cone photoreceptor outer segment and will refocus the search for the cause of similar currently idiopathic conditions. To describe the structural and functional associations found in a patient with acute zonal occult photoreceptor loss. A case report of an adolescent boy with acute visual field loss despite a normal fundus examination performed at a university teaching hospital. Results of a complete ophthalmic examination, full-field flash electroretinography (ERG) and multifocal ERG, light-adapted achromatic and 2-color dark-adapted perimetry, and microperimetry. Imaging was performed with spectral-domain optical coherence tomography (SD-OCT), near-infrared (NIR) and short-wavelength (SW) fundus autofluorescence (FAF), and NIR reflectance (REF). The patient was evaluated within a week of the onset of a scotoma in the nasal field of his left eye. Visual acuity was 20/20 OU, and color vision was normal in both eyes. Results of the fundus examination and of SW-FAF and NIR-FAF imaging were normal in both eyes, whereas NIR-REF imaging showed a region of hyporeflectance temporal to the fovea that corresponded with a dense relative scotoma noted on light-adapted static perimetry in the left eye. Loss in the photoreceptor outer segment detected by SD-OCT co-localized with an area of dense cone dysfunction detected on light-adapted perimetry and multifocal ERG but with near-normal rod-mediated vision according to results of 2-color dark-adapted perimetry. Full-field flash ERG findings were normal in both eyes. The outer nuclear layer and inner retinal thicknesses were normal. Localized, isolated cone dysfunction may represent the earliest photoreceptor abnormality or a distinct entity within the acute zonal occult outer retinopathy complex. Acute zonal occult outer retinopathy should be considered in patients with acute vision loss and abnormalities on NIR-REF imaging, especially if

  2. Tube spacer grid for a heat-exchanger tube bundle

    International Nuclear Information System (INIS)

    Scheidl, H.

    1976-01-01

    A tube spacer grid for a heat-exchanger tube bundle is formed by an annular grid frame having a groove formed in its inner surface in which the interspaced grid bars have their ends positioned and held in interspaced relationship by short sections of tubes passed through holes axially formed in the grid frame so that the tubes are positioned between the ends of the grid bars in the grooves. The tube sections may be cut from the same tubes used to form the tube bundle. 5 claims, 3 drawing figures

  3. Design and economic optimization of shell and tube heat exchangers using Artificial Bee Colony (ABC) algorithm

    International Nuclear Information System (INIS)

    Sencan Sahin, Arzu; Kilic, Bayram; Kilic, Ulas

    2011-01-01

    Highlights: → Artificial Bee Colony for shell and tube heat exchanger optimization is used. → The total cost is minimized by varying design variables. → This new approach can be applied for optimization of heat exchangers. - Abstract: In this study, a new shell and tube heat exchanger optimization design approach is developed. Artificial Bee Colony (ABC) has been applied to minimize the total cost of the equipment including capital investment and the sum of discounted annual energy expenditures related to pumping of shell and tube heat exchanger by varying various design variables such as tube length, tube outer diameter, pitch size, baffle spacing, etc. Finally, the results are compared to those obtained by literature approaches. The obtained results indicate that Artificial Bee Colony (ABC) algorithm can be successfully applied for optimal design of shell and tube heat exchangers.

  4. Design and economic optimization of shell and tube heat exchangers using Artificial Bee Colony (ABC) algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Sencan Sahin, Arzu, E-mail: sencan@tef.sdu.edu.tr [Department of Mechanical Education, Technical Education Faculty, Sueleyman Demirel University, 32260 Isparta (Turkey); Kilic, Bayram, E-mail: bayramkilic@hotmail.com [Bucak Emin Guelmez Vocational School, Mehmet Akif Ersoy University, Bucak (Turkey); Kilic, Ulas, E-mail: ulaskilic@mehmetakif.edu.tr [Bucak Emin Guelmez Vocational School, Mehmet Akif Ersoy University, Bucak (Turkey)

    2011-10-15

    Highlights: {yields} Artificial Bee Colony for shell and tube heat exchanger optimization is used. {yields} The total cost is minimized by varying design variables. {yields} This new approach can be applied for optimization of heat exchangers. - Abstract: In this study, a new shell and tube heat exchanger optimization design approach is developed. Artificial Bee Colony (ABC) has been applied to minimize the total cost of the equipment including capital investment and the sum of discounted annual energy expenditures related to pumping of shell and tube heat exchanger by varying various design variables such as tube length, tube outer diameter, pitch size, baffle spacing, etc. Finally, the results are compared to those obtained by literature approaches. The obtained results indicate that Artificial Bee Colony (ABC) algorithm can be successfully applied for optimal design of shell and tube heat exchangers.

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

    International Nuclear Information System (INIS)

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

    1994-09-01

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

  6. Commissioning of the 4 K Outer Cryostat for the CUORE Experiment

    CERN Document Server

    Ferri, E; Biassoni, M; Bucci, C; Ceruti, G; Chiarini, A; Clemenza, M; Cremonesi, O; Datskov, V; Dossena, S; Faverzani, M; Franceschi, M A; Gaigher, R; Gorla, P; Guetti, M; Ligi, C; Napolitano, T; Nucciotti, A; Pelosi, A; Perego, M; Previtali, E; Sisti, M; Taffarello, L; Terranova, F

    2014-01-01

    The Cryogenic Underground Observatory for Rare Events (CUORE) is a 1-ton scale bolometric experiment. The CUORE detector is an array of 988 TeO crystals arranged in a cylindrical, compact, and granular structure of 19 towers. These detectors will need a base temperature lower than 10 mK in order to meet the performance specifications. To cool the CUORE detector, a large cryogen free cryostat with five pulse tubes and one custom designed high power dilution refrigerator has been designed. The three vessels that form the outer shell of the CUORE cryostat were produced in 2012 and are now assembled in the Gran Sasso National Laboratories (LNGS). We report here the detailed description of the 4 K outer cryostat for the CUORE experiment together with the results of the validation tests done at the production site in 2012 and of the first commissioning to 4 K at LNGS in 2013.

  7. A comparison of R-22, R-134a, R-410a, and R-407c condensation performance in smooth and enhanced tubes: Part 1, Heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Eckels, S J; Tesene, B A

    1999-07-01

    Local and average heat transfer coefficients during condensation are reported for R-22, R-134a, R-410a, and R-407c in one smooth tube and three enhanced surface tubes. The test tubes included a 3/8 inch outer diameter smooth tube, a 3/8 inch outer diameter microfin tube, a 5/16 inch outer diameter microfin tube, and a 5/8 inch outer diameter microfin tube. The local and average heat transfer coefficients were measured over a mass flux range of 92,100 lb/ft{sup 2}{center_dot}h to 442,200 lb/ft{sup 2}{center_dot}h and at saturation temperatures of 104 F and 122 F. A comparison of the performance of the different refrigerants reveals that R-134a has the highest heat transfer performance followed by R-22 and R-410a, which have similar performances. In general, R-407c had the lowest performance of the refrigerants tested. The microfin tube more than doubles the heat transfer coefficient compared to the smooth tube for all refrigerants at the low mass fluxes, but only increases the heat transfer coefficients by 50% at the highest mass flux tested. The measured heat transfer coefficients are also compared with a number of correlations for condensation.

  8. 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.

  9. 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

  10. Free electron laser amplifier experiments on SG-1

    International Nuclear Information System (INIS)

    Hui Zhongxi; Zhou Chuanming; Wu Ruian

    1994-01-01

    The SG-1 FEL facility is composed of a linear induction accelerator (LIA), an electron beam transport system, a wiggler, a microwave source and a diagnostic system. SG-1 LIA provides a 2 kA, 3.0 MeV beam with a normalized emittance of 0.4∼0.6 (π rad·cm), an energy spread (FWHM) of 4%, resulting in a beam brightness of nearly 10 8 A/πm·rad) 2[1] . The beam current through the wiggler is about 600 A. The first ASE experiments began in September 1991. A 2.6-m long wiggler with a peak magnetic field of 0.3 T was used. At 35.8∼36.5 GHz an ASE output of 0.5 W was obtained for a beam current of nearly 50 A. After a shutdown of about 8 months, the second series of ASE experiments began in October 1992. The second series of ASE experiments were performed with a wiggler magnetic field between 0.25∼0.27 T. The maximum output power is about 100 kw for B w = 0.24 T, I = 600 A, At ν = 35.2 GHz. Based on the ASE experiments the amplifier experiments was carried out on SG-1. Using an 300 W input signal (TE 01 ), a beam current of about 600 A and wiggler magnetic fields of 0.24∼0.28 T, the authors measured the FEL output power as a function of the wiggler magnetic field. The resonant magnetic field was about 0.25 T. Meanwhile, in order to study the amplifier gain, the authors measured the FEL output power as a function of the wiggler length at a peak wiggler magnetic field of 0.26 T. The exponential gain is approximately 19 dB/m and the maximum output power is about 10 MW

  11. 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

  12. 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)

  13. The Outer Space as an Educational Motivation

    Science.gov (United States)

    Pérez-Pérez, Melquíades; Hernández-López, Montserrat

    2017-06-01

    STEAM is an educational approach to learning that uses Science, Technology, Engineering, the Arts and Mathematics as access points for guiding student inquiry, dialogue, and critical thinking. The end results are students who take thoughtful risks, engage in experiential learning, persist in problem-solving, embrace collaboration, and work through the creative process. The Outer Space is a window to the past and the future of our travel around the history of the Universe and can be used as a educational tool in primary and secondary education. This paper talks about the integration of the resources of European Space Agency, Space Awareness, Nuclio, Scientix and Schoolnet as motivation to integrate STEAM methodology in secondary education. Keywords: STEAM, outer space, motivation, methodology

  14. The Outer Banks of North Carolina

    Science.gov (United States)

    Dolan, Robert; Lins, Harry F.; Smith, Jodi Jones

    2016-12-27

    The Outer Banks of North Carolina are excellent examples of the nearly 300 barrier islands rimming the Atlantic and Gulf coasts of the United States. These low, sandy islands are among the most dynamic natural landscapes occupied by man. Beach sands move offshore, onshore, and along the shore in the direction of the prevailing longshore currents. In this way, sandy coasts continuously adjust to different tide, wave, and current conditions and to rising sea level that causes the islands to migrate landward.Despite such changes, barrier islands are of considerable environmental importance. The Outer Banks are home to diverse natural ecosystems that are adapted to the harsh coastal environment. Native species tend to be robust and many are specifically adapted to withstand salt spray, periodic saltwater flooding, and the islands’ well-drained sandy soil. The Outer Banks provide an important stopover for birds on the Atlantic flyway, and many species inhabit the islands year round. In addition, Outer Banks beaches provide an important nesting habitat for five endangered or threatened sea turtle species.European explorers discovered North Carolina’s barrier islands in the 16th century, although the islands were not permanently settled until the middle 17th century. By the early 19th century, shipbuilding and lumber industries were among the most successful, until forest resources were depleted. Commercial fishing eventually followed, and it expanded considerably after the Civil War. By the Great Depression, however, little industry existed on the Outer Banks. In response to the effects of a severe hurricane in 1933, the National Park Service and the Civilian Conservation Corps proposed a massive sand-fixation program to stabilize the moving sand and prevent storm waves from sweeping across the entire width of some sections of the islands. Between 1933 and 1940, this program constructed sand fencing on 185 kilometers (115 miles) of beach and planted grass seedlings

  15. Cryovolcanism in the outer solar system

    Science.gov (United States)

    Geissler, Paul E.

