WorldWideScience

Sample records for generator tube plugging

  1. 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)

  2. Maintenance and plugging technology for CANDU steam generator tubing

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  4. 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)

  5. Grooved tube plug rolls in

    International Nuclear Information System (INIS)

    Krausser, P.

    1991-01-01

    The removable plugs used to date by the Power Generation Group (KWU) of Siemens to seal defective steam generator tubes have a good track record. Their sealing principle is based on the elastic tensioning of three seal disks against the inside wall of the tube. Now a further removable plug is available -a roll-in plug with a metal-coated surface. It is particularly suitable for use in the roller-expanded zone of the tubes at the tube sheet. The plugs can be used in both Siemens-KWU steam generators and in steam generators manufactured in compliance with the guidelines of the ASME Code. (author)

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  7. 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)

  8. 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

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  11. 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

  12. 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

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

    International Nuclear Information System (INIS)

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

    1981-05-01

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

  14. Eddy current inspection of steam generator tubing plugs

    International Nuclear Information System (INIS)

    Cullen, W.K.

    1990-01-01

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

  15. Roll-expanded plugs for steam generator heating tubes verification of leak tightness over the component lifetime

    International Nuclear Information System (INIS)

    Beck, J.; Ziegler, R.; Schönheit, N.

    2013-01-01

    Highlights: • Design description of roll-expanded plugs. • Experimental simulation of 40 years lifetime of plugged steam generator tubes. • Destructive testing for off-design loads. • Evaluation of release pressure and tightness before and after the tests. -- Abstract: Steam generator heating tubes are the boundary between the irradiated primary cycle and the conventional secondary cycle in a pressurized water reactor. Despite their operational task to transfer the heat from the primary to the secondary cycle, these tubes have a crucial safety function: the retention of irradiated primary coolant inside the circuit in all operating, emergency and off-design conditions. The heating tubes are subject to various degradation mechanisms during operation. To verify the integrity of each single tube, nuclear power plants carry out frequent in-service inspections. In case of a tube wall degradation beyond the permissible limit, the tube needs to be taken out of service in order to maintain the overall component integrity. The most common method to do so is to plug a damaged tube by a roll-expanded plug. After plugging, the roll-expanded plug acts as pressure boundary between the primary and the secondary cycle instead of the damaged heating tube. The plug must be able to maintain this function, previously provided by the heating tube, in all operational, emergency and off-design conditions. This article describes the approach to this verification by launching several comprehensive process qualification programmes consisting of mechanical analyses as well as static and dynamic testing programmes. The result was a qualified roll-expanded plug which remains leak-tight even during off-design conditions

  16. The impact of NPP Krsko steam generator tube plugging on minimum DNBR at nominal conditions

    International Nuclear Information System (INIS)

    Lajtman, S.

    1996-01-01

    Typically, steam generator tube plugging (SGTP) both decreases the reactor coolant system (RCS) flow rate and the heat transfer surface area of the steam generator. At a constant thermal power and vessel outlet temperature, as tube plugging increases, the vessel average temperature, vessel inlet temperature and steam generator secondary side steam pressure decrease. This paper presents the analysis of impact of SGTP on Minimum Departure from Nucleate Boiling Ratio (MDNBR) at NPP Krsko (NEK), using the Improved Thermal Design Procedure (ITDP), WRB-1 correlation, and COBRA-III-C computer code. No credit was given to high plugging percentage region power reduction resulting from turbine volumetric flow limitations. MDNBR is found to be decreasing with increasing plugging, but not under the limiting values. (author)

  17. The aspects regarding plugging the defective tubes of the steam generator using plastic deformation of the plug wall by conventional or unconventional techniques

    International Nuclear Information System (INIS)

    Gyongyosi, Tiberiu

    2006-01-01

    After a brief introduction the advantages and disadvantages of two plugging methods of the defective tubes from steam generator by plastic deformation of the wall of the plug, deformation performed by mechanical rolling (conventional technique) and by electrohydraulic shock (unconventional technique), respectively, are showed. The paper gives the results of the experimental tests to install the plugs at the end of the tube having the same geometry with those of the steam generator, the performance reached in this stage and some conclusions in the end

  18. 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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  1. Repair of steam generator heating tubes by roll-expanded plugs: approach to cover multiple national regulations