    2015-01-01

    Cryovolcanism is defined as the extrusion of liquids and vapors of materials that would be frozen solid at the planetary surface temperatures of the icy bodies of the outer solar system. Active cryovolcanism is now known to occur on Saturn's moon Enceladus and on Neptune's moon Triton and is suspected on Jupiter's moon Europa, while evidence for past cryovolcanic activity is widespread throughout the outer solar system. This chapter examines the mechanisms and manifestations of cryovolcanism, beginning with a review of the materials that make up these unusual ‘‘magmas’’ and the means by which they might erupt and concluding with a volcanologist's tour of the farthest reaches of the solar system.

  16. Protection of nuclear facilities against outer aggressions

    International Nuclear Information System (INIS)

    Aussourd, P.; Candes, P.; Le Quinio, R.

    1976-01-01

    The various types of outer aggressions envisaged in safety analysis for nuclear facilities are reviewed. These outer aggressions are classified as natural and non-natural phenomena, the latter depending on the human activities in the vicinity of nuclear sites. The principal natural phenomena able to constitute aggressions are atmospheric phenomena (strong winds, snow storms, hail, frosting mists), hydrologie phenomena such as tides, surges, flood, low waters, and geologic phenomena such as earthquakes. Artificial phenomena are concerned with aircraft crashes, projectiles, fire, possible ruptures of dams, and intentional human aggressions. The protection against intentional human aggressions is of two sorts: first, the possibility of access to the installations mostly sensitive to sabotage are to be prevented or reduced, secondly redundant circuits and functions must be separated for preventing their simultaneous destruction in the case when sabotage actors have reach the core of the facility [fr

  17. Outer crust of nonaccreting cold neutron stars

    International Nuclear Information System (INIS)

    Ruester, Stefan B.; Hempel, Matthias; Schaffner-Bielich, Juergen

    2006-01-01

    The properties of the outer crust of nonaccreting cold neutron stars are studied by using modern nuclear data and theoretical mass tables, updating in particular the classic work of Baym, Pethick, and Sutherland. Experimental data from the atomic mass table from Audi, Wapstra, and Thibault of 2003 are used and a thorough comparison of many modern theoretical nuclear models, both relativistic and nonrelativistic, is performed for the first time. In addition, the influences of pairing and deformation are investigated. State-of-the-art theoretical nuclear mass tables are compared to check their differences concerning the neutron drip line, magic neutron numbers, the equation of state, and the sequence of neutron-rich nuclei up to the drip line in the outer crust of nonaccreting cold neutron stars

  18. Boundary layers of the earth's outer magnetosphere

    Science.gov (United States)

    Eastman, T. E.; Frank, L. A.

    1984-01-01

    The magnetospheric boundary layer and the plasma-sheet boundary layer are the primary boundary layers of the earth's outer magnetosphere. Recent satellite observations indicate that they provide for more than 50 percent of the plasma and energy transport in the outer magnetosphere although they constitute less than 5 percent by volume. Relative to the energy density in the source regions, plasma in the magnetospheric boundary layer is predominantly deenergized whereas plasma in the plasma-sheet boundary layer has been accelerated. The reconnection hypothesis continues to provide a useful framework for comparing data sampled in the highly dynamic magnetospheric environment. Observations of 'flux transfer events' and other detailed features near the boundaries have been recently interpreted in terms of nonsteady-state reconnection. Alternative hypotheses are also being investigated. More work needs to be done, both in theory and observation, to determine whether reconnection actually occurs in the magnetosphere and, if so, whether it is important for overall magnetospheric dynamics.

  19. Boundary layers of the earth's outer magnetosphere

    International Nuclear Information System (INIS)

    Eastman, T.E.; Frank, L.A.

    1984-01-01

    The magnetospheric boundary layer and the plasma-sheet boundary layer are the primary boundary layers of the earth's outer magnetosphere. Recent satellite observations indicate that they provide for more than 50 percent of the plasma and energy transport in the outer magnetosphere although they constitute less than 5 percent by volume. Relative to the energy density in the source regions, plasma in the magnetospheric boundary layer is predominantly deenergized whereas plasma in the plasma-sheet boundary layer has been accelerated. The reconnection hypothesis continues to provide a useful framework for comparing data sampled in the highly dynamic magnetospheric environment. Observations of flux transfer events and other detailed features near the boundaries have been recently interpreted in terms of nonsteady-state reconnection. Alternative hypotheses are also being investigated. More work needs to be done, both in theory and observation, to determine whether reconnection actually occurs in the magnetosphere and, if so, whether it is important for overall magnetospheric dynamics. 30 references

  20. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... search for current job openings visit HHS USAJobs Home >> NEI YouTube Videos >> NEI YouTube Videos: Amblyopia Listen NEI YouTube Videos YouTube Videos Home Age-Related Macular Degeneration Amblyopia Animations Blindness Cataract ...

  1. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... search for current job openings visit HHS USAJobs Home » NEI YouTube Videos » NEI YouTube Videos: Amblyopia Listen NEI YouTube Videos YouTube Videos Home Age-Related Macular Degeneration Amblyopia Animations Blindness Cataract ...

  2. Minimizing shell-and-tube heat exchanger cost with genetic algorithms and considering maintenance

    Energy Technology Data Exchange (ETDEWEB)

    Wildi-Tremblay, P.; Gosselin, L. [Universite Laval, Quebec (Canada). Dept. de genie mecanique

    2007-07-15

    This paper presents a procedure for minimizing the cost of a shell-and-tube heat exchanger based on genetic algorithms (GA). The global cost includes the operating cost (pumping power) and the initial cost expressed in terms of annuities. Eleven design variables associated with shell-and-tube heat exchanger geometries are considered: tube pitch, tube layout patterns, number of tube passes, baffle spacing at the centre, baffle spacing at the inlet and outlet, baffle cut, tube-to-baffle diametrical clearance, shell-to-baffle diametrical clearance, tube bundle outer diameter, shell diameter, and tube outer diameter. Evaluations of the heat exchangers performances are based on an adapted version of the Bell-Delaware method. Pressure drops constraints are included in the procedure. Reliability and maintenance due to fouling are taken into account by restraining the coefficient of increase of surface into a given interval. Two case studies are presented. Results show that the procedure can properly and rapidly identify the optimal design for a specified heat transfer process. (author)

  3. Differential Rotation within the Earth's Outer Core

    Science.gov (United States)

    Hide, R.; Boggs, D. H.; Dickey, J. O.

    1998-01-01

    Non-steady differential rotation drive by bouyancy forces within the Earth's liquid outer core (OC) plays a key role not only in the generation of the main geomagnetic field by the magnetohydrodynamic (MHD) dynamo process but also in the excitation of irregular fluctuations in the angular speed of rotation of the overlying solid mantle, as evidenced by changes in the length of the day (LOD) on decadal and longer timescales (1-8).

  4. Fluxgate magnetometers for outer planets exploration

    Science.gov (United States)

    Acuna, M. H.

    1974-01-01

    The exploration of the interplanetary medium and the magnetospheres of the outer planets requires the implementation of magnetic field measuring instrumentation with wide dynamic range, high stability, and reliability. The fluxgate magnetometers developed for the Pioneer 11 and Mariner-Jupiter-Saturn missions are presented. These instruments cover the range of .01 nT to 2 million nT with optimum performance characteristics and low power consumption.

  5. Outer Limits of Biotechnologies: A Jewish Perspective

    Directory of Open Access Journals (Sweden)

    John D. Loike

    2018-01-01

    Full Text Available A great deal of biomedical research focuses on new biotechnologies such as gene editing, stem cell biology, and reproductive medicine, which have created a scientific revolution. While the potential medical benefits of this research may be far-reaching, ethical issues related to non-medical applications of these technologies are demanding. We analyze, from a Jewish legal perspective, some of the ethical conundrums that society faces in pushing the outer limits in researching these new biotechnologies.

  6. Performance analysis of double basin solar still with evacuated tubes

    International Nuclear Information System (INIS)

    Hitesh N Panchal; Shah, P. K.

    2013-01-01

    Solar still is a very simple device, which is used for solar distillation process. In this research work, double basin solar still is made from locally available materials. Double basin solar still is made in such a way that, outer basin is exposed to sun and lower side of inner basin is directly connected with evacuated tubes to increase distillate output and reducing heat losses of a solar still. The overall size of the lower basin is about 1006 mm x 325 mm x 380 mm, the outer basin is about 1006 mm x 536 mm x 100 mm Black granite gravel is used to increase distillate output by reducing quantity of brackish or saline water in the both basins. Several experiments have conducted to determine the performance of a solar still in climate conditions of Mehsana (latitude of 23 degree 59' and longitude of 72 degree 38'), Gujarat, like a double basin solar still alone, double basin solar still with different size black granite gravel, double basin solar still with evacuated tubes and double basin solar still with evacuated tubes and different size black granite gravel. Experimental results show that, connecting evacuated tubes with the lower side of the inner basin increases daily distillate output of 56% and is increased by 60%, 63% and 67% with average 10 mm, 20 mm and 30 mm size black granite gravel. Economic analysis of present double basin solar still is 195 days. (authors)

  7. Computational Fluid Dynamics (CFD) simulations of a Heisenberg Vortex Tube

    Science.gov (United States)

    Bunge, Carl; Sitaraman, Hariswaran; Leachman, Jake

    2017-11-01

    A 3D Computational Fluid Dynamics (CFD) simulation of a Heisenberg Vortex Tube (HVT) is performed to estimate cooling potential with cryogenic hydrogen. The main mechanism driving operation of the vortex tube is the use of fluid power for enthalpy streaming in a highly turbulent swirl in a dual-outlet tube. This enthalpy streaming creates a temperature separation between the outer and inner regions of the flow. Use of a catalyst on the peripheral wall of the centrifuge enables endothermic conversion of para-ortho hydrogen to aid primary cooling. A κ- ɛ turbulence model is used with a cryogenic, non-ideal equation of state, and para-orthohydrogen species evolution. The simulations are validated with experiments and strategies for parametric optimization of this device are presented.

  8. Method of making a composite tube to metal joint

    Energy Technology Data Exchange (ETDEWEB)

    Leslie, James C.; Leslie, II, James C.; Heard, James; Truong, Liem V.; Josephson, Marvin

    2017-11-07

    A method for making a metal to composite tube joint including selecting an elongated interior fitting constructed with an exterior barrel, reduced in exterior diameter to form a distally facing annular shoulder and then projecting still further distally to form an interior sleeve having a radially outwardly facing bonding surface. Selecting an elongated metal outer sleeve formed proximally with a collar constructed for receipt over the barrel and increased in interior diameter and projecting distally to form an exterior sleeve having a radially inwardly facing bonding surface cooperating with the first bonding surface to form an annulus receiving an extremity of a composite tube and a bond bonding the extremity of the tube to the bonding surfaces.

  9. Eddy current inspection of weld defects in tubing

    Science.gov (United States)

    Katragadda, G.; Lord, W.

    1992-01-01

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

  10. Development of an innovative plate-type SG for fast breeder reactor. Proposal of the concept and the evaluation of the fabricating method by the test fabrication of the partial model

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Kinoshita, Izumi

    2006-01-01

    The concept of an innovative plate type SG for the fast reactor fabricated by using the HIP (Hot Isostatic Pressing) method was proposed. The heat transfer plate, which is assembled with rectangular tubes and is fabricated by HIP method, is surrounded by leakage detection spaces. It is possible to apply it to both the pool-type and the loop-type LMFR. In this report, the fabrication technique was studied about the concept for the loop-type LMFR, and the following results were obtained. Hip tests, tensile tests, and structure observation were performed to clarify the suitable HIP condition for the modified 9Cr-1Mo steel. As a result, the optimum condition of 1150 deg C x 1200 kgf/cm 2 x 3 hr was found. Nickel-type solder (BNi-5) and gold-type solder (BAu-4) were selected as a joining material to laminate the heat transfer tube plates. Through the comparison of tensile tests, BAu-4 that showed a more excellent joining performance was selected on the assumption of the margin of 5 mm from the welding line. After buckling load had been clarified, the BAu-4 brazing of the heat transfer tube plates was performed using a hot pressing method. Problems were not observed in the welding of simulated header, and in the fabricating of the partial model of SG. (author)

  11. Gel spinning of silk tubes for tissue engineering

    Science.gov (United States)

    Lovett, Michael; Cannizzaro, Christopher; Vunjak-Novakovic, Gordana; Kaplan, David L.

    2011-01-01

    Tubular vessels for tissue engineering are typically fabricated using a molding, dipping, or electrospinning technique. While these techniques provide some control over inner and outer diameters of the tube, they lack the ability to align the polymers or fibers of interest throughout the tube. This is an important aspect of biomaterial composite structure and function for mechanical and biological impact of tissue outcomes. We present a novel aqueous process system to spin tubes from biopolymers and proteins such as silk fibroin. Using silk as an example, this method of winding an aqueous solution around a reciprocating rotating mandrel offers substantial improvement in the control of the tube properties, specifically with regard to winding pattern, tube porosity, and composite features. Silk tube properties are further controlled via different post-spinning processing mechanisms such as methanol-treatment, air-drying, and lyophilization. This approach to tubular scaffold manufacture offers numerous tissue engineering applications such as complex composite biomaterial matrices, blood vessel grafts and nerve guides, among others. PMID:18801570

  12. Tube plug removal machine

    International Nuclear Information System (INIS)

    Hawkins, P.J.

    1987-01-01

    In a nuclear steam generator wherein some faulty tubes have been isolated by mechanical plugging, to remove a selected plug without damaging the associated tube, a plug removal machine is used. The machine drills into a plug portion with a tap drill bit having a drill portion a tap portion and a threaded portion, engaging that plug portion with the threaded portion after the drilled hole has been threaded by the tap portion thereof, and removing a portion of the plug in the tube with a counterbore drill bit mounted concentrically about the tap drill bit. A trip pin and trip spline disengage the tap drill bit from the motor. The counterbore drill bit is thereafter self-centered with respect to the tube and plug about the now stationary tap drill bit. After a portion of the plug has been removed by the counterbore drill bit, pulling on the top drill bit by grippers on slots will remove the remaining plug portion from the tube. (author)

  13. Categorising YouTube

    Directory of Open Access Journals (Sweden)

    Thomas Mosebo Simonsen

    2011-09-01

    Full Text Available This article provides a genre analytical approach to creating a typology of the User Generated Content (UGC of YouTube. The article investigates the construction of navigationprocesses on the YouTube website. It suggests a pragmatic genre approach that is expanded through a focus on YouTube’s technological affordances. Through an analysis of the different pragmatic contexts of YouTube, it is argued that a taxonomic understanding of YouTube must be analysed in regards to the vacillation of a user-driven bottom-up folksonomy and a hierarchical browsing system that emphasises a culture of competition and which favours the already popular content of YouTube. With this taxonomic approach, the UGC videos are registered and analysed in terms of empirically based observations. The article identifies various UGC categories and their principal characteristics. Furthermore, general tendencies of the UGC within the interacting relationship of new and old genres are discussed. It is argued that the utility of a conventional categorical system is primarily of analytical and theoretical interest rather than as a practical instrument.

  14. Investigation of the integrity of u-bend tube bundles subjected to flow-induced vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, M. [University of Guelph, Guelph, Ontario (Canada); Riznic, J. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

    2012-07-01

    Maintaining the integrity of nuclear steam generator (SG) tubes in CANDU reactors is a major safety issue since they maintain the physical barrier between the primary and secondary coolants. The integrity of these tubes can be compromised due to flow-induced vibrations in the form of fatigue and fretting wear damage. Wear is a result of the tube impacting and sliding against its loose supports, and it becomes more severe as the tube/support clearance increases. The vibration is caused by fluid flow around these tubes through turbulence and fluidelastic instability mechanisms. Supports are installed to stiffen the structure and to ensure safe and stable operation. The U-bend region is the most critical part since it is subjected to high cross flow. Therefore, special attention is paid to properly supporting this region. However, in some situations, tube support plates (TSP) located on the straight part of the tube may deteriorate to the point where extremely large clearances, or even total wastage of the supports, may result. One possible cause for such a situation is corrosion and/or excessive fretting wear. This loss of TSP may affect the rate of wear in the U-bend portion of the tube due to the increased flexibility in this region. The integrity could be seriously breached as result of a potential support loss. This paper addresses the flow-induced vibrations (FIV) aspect, consequences, and suggested remedies for support degradation. This analysis will include fretting wear producing parameters, such as impact force and normal work rate. Turbulence and fluidelastic instability (FEI) are considered to be the main excitation mechanisms. The investigation is conducted through a numerical simulation of the full Ubend tube bundles including modelling the variable flow distribution, flow excitation, impact, and friction at the supports. (author)

  15. The Sg-1 Glycosyltransferase Locus Regulates Structural Diversity of Triterpenoid Saponins of Soybean[W][OA

    Science.gov (United States)

    Sayama, Takashi; Ono, Eiichiro; Takagi, Kyoko; Takada, Yoshitake; Horikawa, Manabu; Nakamoto, Yumi; Hirose, Aya; Sasama, Hiroko; Ohashi, Mihoko; Hasegawa, Hisakazu; Terakawa, Teruhiko; Kikuchi, Akio; Kato, Shin; Tatsuzaki, Nana; Tsukamoto, Chigen; Ishimoto, Masao

    2012-01-01

    Triterpene saponins are a diverse group of biologically functional products in plants. Saponins usually are glycosylated, which gives rise to a wide diversity of structures and functions. In the group A saponins of soybean (Glycine max), differences in the terminal sugar species located on the C-22 sugar chain of an aglycone core, soyasapogenol A, were observed to be under genetic control. Further genetic analyses and mapping revealed that the structural diversity of glycosylation was determined by multiple alleles of a single locus, Sg-1, and led to identification of a UDP-sugar–dependent glycosyltransferase gene (Glyma07g38460). Although their sequences are highly similar and both glycosylate the nonacetylated saponin A0-αg, the Sg-1a allele encodes the xylosyltransferase UGT73F4, whereas Sg-1b encodes the glucosyltransferase UGT73F2. Homology models and site-directed mutagenesis analyses showed that Ser-138 in Sg-1a and Gly-138 in Sg-1b proteins are crucial residues for their respective sugar donor specificities. Transgenic complementation tests followed by recombinant enzyme assays in vitro demonstrated that sg-10 is a loss-of-function allele of Sg-1. Considering that the terminal sugar species in the group A saponins are responsible for the strong bitterness and astringent aftertastes of soybean seeds, our findings herein provide useful tools to improve commercial properties of soybean products. PMID:22611180

  16. 75 FR 1076 - Outer Continental Shelf Civil Penalties

    Science.gov (United States)

    2010-01-08

    ... initiate civil penalty proceedings; however, violations that cause injury, death, or environmental damage... DEPARTMENT OF THE INTERIOR Minerals Management Service Outer Continental Shelf Civil Penalties... daily civil penalty assessment. SUMMARY: The Outer Continental Shelf Lands Act requires the MMS to...

  17. Advanced evacuated tube collectors

    Science.gov (United States)

    Schertz, W. W.; Hull, J. R.; Winston, R.; Ogallagher, J.

    1985-04-01

    The essence of the design concept for these new collectors is the integration of moderate levels of nonimaging concentration inside the evacuated tube itself. This permanently protects the reflection surfaces and allows the use of highly reflecting front surface mirrors with reflectances greater than 95%. Previous fabrication and long term testing of a proof-of-concept prototype has established the technical success of the concept. Present work is directed toward the development of a manufacturable unit that will be suitable for the widest possible range of applications. Design alternatives include scaling up the original prototype's tube diameter from 5 cm to 10 cm, using an internal shaped metal concentrating reflector, using a variety of profile shapes to minimize so-called gap losses and accommodate both single ended and double-ended flow geometries, and allowing the use of heat pipes for the absorber tube.

  18. Square through tube

    International Nuclear Information System (INIS)

    Akita, Junji; Honma, Toei.

    1975-01-01

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

  19. Thermal behaviour of pressure tube under fully and partially voided heating conditions using 19 pin fuel element simulator

    International Nuclear Information System (INIS)

    Yadav, Ashwini K.; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, B.; Mukhopadhya, D.; Lele, H.G.

    2011-01-01

    In a nuclear reactor temperature can rise drastically during LOCA due to failure of heat transportation system and subsequently leads to mechanical deformations like sagging, ballooning and breaching of pressure tube. To understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of 220 MWe Indian Pressurised Heavy Water Reactor (IPHWR). The symmetrical heating of pressure tube of 1 m length was done through resistance heating of 19 pins under 13.5 kW power using a rectifier and the variation of temperatures over the circumference of pressure tube (PT), calandria tube (CT) and clad tubes were measured. The sagging of pressure tube was initiated at 460 deg C temperature and highest temperature attained was 650 deg C. The highest temperature attained by clad tubes was 680 deg C (over outer ring) and heat is dissipated to calandria vessel mainly due to radiation and natural convection. Again to simulate partially voided conditions, asymmetrical heating of pressure was carried out by injecting 8 kW power to upper 8 pins of fuel simulator. A maximum temperature difference of 295 deg C was observed over the circumference of pressure tube which highlights the magnitude of thermal stresses and its role in breaching of pressure tube under partially voided conditions. Integrity of pressure tube was retained during both symmetrical and asymmetrical heatup conditions. (author)

  20. Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

    2008-01-01

    For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

  1. PRODUCTION OF URANIUM TUBING

    Science.gov (United States)

    Creutz, E.C.

    1958-04-15

    The manufacture of thin-walled uranium tubing by the hot-piercing techique is described. Uranium billets are preheated to a temperature above 780 d C. The heated billet is fed to a station where it is engaged on its external surface by three convex-surfaced rotating rollers which are set at an angle to the axis of the billet to produce a surface friction force in one direction to force the billet over a piercing mandrel. While being formed around the mandrel and before losing the desired shape, the tube thus formed is cooled by a water spray.

  2. Guide tube sleeve

    International Nuclear Information System (INIS)

    Attix, D.J.

    1983-01-01

    The invention increases the operating capacity of a nuclear reactor by causing a modification in the flow pattern of the coolant which enhances the coolant's effectiveness. The apparatus provides a thin-walled tubular sleeve closely surrounding but not attached to the exterior surface of a guide tube in a fuel assembly. The wall of the sleeve has tabs projecting outwardly into adjacent flow channels. The sleeve is attached to the wall of a cellular void through which passes the guide tube associated with said sleeve. The tabs increase the flow of water in the channel and thus increase the heat transfer

  3. New insights into controlling tube-bundle fouling using alternative amines

    International Nuclear Information System (INIS)

    Turner, C.W.; Klimas, S.J.; Guzonas, D.A.; Frattini, P.L.; Fruzzetti, K.