    Energy Technology Data Exchange (ETDEWEB)

    Beck, J.; Ziegler, B.; Schoenheit, N. [AREVA NP Gmbh, Erlangen (Germany); Kostroun, F. [AREVA NP Canada Ltd., Pickering, ON (Canada)

    2012-07-01

    During operation, steam generators in nuclear power plants are subject to degradation mechanisms which have an impact on the component life-time. Most affected are the heating tubes which constitute the barrier of the contaminated primary cycle to the secondary side. Various corrosive attacks may cause wall thinning which requires tube repair. A common repair method is to plug the tubes by roll expanded plugs. This is a fast method, easily applicable and requires less equipment or personnel qualification as needed for weld plugs. After insertion, the plugs act as a pressure boundary from primary to secondary side. Although the function of the roll plug is simple, the different national regulations define the requirements which need to be fulfilled by a roll plug differently. In order to reduce the tooling as well as the plug types to a minimum, an approach according to one common design for different regulations and steam generator types is profitable. It was found, that the regulations according to the ASME Boiler and Pressure Vessel code in combination with the German Safety Standards of the German Nuclear Safety Standards Commission covers the regulations of the majority of utilities. To develop a roll plug which suits the different regulatory demands, efforts were made to consider all technical and regulatory boundary conditions implied on roll expanded plugs. This covering approach had an impact on the plug design, which was required to be Helium tight after installation and suitable for a 40 year component lifetime also in accident and emergency conditions. To prove the suitability of the plug design a comprehensive testing programme of the mechanical and chemical properties of the designed roll-expanded plug was launched. A summary of the plug design and testing as well as the main test results are described. (author)

  2. Repair of steam generator heating tubes by roll-expanded plugs: approach to cover multiple national regulations

    International Nuclear Information System (INIS)

    Beck, J.; Ziegler, B.; Schoenheit, N.; Kostroun, F.

    2012-01-01

    During operation, steam generators in nuclear power plants are subject to degradation mechanisms which have an impact on the component life-time. Most affected are the heating tubes which constitute the barrier of the contaminated primary cycle to the secondary side. Various corrosive attacks may cause wall thinning which requires tube repair. A common repair method is to plug the tubes by roll expanded plugs. This is a fast method, easily applicable and requires less equipment or personnel qualification as needed for weld plugs. After insertion, the plugs act as a pressure boundary from primary to secondary side. Although the function of the roll plug is simple, the different national regulations define the requirements which need to be fulfilled by a roll plug differently. In order to reduce the tooling as well as the plug types to a minimum, an approach according to one common design for different regulations and steam generator types is profitable. It was found, that the regulations according to the ASME Boiler and Pressure Vessel code in combination with the German Safety Standards of the German Nuclear Safety Standards Commission covers the regulations of the majority of utilities. To develop a roll plug which suits the different regulatory demands, efforts were made to consider all technical and regulatory boundary conditions implied on roll expanded plugs. This covering approach had an impact on the plug design, which was required to be Helium tight after installation and suitable for a 40 year component lifetime also in accident and emergency conditions. To prove the suitability of the plug design a comprehensive testing programme of the mechanical and chemical properties of the designed roll-expanded plug was launched. A summary of the plug design and testing as well as the main test results are described. (author)

  3. Tube Plugging Criteria for the High-pressure Heaters of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyungnam; Cho, Nam-Cheoul; Lee, Kuk-hee [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this paper, a method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the high pressure heaters of a nuclear power plant. This method relies on the similar plugging criteria used in the steam generator tubes. Power generation field urges nuclear power plants to reduce operating and maintaining costs to remain competitive. To reduce the cost by means of preventing the lowering thermal efficiency, the inspection of balance-of-plant heat exchanger, which was treated as not important work, becomes important. The tubing materials and tube thickness of heat exchangers in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. But tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. The feed water heater is a kind of heat exchanger which raises the temperature of water supplied from the condenser. The heat source of high-pressure heaters is the extraction steam from the high-pressure turbine and moisture separator re-heater. If the tube wall of the heater is broken, the feed water flowing inside the tube intrudes to shell side. This forces the turbine to be stop in order to protect it. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. A method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the high pressure heaters of Ulchin NPP No. 3 and 4. This method relies on the similar plugging