    2002-01-01

    A volatile amine is added to the secondary heat-transport system of a nuclear power plant to reduce the rate of corrosion and corrosion product transport in the feedwater and to protect steam generator (SG) crevices and materials exposed to steam condensate. Volatility and base strength of the amine at the SG operating temperature are two important considerations when choosing the optimum amine (or mixture of amines) for corrosion control in the steam cycle. Atomic Energy of Canada Limited (AECL) and Electric Power Research Institute (EPRI) have been collaborating in an extensive investigation of the effectiveness of amines at controlling the rate of tube-bundle fouling under SG operating conditions. Tests have been performed using a radiotracing technique in a high-temperature fouling loop facility at Chalk River Laboratories operated by AECL. This investigation has provided new insights into the role played by the amine in determining the rate of tube-bundle fouling in the SG. These insights are being used by AECL and EPRI to develop criteria for the selection of an amine that has optimum properties for both corrosion control and deposit control in the secondary heat transport system. The investigation has found that the rate of tube-bundle fouling is strongly dependent upon the surface chemistry of the corrosion products. For example, the fouling rates of fully oxidized iron oxides, such as hematite and lepidocrocite, are at least an order of magnitude greater than the fouling rate of magnetite under identical operating conditions. The difference is related to the sign of the surface charge on the corrosion products at temperature. The choice of amine for pH-control also influences the fouling rate. This was originally thought to be a surface-charge effect as well, but recent tests have suggested that it is related to the role that the amine plays in governing the rate of deposit consolidation on the heat-transfer surface. Amines that promote a high rate of

  4. Heat transfer performance of condenser tubes in an MSF desalination system

    International Nuclear Information System (INIS)

    Galal, T.; Kalendar, A.; Al Saftawi, A.; Zedan, M.

    2010-01-01

    The present research examines the amount of condensed fresh water off the outer-side surface of heat exchangers in an MSF system. The quantitative modeling of condensed water on the outer surface of comparable tubes, enhanced and plain, in a simulated MSF technique is investigated. An adapted simulation design on a test-rig facility, accounting for the condenser tubing in actual industrial desalination plate-form, is used with corrugated and smooth aluminum-brass material tubes 1100mm long and 23mm bore. A single phase flow of authentic brine water that typifies real fouling is utilized to simulate the actual environmental life of a multi-stage flashing desalination system, with coolant flow velocity 0.1 m/s in the two delineated types of condenser tubing. It is demonstrated that the condensate water amount from the specified enhanced tube is about 1.22 times the condensate water amount from the smooth tube, adaptive for 140 running hours under deliberated constrains. The topic covers a comparative analysis of thermal performance. Comparing results with fresh water confirm the effect of fouling on significantly lowering the value of the overall heat transfer coefficient versus time. Fouling resistance R f is reported with the critical coolant flow speed of 0.1 m/s. Comparison between the fouling resistance for both smooth and corrugated tubes versus time is performed. The fouling thermal resistance of the corrugated tube is 0.56 of the fouling thermal resistance of the smooth tube after140 running hours of the experiment are concluded. Overall, in the case of real brine, results prove that heat performance for the corrugated tube is superior to the plain tube over the studied time period (140 hrs) for the chosen range of flow speeds

  5. Issues concerning outer space investments in international law ...

    African Journals Online (AJOL)

    Issues concerning outer space investments in international law. ... Recent improvements in technology have in essence increased the viability of outer space as the next frontier for international investment and development. In addition to ... Key words: Outer Space, Investments, International Law, International Space Station ...

  6. Probabilistic analysis of degradation incubation time of steam generator tubing materials

    International Nuclear Information System (INIS)

    Pandey, M.D.; Jyrkama, M.I.; Lu, Y.; Chi, L.

    2012-01-01

    The prediction of degradation free lifetime of steam generator (SG) tubing material is an important step in the life cycle management and decision for replacement of steam generators during the refurbishment of a nuclear station. Therefore, an extensive experimental research program has been undertaken by the Canadian Nuclear Industry to investigate the degradation of widely-used SG tubing alloys, namely, Alloy 600 TT, Alloy 690 TT, and Alloy 800. The corrosion related degradations of passive metals, such as pitting, crevice corrosion and stress corrosion cracking (SCC) etc. are assumed to start with the break down of the passive film at the tube-environment interface, which is characterized by the incubation time for passivity breakdown and then the degradation growth rate, and both are influenced by the chemical environment and coolant temperature. Since the incubation time and growth rate exhibit significant variability in the laboratory tests used to simulate these degradation processes, the use of probabilistic modeling is warranted. A pit is initiated with the breakdown of the passive film on the SG tubing surface. Upon exposure to aggressive environments, pitting corrosion may not initiate immediately, or may initiate and then re-passivate. The time required to initiate pitting corrosion is called the pitting incubation time, and that can be used to characterize the corrosion resistance of a material under specific test conditions. Pitting may be the precursor to other corrosion degradation mechanisms, such as environmentally-assisted cracking. This paper will provide an overview of the results of the first stage of experimental program in which samples of Alloy 600 TT, Alloy 690 TT, and Alloy 800 were tested under various temperatures and potentials and simulated crevice environments. The testing environment was chosen to represent layup, startup, and full operating conditions of the steam generators. Degradation incubation times for over 80 samples were

  7. Qualification of the Improved rotating probe process for steam generator tubes inspection

    International Nuclear Information System (INIS)

    Caston, D.

    2002-01-01

    In 1997, EDF called for bids to Eddy Current (EC) probes manufacturers to supply rotating probes in order to improve the inspection of the Roll Transition Zone of Steam Generator tubes. Several probes met EDF requirements, and after full assessment, EDF chose one between several EC rotating probe prototypes. For the state of its technical study, EDF chose CEGELEC NDTs services among French ISI SG NDT providers, to inspect a limited number of tubes on two French NPP in 2000 with this prototype. Improved Rotating Probe process technical requirements were provided by EDF with the SG contract specifications in June 2000. They dictate technique performances level and acquisition rate of this new process using two techniques at the same time: - STL classic technique applied for detection and sizing of axial cracks; - STT technique, applied for detection and Sizing of circumferential cracks and wear. It has to be used, instead of classic STL process, without increasing inspection duration and SG occupancy. In competition for the qualification, CEGELEC NDT decided to design a new probe with its providers, including the two EC sensors and meeting EDF's requirements. Two another new equipment, designed in CEGELEC NDT laboratories, have been integrated into Improved Rotating Probe Process: - 'STL Lift', new rotating probe push-puller for Roll Transition Zone inspection; - 'ANASTL', on-line STL and STT data quality check, on-line data processing and analysis software. Without talking about performances of the technique and results obtained on site, this paper presents the new equipment, the different phases of the qualification conducted according to RSE-M rules, the first field experiences in August 2001 and the feedback experience of following site inspections. (author)

  8. Dark matter in the outer solar system

    Science.gov (United States)

    Owen, T.; Cruikshank, D.; De Bergh, C.; Geballe, T.

    1994-01-01

    There are now a large number of small bodies in the outer solar system that are known to be covered with dark material. Attempts to identify that material have been thwarted by the absence of discrete absorption features in the reflection spectra of these planetesimals. An absorption at 2.2 micrometers that appeared to be present in several objects has not been confirmed by new observations. Three absorptions in the spectrum of the unusually red planetesimal 5145 Pholus are well-established, but their identity remains a mystery.

  9. Inner and Outer Life at Work

    DEFF Research Database (Denmark)

    Lundgaard Andersen, Linda

    2012-01-01

    involving people to people interactions offered by psychodynamic theories and methods take up a pivotal position. Psychoanalytic organisational and work life research explores how work, organisations and individuals are affected by psychic dynamics, the influence of the unconscious in the forms of human...... development and interaction situated in a societal context. Based on this substantial work I draw upon two influential psychoanalytical positions—the British Tavistock position and German psychoanalytic social psychology in order to situate and identify how to understand the inner and outer life at work...

  10. The urgency of outer territories anthropology

    Directory of Open Access Journals (Sweden)

    Miloš Milenković

    2016-02-01

    Full Text Available In the context of transforming a part of Serbian anthropology into social theoretic management of identity, I suggest both comparative historiographic and ethnographic learning from societies with similar post-colonial experience, with the aim to include the discipline into an urgent defense of Serbia and Belgrade from further ethno-profiteering interests of elites in/from outer territories, left over on the ruins of our ill judged, resource incompatible, exaggerated or immoral twentieth century adventures. Serbian anthropology, written by anthropologists to whom Serbia and Belgrade are "homeland" by origin or civilized choice, should play the key role in the defense of Serbian citizens from the interest of elites in/from the outer "homelands", particularly by revealing the processes for which it is, as a discipline, most expert at – the professionalization of ethnicity, interactive and hybrid nature of identity, instrumental nature of tradition and the identity politics in general. Having in mind the latest attempt, a particularly successful one, conducted by the end of the 20th and the beginning of the 21st century that the lives, health, well-being, dignity and future of persons born in and loyal to the interest of Serbia and Belgrade, in large scale, thoroughly and long term be sacrificed and dedicated to the interests of ethno-profiteering elites in/from outer territories, in this article I point to the possibility to, along with the comparative learning from the above mentioned post-colonial experiences, delicate experiences of urgent anthropology be applied as well as the rich tradition of collective research. This text analyzes the results of first such research, that represenst the initial, praiseworthy and a brave step in the wise striving to engage social sciences and humanities in a search of expert and not mythical/daily-political solutions of the key problem of the Serbian nation – that of how to settle the interests of the

  11. The control of stainless steel tubes and wires of small diameter by the Eddy current method

    International Nuclear Information System (INIS)

    Stossel, A.; Gallet, G.

    1978-01-01

    Stainless steel tubes and wires with an outer diameter greater than 1 mm were studied by Eddy currents. The sensor characteristics and the detection of defects in function of frequency are presented together with the results obtained in the detection of dimensional and metallurgical defects [fr

  12. Fast and Accurate Non-destructive Testing System for Inspection of Canning Tubes

    DEFF Research Database (Denmark)

    Gundtoft, Hans Erik; Nielsen, E.

    1973-01-01

    The authors describe the development of an inspection bench for the non-destructive examination of canning tubes. The bench is original in that the internal diameter is calculated from exact measurement of the outer diameter and the wall thickness. The transducers for inspection and control are r...

  13. IGA resistance of TT Alloy 690 and concentration behavior of Broached Egg Crate tube support configuration

    International Nuclear Information System (INIS)

    Suzuki, S.; Kusakabe, T.; Yamamoto, H.; Arioka, K.; Ochi, T.

    1992-01-01

    In order to improve the reliability of the Steam Generator (SG), TT Alloy 690 and BEC (Broached Egg Crate) type tube support plate has been developed. Some tests are carried out to heighten the reliability for these improvements all the more and the following results are obtained. (1) SERT test (Slow Extension Rate Test) made clear that TT690 has less IGA susceptibility in comparison with MA600. (2) The alkaline susceptibility on the occurrence of IGA/SCC on TT690 and MA600 obtained by SERT corresponds to that obtained by Model Boiler test. (3) By model boiler test, superior concentration behaviors for BEC type tube support plate configuration have been recognized in comparison with Drill type. This result is obtained by the joint research of the five utilities (Kansai Epco, Hokkaido Epco, Shikoku Epco, Kyushu Epco, JAPCO) and MHI

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. 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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  17. Inducement of IGA/SCC in Inconel 600 steam generator tubing during unit outages

    Energy Technology Data Exchange (ETDEWEB)

    Durance, D.; Sedman, K. [Bruce Power, Tiverton, Ontario (Canada); Roberts, J. [CANTECH Associates Ltd., Burlington, Ontario (Canada); King, P. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada); Gorman, J. [Dominion Engineering, Reston, VA (United States); Allen, R. [Kinectrics, Inc., Toronto, Ontario (Canada)

    2008-07-01

    The degradation of Unit 4 SG tubing by IGA/SCC has limited both the operating period and end of life predictions for Unit 4 since restart in late 2003. The circumferential IGA/SCC has been most significant in SG4 with substantial increases in both initiation and growth rates from 2005 through the spring of 2007. A detailed review of the occurrence of circumferential OD IGA/SCC at the RTZ in the HL TTS region of Bruce 4 steam generator tubes has led a conclusion that it is probable that the IGA/SCC has been the result of attack by partially reduced sulfur species such as tetrathionates and thiosulfates during periods of low temperature exposure. It is believed that attack of this type has mostly likely occurred during startup evolutions following outages as the result the development of aggressive reduced sulfur species in the TTS region during periods when the boilers were fully drained for maintenance activities. The modification of outage practices to limit secondary side oxygen ingress in the spring of 2007 has apparently arrested the degradation and has had significant affects on the allowable operating interval and end of life predictions for the entire unit. (author)

  18. Inducement of IGA/SCC in Inconel 600 steam generator tubing during unit outages

    International Nuclear Information System (INIS)

    Durance, D.; Sedman, K.; Roberts, J.; King, P.; Gorman, J.; Allen, R.

    2008-01-01

    The degradation of Unit 4 SG tubing by IGA/SCC has limited both the operating period and end of life predictions for Unit 4 since restart in late 2003. The circumferential IGA/SCC has been most significant in SG4 with substantial increases in both initiation and growth rates from 2005 through the spring of 2007. A detailed review of the occurrence of circumferential OD IGA/SCC at the RTZ in the HL TTS region of Bruce 4 steam generator tubes has led a conclusion that it is probable that the IGA/SCC has been the result of attack by partially reduced sulfur species such as tetrathionates and thiosulfates during periods of low temperature exposure. It is believed that attack of this type has mostly likely occurred during startup evolutions following outages as the result the development of aggressive reduced sulfur species in the TTS region during periods when the boilers were fully drained for maintenance activities. The modification of outage practices to limit secondary side oxygen ingress in the spring of 2007 has apparently arrested the degradation and has had significant affects on the allowable operating interval and end of life predictions for the entire unit. (author)

  19. Economic evaluation of maintenance strategies for steam generator tubes using probabilistic fracture mechanics and financial method

    International Nuclear Information System (INIS)

    Sagisaka, Mitsuyuki; Isobe, Yoshihiro; Yoshimura, Shinobu; Yagawa, Genki

    2004-01-01

    As an application of probabilistic fracture mechanics (PFM) and a financial method, risk-benefit analyses were performed for the purpose of optimizing maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). Parameters such as in-service inspection (ISI) detection accuracy, ISI interval, sampling inspection, replacement of SGs and stress corrosion cracking (SCC) allowance operation were selected for sensitivity analyses. In the analysis of the operation introducing maintenance criteria, the effect of quantitative accuracy of the inspection was also taken into account. Although the analyses were mainly conducted for SG tubes made of Inconel 600 mill anneal (MA) materials, the analyses were also performed for SCC-resistant materials with making assumptions on their crack initiation probabilities and crack propagation laws. To justify whether or not it is worth while implementing the selected maintenance strategies in terms of an economic point of view, net present value (NPV) was calculated as an index which is one of the most fundamental financial indices for decision-making based on the discounted cash flow (DCF) method. (author)

  20. 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.

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

    International Nuclear Information System (INIS)

    Keim, E.; Kastner, W.

    1994-01-01

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

  2. Tube-dwelling invertebrates

    NARCIS (Netherlands)

    Hölker, Franz; Vanni, Michael J.; Kuiper, Jan J.; Meile, Christof; Grossart, Hans Peter; Stief, Peter; Adrian, Rita; Lorke, Andreas; Dellwig, Olaf; Brand, Andreas; Hupfer, Michael; Mooij, Wolf M.; Nützmann, Gunnar; Lewandowski, Jörg

    2015-01-01

    There is ample evidence that tube-dwelling invertebrates such as chironomids significantly alter multiple important ecosystem functions, particularly in shallow lakes. Chironomids pump large water volumes, and associated suspended and dissolved substances, through the sediment and thereby compete

  3. Cladding tube manufacturing technology

    International Nuclear Information System (INIS)

    Hahn, R.; Jeong, Y. H.; Baek, B. J.; Kim, K. H.; Kim, S. J.; Choi, B. K.; Kim, J. M.

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report A lloy Development for High Burnup Cladding . Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs

  4. Thoughts on accelerator tubes

    International Nuclear Information System (INIS)

    Larson, J.D.

    1978-01-01

    A brief, subjective review is given of mechanisms that may be limiting electrostatic accelerator tubes to present levels of performance. Suggestions are made for attacking these limitations with the purpose of stimulating the thinking of designers and users of electrostatic accelerators

  5. Tracheostomy tube - speaking

    Science.gov (United States)

    ... with others. However, you can learn how to speak with a tracheostomy tube. It just takes practice. There ... If it is hard to speak with a trach in place, special devices can help you learn to create sounds. One-way valves, called speaking valves, are placed ...

  6. Thoughts of accelerator tubes

    International Nuclear Information System (INIS)

    Larson, J.D.

    1977-01-01

    A brief, subjective review is given of mechanisms that may be limiting electrostatic accelerator tubes to present levels of performance. Suggestions are made for attacking these limitations with the purpose of stimulating the thinking of designers and users of electrostatic accelerators

  7. Cladding tube manufacturing technology

    Energy Technology Data Exchange (ETDEWEB)

    Hahn, R. [Kraftwerk Union AG, Mulheim (Germany); Jeong, Y.H.; Baek, B.J.; Kim, K.H.; Kim, S.J.; Choi, B.K.; Kim, J.M. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-04-01

    This report gives an overview of the manufacturing routine of PWR cladding tubes. The routine essentially consists of a series of deformation and annealing processes which are necessary to transform the ingot geometry to tube dimensions. By changing shape, microstructure and structure-related properties are altered simultaneously. First, a short overview of the basics of that part of deformation geometry is given which is related to tube reducing operations. Then those processes of the manufacturing routine which change the microstructure are depicted, and the influence of certain process parameters on microstructure and material properties are shown. The influence of the resulting microstructure on material properties is not discussed in detail, since it is described in my previous report 'Alloy Development for High Burnup Cladding.' Because of their paramount importance still up to now, and because manufacturing data and their influence on properties for other alloys are not so well established or published, the descriptions are mostly related to Zry4 tube manufacturing, and are only in short for other alloys. (author). 9 refs., 46 figs.

  8. An experimental study of heat transfer characteristics of single and two-phase flows in an annular tube with external vibrations

    International Nuclear Information System (INIS)

    Zaki, Adel M.; Abou El-Kassem, S.K.; Abdalla Hanafi

    2003-01-01

    An experimental study of the external vibration effect on the heat transfer characteristics of single and two-phase flows in an annular tube is carried out. An experimental set-up was constructed to study the heat transfer in a stationary, as well as, in oscillating annular tube. The annular tube was heated electrically through the inner surface, which is a stainless steel tube (St 304) 13 mm outer diameter, while the outer tube, of 3.7 cm inner diameter, made from a glass. The experimental set-up was equipped with a vibrating system to excite the annular tube in the frequency range of 0 up to 134 Hz. Several sensors for measuring wall and fluid temperatures, heat fluxes and volume flow rates of both phases were used. The obtained results show that the heat transfer coefficient can be significantly increased by vibration of the test section. (author)

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

    International Nuclear Information System (INIS)

    Tonkovic, Zdenko; Skozrit, Ivica; Alfirevic, Ivo

    2008-01-01

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

  10. 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)

  11. Heating of the outer solar atmosphere

    International Nuclear Information System (INIS)

    Parker, E.N.

    1983-01-01

    The author discusses the idea that there must be a source of magnetic fields somewhere below the solar surface. He starts by considering present day ideas about the sun's internal structure. The sun has a radius of approximately 700,000 km, of which the outer 100,000 km or so is the convective zone, according to mixing-length models. The dynamo is believed to operate in the convective zone, across which there may be a 5-10% variation in the angular velocity. There are the stretched east-west fields similar to the ones in the earth's core. Associated with these are poloidal fields which contribute to a net dipole moment of the sun and are generated by a dynamo. The author shows that essentially no magnetic field configuration has an equilibrium; they dissipate quickly in spite of the high conductivity in fluid motions and heating. This is probably the major part of the heating of the sun's outer atmosphere. (Auth.)

  12. Cosmics in the LHCb Outer Tracker

    CERN Document Server

    Aaij, Roel

    2010-01-01

    The LHCb experiment at the Large Hadron Collider studies the decay of B mesons to test the description of CP violation in the Standard Model and to search for new physics. The decay $B_s \\to \\mu^+ \\mu^-$ has been identified as very promising in the search for new physics. An excellent invariant mass resolution is required to suppress backgrounds to this decay. This in turn requires a momentum resolution of dp/p = 0.4%. The Outer Tracker is part of the LHCb tracking system and has been commissioned with cosmic muons. The noise in the Outer Tracker is shown to be less than 0.05%. To use drift time information in the reconstruction of cosmic tracks, the event time must be known. Four methods to obtain the event time are studied and compared. It is shown that the event time can be obtained with a resolution better than 2.6 ns. Using drift time information, tracks are reconstructed with a resolution of 344 $\\mu$m. Knowledge of the event time enables the calibration of electronic time offsets and the r(t)– relati...

  13. Gamma rays from pulsar outer gaps

    International Nuclear Information System (INIS)

    Chiang, J.; Romani, R.W.; Cheng Ho

    1993-01-01

    We describe a gamma ray pulsar code which computes the high energy photon emissivities from vacuum gaps in the outer magnetosphere, after the model outlined by Cheng, Ho and Ruderman (1986) and Ho (1989). Pair-production due to photon-photon interactions and radiation processes including curvature, synchrotron and inverse Compton processes are computed with an iterative scheme which converges to self-consistent photon and particle distributions for a sampling of locations in the outer magnetosphere. We follow the photons from these distributions as they propagate through the pulsar magnetosphere toward a distant observer. We include the effects of relativistic aberration, time-of-flight delays and reabsorption by photon-photon pair-production to determine an intensity map of the high energy pulsar emission on the sky. Using data from radio and optical observations to constrain the geometry of the magnetosphere as well as the possible observer viewing angles, we derive light curves and phase dependent spectra which can be directly compared to data from the Compton Observatory. Observations for Crab, Vela and the recently identified gamma ray pulsars Geminga, PSR1706-44 aNd PSR 1509-58 will provide important tests of our model calculations, help us to improve our picture of the relevant physics at work in pulsar magnetospheres and allow us to comment on the implications for future pulsar discoveries

  14. Residual Stress Testing of Outer 3013 Containers

    International Nuclear Information System (INIS)

    Dunn, K.

    2004-01-01

    A Gas Tungsten Arc Welded (GTAW) outer 3013 container and a laser welded outer 3013 container have been tested for residual stresses according to the American Society for Testing Materials (ASTM) Standard G-36-94 [1]. This ASTM standard describes a procedure for conducting stress-corrosion cracking tests in boiling magnesium chloride (MgCl2) solution. Container sections in both the as-fabricated condition as well as the closure welded condition were evaluated. Significantly large residual stresses were observed in the bottom half of the as-fabricated container, a result of the base to can fabrication weld because through wall cracks were observed perpendicular to the weld. This observation indicates that regardless of the closure weld technique, sufficient residual stresses exist in the as-fabricated container to provide the stress necessary for stress corrosion cracking of the container, at the base fabrication weld. Additionally, sufficiently high residual stresses were observed in both the lid and the body of the GTAW as well as the laser closure welded containers. The stresses are oriented perpendicular to the closure weld in both the container lid and the container body. Although the boiling MgCl2 test is not a quantitative test, a comparison of the test results from the closure welds shows that there are noticeably more through wall cracks in the laser closure welded container than in the GTAW closure welded container

  15. Radioiodination of an outer membrane protein in intact Rickettsia prowazekii

    International Nuclear Information System (INIS)

    Smith, D.K.; Winkler, H.H.