  4. 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

  5. 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

  6. A Method to Establishing Tube Plugging Criterion for Heat Exchangers with Straight Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyungnam [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The difference of thermal expansion coefficients between the shell and tube materials causes the stress in axial direction of tube. Because of the axial stress due to thermal load, the straight tubes are used for heat exchangers operated in low temperature such as CCW (Component Cooling Water) heat exchangers and condensers. It is inevitable for the materials of the components to be degraded as the power plants become older. The degradation accompanies increasing maintenance cost as well as creating safety issues. The materials and wall thickness of heat exchanger tubes in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers, which is based on the USNRC Regulatory Guide 1.121, is introduced and the tube plugging criteria for the TPCCW heat exchanger of Yonggwang NPP No. 1 and 2. A method to establish the tube plugging criteria of heat exchangers with straight tubes are introduced based on the USNRC Regulatory Guide 1.121. As an example, the tube plugging criterion for the CCW heat exchanger of a nuclear power plant is provided.

  7. 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.

  8. 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.

  9. Apparatus for assembling and welding end plugs to nuclear fuel cladding tubes and inspecting the end plug welds on an automated basis

    International Nuclear Information System (INIS)

    Schoenig, F.C. Jr.; Walker, E.S.; Cueman, M.K.; Haughton, R.A.; Zuloaga, J.A. Jr.

    1989-01-01

    This patent describes an automated apparatus for welding a separate end plug to one open end of each of a succession of nuclear fuel cladding tubes and for inspecting each end plug weld. The apparatus comprising, in combination: a welding station; a cooldown station for cooling each end plug weld in an inert gas atmosphere; a serial number reader station for reading a serial number on each end plug; a first weld inspection station; a second weld inspection station for generating second weld inspection data; a computer system linked with the serial number reader and the first and second weld inspection stations; an input queue for holding a plurality of tubes; a tube transporter for periodically picking individual tubes from the input queque and conveying the tubes in a direction transverse to their tube axis in indexing steps to index positions respectively axially aligned with the welding, serial number reader, and first and second weld inspection stations; and a sorter positioned at an output end of the tube transporter

  10. Process and device for extracting a cylindrical element such a plug of a small diameter tube

    International Nuclear Information System (INIS)

    Ferlay, J.C.; Gente, D.; Jouin, D.

    1992-01-01

    The plug engaged in a steam generator tube is extracted by first machining away the expander and a part of its side wall then pulling the remainder of the plug from the tube with a tool engaging its closed end

  11. Tube Plugging Criterion for the TPCCW Heat Exchanger of Yonggwang NPP 1 and 2

    International Nuclear Information System (INIS)

    Kim, Hyung Nam; Yoo, Hyun Ju; Choi, Sung Nam; Song, Seok Yoon

    2009-01-01

    The turbine plant component cooling water(TPCCW) system circulates the cooling water to cool the components in the turbine building and discharges the heat from the components through the TPCCW heat exchanger. Recently, Yonggwang NPP 1 and 2 replaced the TPCCW heat exchanger because of tube degradation. The tubing material of new TPCCW heat exchanger of Yonggwang NPP 1 and 2 is titanium. If the tube wall cannot withstand the pressure, the cooling water with the chemicals flows into the tube side and it is discharged to the open water. The chemicals can pollute the open water. Therefore, the tubes of the TPCCW heat exchanger should be inspected and degraded tubes should be plugged. It is inevitable for the materials of the components to be degraded as the power plants become older. The degradation accompanies increasing maintenance cost as well as creating safety issues. The materials and wall thickness of heat exchanger tubes in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. However, tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers, which is based on the USNRC Regulatory Guide 1.121, is introduced and the tube plugging criteria for the TPCCW heat exchanger of Yonggwang NPP No. 1 and 2. This method relies on the similar plugging criteria used in the steam generator

  12. Tube Plugging Criterion for the TPCCW Heat Exchanger of Yonggwang NPP 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Nam; Yoo, Hyun Ju; Choi, Sung Nam; Song, Seok Yoon [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    2009-05-15