    1980-01-01

    Intact Rickettsia prowazekii was radiolabeled with the glucose oxidase-lactoperoxidase method of iodination. Separation of the rickettsial extract into cytoplasmic, outer and inner membrane fractions demonstrated that the outer membrane was preferentially labeled. Analysis of the polypeptides of these fractions on high-resolution slab polyacrylamide gels showed that most of the 125 I was in polypeptide T49, an outer membrane constituent. Additional outer membrane polypeptides were iodinated in broken envelope preparations, demonstrating that T49 is uniquely accessible to the external environment and the asymmetric polypeptide organization of the outer membrane

  16. The simulation of acoustic propagation within SG water leak detection system

    International Nuclear Information System (INIS)

    Suzuki, Takehiko; Shioyama, Tsutomu

    1996-01-01

    It is important to detect the leak sound signal in a steam generator tube. For this purpose, it is necessary to develop the detection system capable of detecting the leak sound signal buried in external noises. This leak sound signal is measured the acceleration on the wall of the steam generator. The authors used a simulation technique to investigate how the sound generated in a steam generator propagates out of the generator. The results obtained using the simulation technique clarify that the observed signal had many resonation frequency courses due to scattering from complex structures. Therefore, the information of the original signal is lost. However, if the acceleration value many points on the outer wall is detected, and cross-correlation are obtained from each coupled measurement point, it is possible to separate the direct wave from a source point from the scattering waves in a measurement signal. Using the cross-correlation value, the source point of the leak sound signal in a steam generator tube is determined by the synthetic aperture focusing technique

  17. Synthesis and characterization of a novel tube-in-tube nanostructured PPy/MnO{sub 2}/CNTs composite for supercapacitor

    Energy Technology Data Exchange (ETDEWEB)

    Li, Juan, E-mail: lj-panpan@163.com [College of Chemistry and Chemical Engineering, Xinjiang University, Urumqi 830046 (China); Beijing National Laboratory for Molecular Sciences (BNLMS), State Key Laboratory for Structural Chemistry of Unstable and Stable Species, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871 (China); Que, Tingli [College of Chemistry and Chemical Engineering, Xinjiang University, Urumqi 830046 (China); Huang, Jianbin, E-mail: JBhuang@pku.edu.cn [College of Chemistry and Chemical Engineering, Xinjiang University, Urumqi 830046 (China); Beijing National Laboratory for Molecular Sciences (BNLMS), State Key Laboratory for Structural Chemistry of Unstable and Stable Species, College of Chemistry and Molecular Engineering, Peking University, Beijing 100871 (China)

    2013-02-15

    Graphical abstract: A novel tube-in-tube nanostructured PPy/MnO{sub 2}/CNTs composite have been successfully fabricated. Its inner tubules are CNTs and the outer tubules are template-synthesized PPy. Most MnO{sub 2} nanoparticles are sandwiched between the inner and outer wall, some relatively large particles are also latched onto the outside wall of the PPy tube. The composite yields a good electrochemical reversibility through 1000 cycles’ cyclic voltammogram (CV) test and galvanostatic charge–discharge experiments at different current densities. Display Omitted Highlights: ► We fabricate a ternary organic–inorganic complex of PPy/MnO{sub 2}/CNTs composite. ► We characterize its morphological structures and properties by several techniques. ► The composite possesses the typical tube-in-tube nanostructures. ► Most MnO{sub 2} nanoparticles are sandwiched between the inner CNTs and outer PPy wall. ► The composite has good electrochemical reversibility for supercapacitor. -- Abstract: Ternary organic–inorganic complex of polypyrrole/manganese dioxide/carbon nanotubes (PPy/MnO{sub 2}/CNTs) composite was prepared by in situ chemical oxidation polymerization of pyrrole in the host of inorganic matrix of MnO{sub 2} and CNTs, using complex of methyl orange (MO)/FeCl{sub 3} was used as a reactive self-degraded soft-template. The morphological structures of the composite were characterized by scanning electron microscopy (SEM), transmission electron microscopy (TEM), high-resolution transmission electron microscopic (HRTEM), Fourier transform infrared spectroscopy (FT-IR) and X-ray diffraction (XRD), respectively. All the results indicate that the PPy/MnO{sub 2}/CNTs composite possesses the typical tube-in-tube nanostructures: the inner tubules are CNTs and the outer tubules are template-synthesized PPy. MnO{sub 2} nanoparticles may either sandwich the space between the inner and outer tubules or directly latch onto the wall of the PPy tubes. The composite

  18. Diaphragm flange and method for lowering particle beam impedance at connected beam tubes of a particle accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Biallas, George Herman

    2017-07-04

    A diaphragm flange for connecting the tubes in a particle accelerator while minimizing beamline impedance. The diaphragm flange includes an outer flange and a thin diaphragm integral with the outer flange. Bolt holes in the outer flange provide a means for bolting the diaphragm flange to an adjacent flange or beam tube having a mating bolt-hole pattern. The diaphragm flange includes a first surface for connection to the tube of a particle accelerator beamline and a second surface for connection to a CF flange. The second surface includes a recessed surface therein and a knife-edge on the recessed surface. The diaphragm includes a thickness that enables flexing of the integral diaphragm during assembly of beamline components. The knife-edge enables compression of a soft metal gasket to provide a leak-tight seal.

  19. Data report of ROSA/LSTF experiment SB-HL-12. 1% hot leg break LOCA with SG depressurization and gas inflow

    International Nuclear Information System (INIS)

    Takeda, Takeshi

    2016-01-01

    An experiment SB-HL-12 was conducted on February 24, 1998 using the Large Scale Test Facility (LSTF) in the Rig of Safety Assessment-V (ROSA-V) Program. The ROSA/LSTF experiment SB-HL-12 simulated a 1% hot leg small-break loss-of-coolant accident in a pressurized water reactor under assumptions of total failure of high pressure injection system and non-condensable gas (nitrogen gas) inflow to the primary system from accumulator (ACC) tanks of emergency core cooling system (ECCS). Steam generator (SG) secondary-side depressurization by fully opening the relief valves in both SGs as an accident management (AM) action was initiated immediately after maximum surface temperature of simulated fuel rod reached 600 K. Auxiliary feedwater injection into the secondary-side of both SGs was started immediately after the initiation of AM action. After the onset of AM action due to first core uncovery by core boil-off, the primary pressure decreased following the SG secondary-side pressure, causing core mixture level swell. The fuel rod surface temperature then increased up to 635 K. Second core uncovery by core boil-off took place before loop seal clearing (LSC) induced by steam condensation on ACC coolant injected into cold legs. The core liquid level recovered rapidly after the LSC. The fuel rod surface temperature then increased up to 696 K. The pressure difference became larger between the primary and SG secondary sides after the ACC tanks started to discharge nitrogen gas, which resulted in no actuation of LPI system of ECCS during the experiment. Third core uncovery by core boil-off occurred during the reflux condensation in the SG U-tubes under nitrogen gas inflow. The core power was automatically decreased by the LSTF core protection system when the maximum fuel rod surface temperature exceeded 908 K. The obtained data would be useful to define the conditions for counterpart testing of other integral test facilities to address scaling problems through thermal

  20. 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

  1. A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herer, C. [RRAMATOME EP/TC, Paris (France); Souyri, A. [EdF DER/RNE/TTA, Chatou (France); Garnier, J. [CEA DRN/DTP/STR/LETC, Grenoble (France)

    1995-09-01

    Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to the annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.

  2. Track chambers based on precision drift tubes housed inside 30 mm mylar pipe

    International Nuclear Information System (INIS)

    Borisov, A; Bozhko, N; Fakhrutdinov, R; Kozhin, A; Leontiev, B; Levin, A

    2014-01-01

    We describe drift chambers consisting of 3 layers of 30 mm (OD) drift tubes made of double sided aluminized mylar film with thickness 0.125 mm. A single drift tube is self-supported structure withstanding 350 g tension of 50 microns sense wire located in the tube center with 10 microns precision with respect to end-plug outer surface. Such tubes allow to create drift chambers with small amount of material, construction of such chambers doesn't require hard frames. Twenty six chambers with working area from 0.8 × 1.0 to 2.5 × 2.0 m 2 including 4440 tubes have been manufactured for experiments at 70-GeV proton accelerator at IHEP(Protvino)

  3. Track chambers based on precision drift tubes housed inside 30 mm mylar pipe

    Science.gov (United States)

    Borisov, A.; Bozhko, N.; Fakhrutdinov, R.; Kozhin, A.; Leontiev, B.; Levin, A.

    2014-06-01

    We describe drift chambers consisting of 3 layers of 30 mm (OD) drift tubes made of double sided aluminized mylar film with thickness 0.125 mm. A single drift tube is self-supported structure withstanding 350 g tension of 50 microns sense wire located in the tube center with 10 microns precision with respect to end-plug outer surface. Such tubes allow to create drift chambers with small amount of material, construction of such chambers doesn't require hard frames. Twenty six chambers with working area from 0.8 × 1.0 to 2.5 × 2.0 m2 including 4440 tubes have been manufactured for experiments at 70-GeV proton accelerator at IHEP(Protvino).

  4. Investigation into Cause of High Temperature Failure of Boiler Superheater Tube

    Science.gov (United States)

    Ghosh, D.; Ray, S.; Roy, H.; Shukla, A. K.

    2015-04-01

    The failure of the boiler tubes occur due to various reasons like creep, fatigue, corrosion and erosion. This paper highlights a case study of typical premature failure of a final superheater tube of 210 MW thermal power plant boiler. Visual examination, dimensional measurement, chemical analysis, oxide scale thickness measurement, microstructural examination are conducted as part of the investigations. Apart from these investigations, sulfur print, Energy Dispersive spectroscopy (EDS) and X ray diffraction analysis (XRD) are also conducted to ascertain the probable cause of failure of final super heater tube. Finally it has been concluded that the premature failure of the super heater tube can be attributed to the combination of localized high tube metal temperature and loss of metal from the outer surface due to high temperature corrosion. The corrective actions have also been suggested to avoid this type of failure in near future.

  5. 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

  6. Framatome technologies successfully performs SG re-sleeving at Kewaunee

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    Framatome Technologies teamed with ABB-CE (Connecticut) to remove previously installed sleeves from the Kewaunee nuclear power stations steam generator tubes and install new sleeves. Recent inspections indicated problems at the upper joint of many of the previous sleeves. The process involved removing most of the hydraulic expansion joint (HEJ) sleeve, expanding the sleeve remnant and the tube section at the upper joint, and installing a TIG-welded sleeve that bridges the problem portion of the tube that the HEJ sleeve covered

  7. Drift tubes of Linac 2

    CERN Multimedia

    Photographic Service

    1977-01-01

    Being redied for installation, those at the right are for tank 1, those on the left for tank 2. Contrary to Linac 1, which had drift-tubes supported on stems, here the tubes are suspended, for better mechanical stability.

  8. NEI You Tube Videos: Amblyopia

    Medline Plus

    Full Text Available ... YouTube Videos: Amblyopia Embedded video for NEI YouTube Videos: Amblyopia ... *PDF files require the free Adobe® Reader® software for viewing. This website is maintained by the ...