    The turbine plant component cooling water(TPCCW) system circulates the cooling water to cool the components in the turbine building and discharges the heat from the components through the TPCCW heat exchanger. Recently, Yonggwang NPP 1 and 2 replaced the TPCCW heat exchanger because of tube degradation. The tubing material of new TPCCW heat exchanger of Yonggwang NPP 1 and 2 is titanium. If the tube wall cannot withstand the pressure, the cooling water with the chemicals flows into the tube side and it is discharged to the open water. The chemicals can pollute the open water. Therefore, the tubes of the TPCCW heat exchanger should be inspected and degraded tubes should be plugged. It is inevitable for the materials of the components to be degraded as the power plants become older. The degradation accompanies increasing maintenance cost as well as creating safety issues. The materials and wall thickness of heat exchanger tubes in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. However, tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers, which is based on the USNRC Regulatory Guide 1.121, is introduced and the tube plugging criteria for the TPCCW heat exchanger of Yonggwang NPP No. 1 and 2. This method relies on the similar plugging criteria used in the steam generator

  13. Extending service life of steam generators by sleeving tubes

    International Nuclear Information System (INIS)

    Gutzwiller, J.E.

    1982-01-01

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

  14. Automated WWER steam generator eddy current testing and plugging control system

    International Nuclear Information System (INIS)

    Gorecan, I.; Gortan, K.; Grzalja, I.

    2004-01-01

    The structural architecture of the system contains three main components which are described as follows: Manipulator Guidance System; Eddy Current Testing System; Plugging System. The manipulator system has the task to position the end-effectors to the desired tube position. When the final position is reached, the Eddy Current testing system performs data acquisition. In case defects are found, the plugging system performs tube plug installment. Each system is composed of 3 layers. The first layer is the hardware layer consisting of motors driving the effectors along with sensors needed to obtain the positioning data, pusher motors used to push the test probes into tubes of the WWER steam generator, and plugging hardware tool. The second layer is the control box performing basic monitoring and control routines as an interconnection between first and third layer. The highest layer is the control software, running on the PC, which is used as a human-machine-interface.(author)

  15. Plugger guide for aligning an end plug and a fuel rod tube end

    International Nuclear Information System (INIS)

    Klapper, K.K.; Boatwright, D.A.

    1987-01-01

    A pin driving tool is described for inserting or removing pins from teeth on a digging means, comprising: fuel rod tube toward an end plug for application of the end plug into the tube end, the apparatus comprising: (a) a guide housing having an elongated central longitudinal bore with one end for receiving the end plug and an opposite end for receiving the fuel rod tube end; (b) sets of rolling elements disposed in the housing at axially spaced positions along and about the bore thereof. The rolling elements in each set are positioned in fixed relation with respect to one another to receive the fuel rod tube end therebetween and align the tube end with the end plug as the tube end is moved through the bore and into engagement with the end plug; and (c) retaining means disposed adjacent to the open end of the housing bore for engaging the end plug so as to maintain it in a stationary seated position at the one end of the housing bore

  16. Plugging of feed inlet tube upstands with Ni/Ti shape memory alloy plugs - Heysham 1 power station

    International Nuclear Information System (INIS)

    Mathews, A.J.

    1988-01-01

    The paper contains a description of a new approach for Plugging feed inlet tubes of Gas-Cooled Reactors. Instead of utilizing the original explosive method plugging by fitting a shape memory alloy plug into the upstand is being described. (author)

  17. 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

  18. 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

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

    International Nuclear Information System (INIS)

    Boehm, H.H.; Foerch, H.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Glass, S.W.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Glass, S.W.

    1985-01-01

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

  2. 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

  3. 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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  5. Lessons learned from tubes pulled from French steam generators

    International Nuclear Information System (INIS)

    Berge, Ph.; Boursier, J.M.; Dallery, D.; De Keroulas, F.; Rouillon, Y.