  9. Tubing For Sampling Hydrazine Vapor

    Science.gov (United States)

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

    1993-01-01

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

  10. Prospects for stronger calandria tubes

    International Nuclear Information System (INIS)

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

    1990-12-01

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

  11. Frequent sgRNA-barcode recombination in single-cell perturbation assays.

    Directory of Open Access Journals (Sweden)

    Shiqi Xie

    Full Text Available Simultaneously detecting CRISPR-based perturbations and induced transcriptional changes in the same cell is a powerful approach to unraveling genome function. Several lentiviral approaches have been developed, some of which rely on the detection of distally located genetic barcodes as an indirect proxy of sgRNA identity. Since barcodes are often several kilobases from their corresponding sgRNAs, viral recombination-mediated swapping of barcodes and sgRNAs is feasible. Using a self-circularization-based sgRNA-barcode library preparation protocol, we estimate the recombination rate to be ~50% and we trace this phenomenon to the pooled viral packaging step. Recombination is random, and decreases the signal-to-noise ratio of the assay. Our results suggest that alternative approaches can increase the throughput and sensitivity of single-cell perturbation assays.

  12. 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.

  13. Expansion lyre-shaped tube

    International Nuclear Information System (INIS)

    Andro, Jean.

    1973-01-01

    The invention relates the expansion lyre-shaped tube portions formed in dudgeoned tubular bundles between two bottom plates. An expansion lyre comprises at least two sets of tubes of unequal lengths coplanar and symmetrical with respect to the main tube axis, with connecting portions between the tubes forming said sets. The invention applies to apparatus such as heat exchangers, heaters, superheaters or breeders [fr

  14. Metal diffusion from furnace tubes depends on location

    International Nuclear Information System (INIS)

    Albright, L.F.

    1988-01-01

    Studies of metal samples from an ethylene furnace on the Texas Gulf Coast, using a scanning electron microscope (SEM) and an energy dispersive X-ray analyzer (EDAX), reveal preferential diffusion of chromium, titanium, and aluminum in the coil wall to the surfaces of the tube where they form metal oxides. These elements are gradually depleted from the tube wall. Complicated surface reactions that include the formation of several metal oxides, metal sulfides, and metal-catalyzed coke also occur. Several mechanisms can be postulated as to how metal fines or compounds are formed and transferred in the coil and transfer lines exchanger (TLX) of ethylene units. These surface reactions directly or indirectly affect coke formation in the tube. Finally, creep in the coils is likely a factor in promoting corrosion. Such creep is promoted by variable temperature-time patterns to which a coil is exposed during pyrolysis, and then decoking. Periods of stress and compression occur in the coil walls. Knowledge of the diffusion and reactions that take place can result in better furnace operations and decoking procedures to extend the life of the furnace tubes. In this second installment of a four-part series, photomicrographs of four pyrolysis tube samples from the ethylene furnace indicate that significant differences existed between the outer surfaces, inner surfaces, and cross-sectional areas of the samples. The first installment of the series dealt with coke

  15. Numerical simulation of triple concentric-tube heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Valladares, O. [Centro de Investigacion en Energia (CIE), Universidad Nacional Autonoma de Mexico (UNAM), Privada Xochicalco S/N, Temixco, 62580, Morelos (Mexico)

    2004-10-01

    A detailed one-dimensional steady and transient numerical simulation of the thermal and fluid-dynamic behaviour of triple concentric-tube heat exchangers has been developed. The governing equations (continuity, momentum and energy) inside the inner tube and the annulus (inner and outer), together with the energy equations in the inner, intermediate and outermost tube wall and insulation, are solved iteratively in a segregated manner. The discretized governing equations in the zones with fluid flow are coupled using an implicit step by step method. This formulation requires the use of empirical information for the evaluation of convective heat transfer, shear stress and void fraction. An implicit central difference numerical scheme and a line-by-line solver was used in the inner and intermediate tube walls and the outermost tube wall with insulation. All the flow variables (enthalpies, temperatures, pressures, mass fractions, velocities, heat fluxes, etc.) together with the thermophysical properties are evaluated at each point of the grid in which the domain is discretized. Different numerical aspects and comparisons with results obtained from the technical literature are presented in order to verify and validate the model. (authors)

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

    International Nuclear Information System (INIS)

    Velasco, Raphael; Boudeau, Nathalie

    2007-01-01

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

  17. Tapered leaf support pin for operating plant guide tubes

    International Nuclear Information System (INIS)

    Land, J.T.; Hopkins, R.J.; Ford, D.E.

    1991-01-01

    This patent describes a mounting system for removably mounting the lower flange of a control rod guide tube over an opening in the upper core plate of a nuclear reactor comprising at least one elongated support pin mounted on the guide tube lower flange and resiliently receivable in a bore formed in the upper core plate. It comprises a support pin having a longitudinal axis and comprising a first pin portion mountable on the guide tube lower flange, and a second pin portion receivable within the upper core plate bore, the second pin portion including a solid body section adjacent the first pin portion and having an outer diameter which is accommodated by the bore by a close clearance fit; locking means mounted on the first pin portion of the support pin for retaining the guide tube lower flange between the solid body section of the second pin portion and the locking means; and a washer disposed around the first pin portion between the locking means and the control rod guide tube flange, the washer and the locking means including mutually engaging rounded surfaces for eliminating bending moments and stresses on the support pin during mounting of the locking means on the first pin portion of the support pin

  18. Fluisd elastic instability and fretting-wear characteristics of steam generator helical tubes subjected to single-phase external flow and two-phase internal flow

    International Nuclear Information System (INIS)

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

    2004-01-01

    This study investigates the fluid elastic instability characteristics of steam generator (SG) helical type tubes and the safety assessment of the potential for fretting-wear damages caused by foreign object in operating nuclear power plants. The thermal-hydraulic conditions of both tube side and shell side flow fields are predicted by a general purpose computational fluid dynamics code employing the finite volume element modeling. 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 on the effects of coil diameter and the number of turns 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 helical type 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 external pressure on the vibration and fretting wear characteristics of the tube

  19. 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

  20. Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

    International Nuclear Information System (INIS)

    Jiang, J.T.; Luxat, J.C.

    2008-01-01

    Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

  1. Modeling the quenching of a calandria tube following a critical break LOCA in a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

    2008-07-01

    Following a postulated critical large break LOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a CANDU CT (approximately 130mm). The model has been developed to analyze the variation of steady state vapor film thickness as a function of sub-cooling temperature, wall superheat and incident heat flux. The CT outer surface heat flux and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (author)

  2. Chest tube insertion - series (image)

    Science.gov (United States)

    Chest tubes are inserted to drain blood, fluid, or air and allow full expansion of the lungs. The tube is placed in the pleural space. The area where the tube will be inserted is numbed (local anesthesia). The patient may also be sedated. The chest ...

  3. Tube Length and Water Flow

    Directory of Open Access Journals (Sweden)

    Ben Ruktantichoke

    2011-06-01

    Full Text Available In this study water flowed through a straight horizontal plastic tube placed at the bottom of a large tank of water. The effect of changing the length of tubing on the velocity of flow was investigated. It was found that the Hagen-Poiseuille Equation is valid when the effect of water entering the tube is accounted for.

  4. Tubing cutter for tight spaces

    Science.gov (United States)

    Girala, A. S.

    1980-01-01

    Cutter requires few short swings of handle to rotate its cutting edge full 360 around tube. It will cut tubing installed in confined space that prevents free movement of conventional cutter. Cutter is snapped onto tube and held in place by spring-loaded clamp. Screw ratchet advances cutting wheel.

  5. Heat Exchanger Tube to Tube Sheet Joints Corrosion Behavior

    Directory of Open Access Journals (Sweden)

    M. Iancu

    2013-03-01

    Full Text Available Paper presents the studies made by the authors above the tube to tube sheet fittings of heat exchanger with fixed covers from hydrofining oil reforming unit. Tube fittings are critical zones for heat exchangers failures. On a device made from material tube and tube sheet at real joints dimensions were establish axial compression force and traction force at which tube is extracted from expanded joint. Were used two shapes joints with two types of fittings surfaces, one with smooth hole of tube sheet and other in which on boring surface we made a groove. From extracted expanded tube zones were made samples for corrosion tests in order to establish the corrosion rate, corrosion potential and corrosion current in working mediums such as hydrofining oil and industrial water at different temperatures. The corrosion rate values and the temperature influence are important to evaluate joints durability and also the results obtained shows that the boring tube sheet shape with a groove on hole tube shape presents a better corrosion behavior then the shape with smooth hole tube sheet.

  6. Active Intracellular Delivery of a Cas9/sgRNA Complex Using Ultrasound-Propelled Nanomotors.

    Science.gov (United States)

    Hansen-Bruhn, Malthe; de Ávila, Berta Esteban-Fernández; Beltrán-Gastélum, Mara; Zhao, Jing; Ramírez-Herrera, Doris E; Angsantikul, Pavimol; Vesterager Gothelf, Kurt; Zhang, Liangfang; Wang, Joseph

    2018-03-01

    Direct and rapid intracellular delivery of a functional Cas9/sgRNA complex using ultrasound-powered nanomotors is reported. The Cas9/sgRNA complex is loaded onto the nanomotor surface through a reversible disulfide linkage. A 5 min ultrasound treatment enables the Cas9/sgRNA-loaded nanomotors to directly penetrate through the plasma membrane of GFP-expressing B16F10 cells. The Cas9/sgRNA is released inside the cells to achieve highly effective GFP gene knockout. The acoustic Cas9/sgRNA-loaded nanomotors display more than 80 % GFP knockout within 2 h of cell incubation compared to 30 % knockout using static nanowires. More impressively, the nanomotors enable highly efficient knockout with just 0.6 nm of the Cas9/sgRNA complex. This nanomotor-based intracellular delivery method thus offers an attractive route to overcome physiological barriers for intracellular delivery of functional proteins and RNAs, thus indicating considerable promise for highly efficient therapeutic applications. © 2018 Wiley-VCH Verlag GmbH & Co. KGaA, Weinheim.

  7. Antagonistic action of Bacillus subtilis strain SG6 on Fusarium graminearum.

    Science.gov (United States)

    Zhao, Yueju; Selvaraj, Jonathan Nimal; Xing, Fuguo; Zhou, Lu; Wang, Yan; Song, Huimin; Tan, Xinxin; Sun, Lichao; Sangare, Lancine; Folly, Yawa Minnie Elodie; Liu, Yang

    2014-01-01

    Fusarium graminearum causes Fusarium head blight (FHB), a devastating disease that leads to extensive yield and quality loss of wheat and barley. Bacteria isolated from wheat kernels and plant anthers were screened for antagonistic activity against F. graminearum. Based on its in vitro effectiveness, strain SG6 was selected for characterization and identified as Bacillus subtilis. B. subtilis SG6 exhibited a high antifungal effect on the mycelium growth, sporulation and DON production of F. graminearum with the inhibition rate of 87.9%, 95.6% and 100%, respectively. In order to gain insight into biological control effect in situ, we applied B. subtilis SG6 at anthesis through the soft dough stage of kernel development in field test. It was revealed that B. subtilis SG6 significantly reduced disease incidence (DI), FHB index and DON (P ≤ 0.05). Further, ultrastructural examination shows that B. subtilis SG6 strain induced stripping of F. graminearum hyphal surface by destroying the cellular structure. When hypha cell wall was damaged, the organelles and cytoplasm inside cell would exude, leading to cell death. The antifungal activity of SG6 could be associated with the coproduction of chitinase, fengycins and surfactins.