    1998-01-01

    Since 1981, the Chinon Hot Laboratory has completed more than 380 metallurgical examinations of pulled French steam generator tubes. Electricite de France decided to perform such investigations from the very outset of the French nuclear program, in order to contribute to nuclear power plant safety. The main reasons for withdrawing tubes are to evaluate the degradation, to validate non destructive examination (NDE) techniques, to gain a better understanding of cracking phenomena, and to ensure that the criteria on which plugging operations are based remain conservative. Considerable experience has been accumulated in the field of primary water stress corrosion cracking (PWSCC), OD (secondary) side corrosion, leak and burst tests, and various tube plugging techniques. This paper focuses on the PWSCC phenomenon and on the secondary side corrosion process, and in particular, attempts to correlate French data from pulled tubes with the results of fundamental R and D studies. Finally, within the framework of the Nuclear Power Plant Safety and Maintenance Policy, all these results are discussed in terms of optimization of the field inspection of tube bundles and plugging criteria. (author)

  6. Integrity Analysis of Damaged Steam Generator Tubes

    International Nuclear Information System (INIS)

    Stanic, D.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-12-01

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

  8. Apparatus and method for applying an end plug to a fuel rod tube end

    International Nuclear Information System (INIS)

    Rieben, S.L.; Wylie, M.E.

    1987-01-01

    An apparatus is described for applying an end plug to a hollow end of a nuclear fuel rod tube, comprising: support means mounted for reciprocal movement between remote and adjacent positions relative to a nuclear fuel rod tube end to which an end plug is to be applied; guide means supported on the support means for movement; and drive means coupled to the support means and being actuatable for movement between retracted and extended positions for reciprocally moving the support means between its respective remote and adjacent positions. A method for applying an end plug to a hollow end of a nuclear fuel rod tube is also described

  9. Corrosion and Rupture of Steam Generator Tubings in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-15

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

  10. Corrosion and Rupture of Steam Generator Tubings in PWRs

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-01

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

  11. Motion of liquid plugs between vapor bubbles in capillary tubes: a comparison between fluids

    Science.gov (United States)

    Bertossi, Rémi; Ayel, Vincent; Mehta, Balkrishna; Romestant, Cyril; Bertin, Yves; Khandekar, Sameer

    2017-11-01

    Pulsating heat pipes (PHP) are now well-known devices in which liquid/vapor slug flow oscillates in a capillary tube wound between hot and cold sources. In this context, this paper focuses on the motion of the liquid plug, trapped between vapor bubbles, moving in capillary tubes, to try to better understand the thermo-physical phenomena involved in such devices. This study is divided into three parts. In the first part, an experimental study presents the evolution of the vapor pressure during the evaporation process of a liquid thin film deposited from a liquid plug flowing in a heated capillary tube: it is found that the behavior of the generated and removed vapor can be very different, according to the thermophysical properties of the fluids. In the second part, a transient model allows to compare, in terms of pressure and duration, the motion of a constant-length liquid plug trapped between two bubbles subjected to a constant difference of vapor pressure: the results highlight that the performances of the four fluids are also very different. Finally, a third model that can be considered as an improvement of the second one, is also presented: here, the liquid slug is surrounded by two vapor bubbles, one subjected to evaporation, the pressure in both bubbles is now a result of the calculation. This model still allows comparing the behaviors of the fluid. Even if our models are quite far from a complete model of a real PHP, results do indicate towards the applicability of different fluids as suitable working fluids for PHPs, particularly in terms of the flow instabilities which they generate.

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

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1996-01-01

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

  13. Guidelines for Selecting Plugs Used in Thin-Walled Tube Drawing Processes of Metallic Alloys

    Directory of Open Access Journals (Sweden)

    Eva María Rubio

    2017-12-01

    Full Text Available In this paper, some practical guidelines to select the plug or set of plugs more adequate to carry out drawing processes of thin-walled tubes carried out with fixed conical inner plug are presented. For this purpose, the most relevant input parameters have been considered in this study: the tube material, the most important geometrical parameters of the process (die semiangle, α , and cross-sectional area reduction, r and the friction conditions (Coulomb friction coefficients, μ 1 , between the die and the tube outer surface, and μ 2 , between the plug and the tube inner surface. Three work-hardening materials are analyzed: the annealed copper UNS C11000, the aluminum UNS A91100, and the stainless steel UNS S34000. The analysis is realized by means of the upper bound method (UBM, modelling the plastic deformation zone by triangular rigid zones (TRZ, under the validated assumption that the process occurs under plane strain conditions. The obtained results allow establishing, for each material, a group of geometrical parameters, friction conditions, a set of plugs that make possible to carry out the process under good conditions, and the optimum plug to carry out the process using the minimum amount of energy. The proposed model is validated by means of an own finite element analysis (FEA carried out under different conditions and, in addition, by other finite element method (FEM simulations and real experiments taken from other researchers found in the literature (called literature simulations and literature experimental results, respectively. As a main conclusion, it is possible to affirm that the plug that allows carrying out the process with minimum quantity of energy is cylindrical in most cases.

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

    International Nuclear Information System (INIS)

    Dvorsek, T.; Cizelj, L.

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  17. WWER Steam Generators Tubing Performance and Aging Management

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  18. Friction stir welding of F/M ODS steel plug and F/M steel tube

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Suk Hoon, E-mail: shkang77@kaeri.re.kr [Nuclear Materials Division, Korea Atomic Energy Research Institute (Korea, Republic of); Vasudevan, M. [Materials Technology Division, Indira Gandhi Centre for Atomic Research, Kalpakkam (India); Noh, Sanghoon; Jin, Hyun Ju; Jang, Jinsung; Kim, Tae Kyu [Nuclear Materials Division, Korea Atomic Energy Research Institute (Korea, Republic of)

    2016-11-01

    Highlights: • Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. • The curvature and smaller thickness of tube was the major limitation for applying FSW method, it was solved using specially designed jig. • Considerable hardening occurs in the joint because the cooling rate was sufficient to reproduce a martensitic microstructure. • The measured hoop strength of the FSWed joint was 70–90 MPa, the value was at around 70% of the tube. - Abstract: Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. The dimensions of the tube included outer diameter of 7 mm, wall thickness of 0.5 mm. The objective was to find suitable process variables for gaining enough frictional heat from those thin and curved pieces. A specially designed jig was used for stabilization and slow rotation of tube during FSW. Additionally, the plug was designed to overlap the tube. Inconel 718 was used as FSW tool, the diameter was 3.5 mm. The adequate rotation speed of the tool and jig were 1200 rpm and 1.5 rpm, respectively. The joining was successfully accomplished using above combination, showing a good possibility. The hoop stress tests of joint were conducted by blowing Ar gas into the tube, the flow rate of gas was 10 MPa/min. The measured hoop stress was 70–90 MPa, the value was at around 70% of the tube.

  19. Friction stir welding of F/M ODS steel plug and F/M steel tube

    International Nuclear Information System (INIS)

    Kang, Suk Hoon; Vasudevan, M.; Noh, Sanghoon; Jin, Hyun Ju; Jang, Jinsung; Kim, Tae Kyu

    2016-01-01

    Highlights: • Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. • The curvature and smaller thickness of tube was the major limitation for applying FSW method, it was solved using specially designed jig. • Considerable hardening occurs in the joint because the cooling rate was sufficient to reproduce a martensitic microstructure. • The measured hoop strength of the FSWed joint was 70–90 MPa, the value was at around 70% of the tube. - Abstract: Friction stir welding (FSW) was used for joining of oxide dispersion strengthened (ODS) steel plug and F/M steel tube. The dimensions of the tube included outer diameter of 7 mm, wall thickness of 0.5 mm. The objective was to find suitable process variables for gaining enough frictional heat from those thin and curved pieces. A specially designed jig was used for stabilization and slow rotation of tube during FSW. Additionally, the plug was designed to overlap the tube. Inconel 718 was used as FSW tool, the diameter was 3.5 mm. The adequate rotation speed of the tool and jig were 1200 rpm and 1.5 rpm, respectively. The joining was successfully accomplished using above combination, showing a good possibility. The hoop stress tests of joint were conducted by blowing Ar gas into the tube, the flow rate of gas was 10 MPa/min. The measured hoop stress was 70–90 MPa, the value was at around 70% of the tube.

  20. A study on development of a plugging margin evaluation method taking into account the fouling of shell-and tube heat exchangers

    International Nuclear Information System (INIS)

    Hwang, Kyeong Mo; Jin, Tae Eun; Kim, Kyung Hoon

    2006-01-01

    As the operating time of heat exchangers progresses, fouling caused by water-borne deposits and the number of plugged tubes increase and thermal performance decreases. Both fouling and tube plugging are known to interfere with normal flow characteristics and to reduce thermal efficiencies of heat exchangers. The heat exchangers of Korean nuclear power plants have been analyzed in terms of heat transfer rate and overall heat transfer coefficient as a means of heat exchanger management. Except for fouling resulting from the operation of heat exchangers, all the tubes of heat exchangers have been replaced when the number of plugged tubes exceeded the plugging criteria based on design performance sheet. This paper describes a plugging margin evaluation method taking into account the fouling of shell-and-tube heat exchangers. The method can evaluate thermal performance, estimate future fouling variation, and consider current fouling level in the calculation of plugging margin. To identify the effectiveness of the developed method, fouling and plugging margin evaluations were performed at a component cooling heat exchanger in a Korean nuclear power plant