  8. Antagonistic action of Bacillus subtilis strain SG6 on Fusarium graminearum.

    Directory of Open Access Journals (Sweden)

    Yueju Zhao

    Full Text Available Fusarium graminearum causes Fusarium head blight (FHB, a devastating disease that leads to extensive yield and quality loss of wheat and barley. Bacteria isolated from wheat kernels and plant anthers were screened for antagonistic activity against F. graminearum. Based on its in vitro effectiveness, strain SG6 was selected for characterization and identified as Bacillus subtilis. B. subtilis SG6 exhibited a high antifungal effect on the mycelium growth, sporulation and DON production of F. graminearum with the inhibition rate of 87.9%, 95.6% and 100%, respectively. In order to gain insight into biological control effect in situ, we applied B. subtilis SG6 at anthesis through the soft dough stage of kernel development in field test. It was revealed that B. subtilis SG6 significantly reduced disease incidence (DI, FHB index and DON (P ≤ 0.05. Further, ultrastructural examination shows that B. subtilis SG6 strain induced stripping of F. graminearum hyphal surface by destroying the cellular structure. When hypha cell wall was damaged, the organelles and cytoplasm inside cell would exude, leading to cell death. The antifungal activity of SG6 could be associated with the coproduction of chitinase, fengycins and surfactins.

  9. Recurrent Neural Network for Computing Outer Inverse.

    Science.gov (United States)

    Živković, Ivan S; Stanimirović, Predrag S; Wei, Yimin

    2016-05-01

    Two linear recurrent neural networks for generating outer inverses with prescribed range and null space are defined. Each of the proposed recurrent neural networks is based on the matrix-valued differential equation, a generalization of dynamic equations proposed earlier for the nonsingular matrix inversion, the Moore-Penrose inversion, as well as the Drazin inversion, under the condition of zero initial state. The application of the first approach is conditioned by the properties of the spectrum of a certain matrix; the second approach eliminates this drawback, though at the cost of increasing the number of matrix operations. The cases corresponding to the most common generalized inverses are defined. The conditions that ensure stability of the proposed neural network are presented. Illustrative examples present the results of numerical simulations.

  10. Ageing of the LHCb outer tracker

    CERN Document Server

    Blom, M R; Tuning, N

    2009-01-01

    The modules of the LHCb outer tracker have shown to suffer severe gain loss under moderate irradiation. This process is called ageing. Ageing of the modules results from contamination of the gas system by glue, araldite AY 103-1, used in their construction. In this thesis the ageing process will be shown. The schemes known to reduce, reverse, or prevent ageing have been investigated to determine their effect on the detector performance. The addition of O2 to the gas mixture lowers the detector response by an acceptable amount and does not affect the gas transport properties significantly. The ageing rate is decreased after extensive flushing and HV training could eventually repair the irradiation damage. The risks of HV training have been assessed. Furthermore, several gaseous and aquatic additions have been tested for their capability to prevent, or moderate ageing, but none showed significant improvement.

  11. YouTube Physics

    Science.gov (United States)

    Riendeau, Diane

    2012-09-01

    To date, this column has presented videos to show in class, Don Mathieson from Tulsa Community College suggested that YouTube could be used in another fashion. In Don's experience, his students are not always prepared for the mathematic rigor of his course. Even at the high school level, math can be a barrier for physics students. Walid Shihabi, a colleague of Don's, decided to compile a list of YouTube videos that his students could watch to relearn basic mathematics. I thought this sounded like a fantastic idea and a great service to the students. Walid graciously agreed to share his list and I have reproduced a large portion of it below.

  12. Neutron image intensifier tubes

    International Nuclear Information System (INIS)

    Verat, M.; Rougeot, H.; Driard, B.

    1983-01-01

    The most frequently used techniques in neutron radiography employ a neutron converter consisting of either a scintillator or a thin metal sheet. The radiation created by the neutrons exposes a photographic film that is in contact with the converter: in the direct method, the film is exposed during the time that the object is irradiated with neutrons; in the transfer method, the film is exposed after the irradiation of the object with neutrons. In industrial non-destructive testing, when many identical objects have to be checked, these techniques have several disadvantages. Non-destructive testing systems without these disadvantages can be constructed around neutron-image intensifier tubes. A description and the operating characteristics of neutron-image intensifier tubes are given. (Auth.)

  13. Tube coupling device

    Science.gov (United States)

    Myers, William N. (Inventor); Hein, Leopold A. (Inventor)

    1987-01-01

    A first annular ring of a tube coupling device has a keyed opening sized to fit around the nut region of a male coupling, and a second annular ring has a keyed opening sized to fit around the nut of a female coupling. Each ring has mating ratchet teeth and these rings are biased together, thereby engaging these teeth and preventing rotation of these rings. This in turn prevents the rotation of the male nut region with respect to the female nut. For tube-to-bulkhead locking, one facet of one ring is notched, and a pin is pressed into an opening in the bulkhead. This pin is sized to fit within one of the notches in the ring, thereby preventing rotation of this ring with respect to the bulkhead.

  14. PEG Tube Placement

    Directory of Open Access Journals (Sweden)

    Saptarshi Biswas

    2014-01-01

    Full Text Available Percutaneous endoscopic gastrostomy (PEG has been used for providing enteral access to patients who require long-term enteral nutrition for years. Although generally considered safe, PEG tube placement can be associated with many immediate and delayed complications. Buried bumper syndrome (BBS is one of the uncommon and late complications of percutaneous endoscopic gastrostomy (PEG placement. It occurs when the internal bumper of the PEG tube erodes into the gastric wall and lodges itself between the gastric wall and skin. This can lead to a variety of additional complications such as wound infection, peritonitis, and necrotizing fasciitis. We present here a case of buried bumper syndrome which caused extensive necrosis of the anterior abdominal wall.

  15. Salvaging of service exposed cast alloy 625 cracker tubes of ammonia based Heavy Water Plants

    International Nuclear Information System (INIS)

    Kumar, Niraj; Misra, B.; Mahajan, M.P.; Mittra, J.; Sundararaman, M.; Chakravartty, J.K.

    2006-01-01

    In ammonia based heavy water plants, cracking of ammonia vapour, enriched in deuterium is carried out inside a cracker tube, packed with catalyst. These cracker tubes are made of alloy 625 (either wrought or cast) having dimensions of about 12.5 metres long, 88 mm outer diameter and 7.9 mm wall thickness. Seventy such tubes are housed in a typical ammonia cracker unit. The anticipated design life of such tube is 1,00,000 hrs. when operated at 720 degC based on creep as main degradation mechanism. Presently, these tubes are being operated at 680 degC skin temperature. Alloy 625 tubes are costly and normally not manufactured in India and are being imported. The cast alloy 625 cracker tubes have outlived their design life of 100,000 hrs. Therefore it has been decided to salvage the cast cracker tubes and extend the life further as it had already been done for wrought tubes. Similar to the earlier attempt of resolutionising of wrought alloy 625 tubes, efforts are in progress to salvage these cast tubes. In this study, cast tubes samples were subjected to solution-annealing treatment at two different temperatures, 1100degC and 1160degC respectively for two hrs. Mechanical properties along with the microstructure of the samples, which were resolutionized at 1160degC were comparable with that of virgin material. The 12.5 metres long cast alloy 625 cracker tubes will also be shortly solution-annealed in a specially designed resistance heating furnace after completing some more tests. (author)

  16. Impulsive ion acceleration in earth's outer magnetosphere

    International Nuclear Information System (INIS)

    Baker, D.N.; Belian, R.D.

    1985-01-01

    Considerable observational evidence is found that ions are accelerated to high energies in the outer magnetosphere during geomagnetic disturbances. The acceleration often appears to be quite impulsive causing temporally brief (10's of seconds), very intense bursts of ions in the distant plasma sheet as well as in the near-tail region. These ion bursts extend in energy from 10's of keV to over 1 MeV and are closely associated with substorm expansive phase onsets. Although the very energetic ions are not of dominant importance for magnetotail plasma dynamics, they serve as an important tracer population. Their absolute intensity and brief temporal appearance bespeaks a strong and rapid acceleration process in the near-tail, very probably involving large induced electric fields substantially greater than those associated with cross-tail potential drops. Subsequent to their impulsive acceleration, these ions are injected into the outer trapping regions forming ion ''drift echo'' events, as well as streaming tailward away from their acceleration site in the near-earth plasma sheet. Most auroral ion acceleration processes occur (or are greatly enhanced) during the time that these global magnetospheric events are occurring in the magnetotail. A qualitative model relating energetic ion populations to near-tail magnetic reconnection at substorm onset followed by global redistribution is quite successful in explaining the primary observational features. Recent measurements of the elemental composition and charge-states have proven valuable for showing the source (solar wind or ionosphere) of the original plasma population from which the ions were accelerated

  17. Wide cross-reactivity between Anopheles gambiae and Anopheles funestus SG6 salivary proteins supports exploitation of gSG6 as a marker of human exposure to major malaria vectors in tropical Africa

    Directory of Open Access Journals (Sweden)

    Petrarca Vincenzo

    2011-07-01

    Full Text Available Abstract Background The Anopheles gambiae gSG6 is an anopheline-specific salivary protein which helps female mosquitoes to efficiently feed on blood. Besides its role in haematophagy, gSG6 is immunogenic and elicits in exposed individuals an IgG response, which may be used as indicator of exposure to the main African malaria vector A. gambiae. However, malaria transmission in tropical Africa is sustained by three main vectors (A. gambiae, Anopheles arabiensis and Anopheles funestus and a general marker, reflecting exposure to at least these three species, would be especially valuable. The SG6 protein is highly conserved within the A. gambiae species complex whereas the A. funestus homologue, fSG6, is more divergent (80% identity with gSG6. The aim of this study was to evaluate cross-reactivity of human sera to gSG6 and fSG6. Methods The A. funestus SG6 protein was expressed/purified and the humoral response to gSG6, fSG6 and a combination of the two antigens was compared in a population from a malaria hyperendemic area of Burkina Faso where both vectors were present, although with a large A. gambiae prevalence (>75%. Sera collected at the beginning and at the end of the high transmission/rainy season, as well as during the following low transmission/dry season, were analysed. Results According to previous observations, both anti-SG6 IgG level and prevalence decreased during the low transmission/dry season and showed a typical age-dependent pattern. No significant difference in the response to the two antigens was found, although their combined use yielded in most cases higher IgG level. Conclusions Comparative analysis of gSG6 and fSG6 immunogenicity to humans suggests the occurrence of a wide cross-reactivity, even though the two proteins carry species-specific epitopes. This study supports the use of gSG6 as reliable indicator of exposure to the three main African malaria vectors, a marker which may be useful to monitor malaria transmission

  18. Recurrent inspection of tubes

    International Nuclear Information System (INIS)

    Andersson, S.

    1984-01-01

    Recommendations concerning the selection of areas for the inspection of tubes have been outlined. The aim is to focus the control on the regions which are important for the safety and where damage is supposed to take place. The number of zones will depend on the risk factors as judged by experts. The localizing will be based upon probable damaging mechanisms. A certain number of areas should be chosen at random. (G.B.)

  19. Cathode ray tube screens

    International Nuclear Information System (INIS)

    Cockayne, B.; Robbins, D.J.; Glasper, J.L.

    1982-01-01

    An improved cathode ray tube screen is described which consists of a single- or a poly-crystalline slice of a material such as yttrium aluminium garnet in which dopants such as Tb 3 + , Eu 3 + , Ce 3 + or Tm 3 + are ion implanted to different depths or in different areas of the screen. Annealing the screen removes lattice damage caused by the ion implanting and assists the diffusion of the dopant into the crystal. (U.K.)

  20. Cathode ray tube

    International Nuclear Information System (INIS)

    1979-01-01

    A cathode ray tube comprises two electron lens means in combination to crossover the electron beam at a second crossover between the two electron lens means with one of the two lens means having a variable voltage applied thereto to control the location of the beam crossover in order to focus the beam onto a display screen at any location away from the screen center. (Auth.)