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

    International Nuclear Information System (INIS)

    Silveira, Helvecio Carlos Klinke da

    2002-01-01

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

  2. Reliability of eddy current examination of steam generator tubes

    International Nuclear Information System (INIS)

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

    1985-04-01

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

  3. The influence of mechanical clearance reduction between plug (m1- m.e. version) and tube sheet simulator, at its experimental installation using a high speed plastic deformation technique

    International Nuclear Information System (INIS)

    Gyongyosi, T.; Panaitescu, V. N.; Prisecaru, I.

    2013-01-01

    The steam generator has to keep its structural integrity during entire operation lifetime. The isolation at both ends of the defective tube, from the primary heat transport system, is made of special shape plugs, using a plastic deformation technique. In the paper, after a brief introduction, are presented the technique developed experimentally for plugging of the steam generator defective tubes followed by a description of the tests performed, the results analysis get and conclusions. The paper is dedicated to the specialists working in the research and technological engineering. (authors)

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

    International Nuclear Information System (INIS)

    Stevens-Guille, P.D.

    1975-01-01

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

  5. AN INVESTIGATION OF THE EFFECT OF THE HOT END PLUGS ON THE EFFICIENCY OF THE RANQUE-HILSCH VORTEX TUBE

    Directory of Open Access Journals (Sweden)

    MAZIAR ARJOMANDI

    2007-12-01

    Full Text Available The phenomenon of temperature distribution in confined steady rotating gas flows is called Ranque-Hilsch effect. The simple counter-flow vortex tube consists of a long hollow cylinder with tangential nozzle at one end for injecting compressed air. The flow inside the vortex tube can be described as rotating air, which moves in a spring-shaped vortex track. The peripheral flow moves toward the hot end where a hot end plug is placed and the axial flow, which is forced back by the plug, moves in the opposite direction toward the cold end. This paper focuses on the effect of the size of hot nozzle on the performance of the Ranque–Hilsch vortex tube. Series of plugs were used in the experiment in order to find the relationship between the diameter of hot end plug and the performance of the vortex tube.

  6. Research perspectives on the evaluation of steam generator tube integrity

    International Nuclear Information System (INIS)

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

    2001-01-01

    Industry effects have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, SCC and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by plug or repair on detection, because current NDE techniques for characterization of flaws are not accurate enough to permit continued operation. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators

  7. 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.

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

    International Nuclear Information System (INIS)

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

    1980-02-01

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

  9. Nuclear steam generator tube to tubesheet joint optimization

    International Nuclear Information System (INIS)

    McGregor, Rod

    1999-01-01

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

  10. From catastrophic acceleration to deceleration of liquid plugs in prewetted capillary tubes

    Science.gov (United States)

    Magniez, Juan; Baudoin, Michael; Zoueshtiagh, Farzam; Lemac/Lics Team

    2016-11-01

    Liquid/gas flows in capillaries are involved in a multitude of systems including flow in porous media, petroleum extraction, imbibition of paper or flows in pulmonary airways in pathological conditions. Liquid plugs, witch compose the biphasic flows, can have a dramatic impact on patients with pulmonary obstructive diseases, since they considerably alter the circulation of air in the airways and thus can lead to severe breathing difficulties. Here, the dynamics of liquid plugs in prewetted capillary tube is investigated experimentally and theoretically, with a particular emphasis on the role of the prewetting films and of the driving condition (constant flow rate, constant pressure). For both driving conditions, the plugs can either experience a continuous increase or decrease of their size. While this phenomenon is regular in the case of imposed flow rate, a constant pressure head can lead to a catastrophic acceleration of the plug and eventually its rupture or a dramatic increase of the plug size. A theoretical model is proposed to explain the transition between theses two regimes. These results give a new insight on the critical pressure required for airways obstruction and reopening. IEMN, International Laboratory LEMAC/LICS, UMR CNRS 8520, University of Lille.

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

    International Nuclear Information System (INIS)

    Park, Jun Su; Jeong, Seung Ha

    1996-06-01

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

  12. Rotary plug seal

    International Nuclear Information System (INIS)

    Ito, Koji; Abiko, Yoshihiro.

    1981-01-01

    Purpose: To enable fuel exchange even upon failure of regular seals and also to enable safety seal exchange by the detection of the reduction in the contact pressure of a rotary plug seal. Constitution: If one of a pair of regular tube seals for the rotary plug is failed during ordinary operation of a FBR type reactor, the reduction in the contact pressure of the seal to the plug gibbousness is detected by a pressure gauge and a solenoid valve is thereby closed. Thus, a back-up-tube seal provided above or below the tube seal is press-contacted by way of argon gas to the gibbousness to enter into operation state and lubricants are supplied from an oil tank. In such a structure, the back-up-tube seal is operated before the failure of the tube seal to enable to continue the fuel exchange work, as well as safety exchange for the tube seal. (Moriyama, K.)

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

    Scheveneels, G.

    1997-01-01

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

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

    International Nuclear Information System (INIS)

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

    2009-01-01

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

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  17. 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)

  18. Analysis of steam generator plugging on core thermohydraulic performance of NPP Krsko; Analiza vpliva cepljenja cevi v upravljaniku na termohidravliko sredice JE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Kostadinov, V; Petelin, S; Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1988-07-01

    Nuclear safety analysis of NPP Krsko core operating at full power with 4% steam generator tubes plugged have been performed. Influence of individual parameters on core thermohydraulic performance have been evaluated. Using COBRA-III-C computer code we have analysed a core design (evaluation) model. The DNBR change was calculated as a consequence of 4% plugging. The influence of thermohydraulic parameters change on DNBR was analysed. (author)

  19. Assesment of integrity of WWER steam generator tubes

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. 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

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  2. Probability of a steam generator tube rupture due to the presence of axial through wall cracks

    International Nuclear Information System (INIS)

    Mavko, B.; Cizelj, L.

    1991-01-01

    Using the Leak-Before-Break (LBB) approach to define tube plugging criteria a possibility to operate with through wall crack(s) in steam generator tubes may be considered. This fact may imply an increase in tube rupture probability. Improved examination techniques (in addition to the 100% tube examination) have been developed and introduced to counterbalance the associated risk. However no estimates of the amount of total increase or decrease of risk due to the introduction of LBB have been made. A scheme to predict this change of risk is proposed in the paper, based on probabilistic fracture mechanics analysis of axial cracks combined with available data of steam generator tube nondestructive examination reliability. (author)

  3. 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.

  4. 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.

  5. 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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  8. Technical basis for the CANDU steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

    Kozluk, M.J.; Scarth, D.A.; Graham, D.B.

    2002-01-01

    Active degradation mechanisms in steam generators and preheaters in Canadian CANDU T M generating stations are managed through Steam Generator Programs that incorporate tube inspection, maintenance (cleaning), fitness-for-service assessment, and preventative plugging as part of the overall steam generator management strategy. Steam generator and preheater tubes are inspected in accordance with the CSA Standard CAN/CSA-N285.4-94[l]. When a detected flaw indication does not satisfy the criteria of acceptance by examination, CSA-N285.4-94 permits a fitness-for-service assessment to determine acceptability. In 1999 Ontario Power Generation issued, for trial use, fitness-for-service guidelines for steam generator and preheater tubes in CANDU nuclear power plants. The main objectives of the Fitness-for-Service Guidelines are to provide reasonable assurance that tube structural integrity is maintained, and to provide reasonable assurance that there are adequate margins between estimated accumulated dose and applicable site dose limits. The Fitness-for-Service Guidelines are intended to provide industry-standard acceptance criteria and evaluation procedures for assessing the condition of steam generator and preheater tubes in terms of tube structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. This paper describes the technical basis for the minimum required safety factors specified in Table IC-1 of the Fitness-for-Service Guidelines and for the flaw models used to develop the flaw stability requirements in the nonmandatory, Appendix C of the Fitness-for-Service Guidelines. (author)

  9. Evaluation of sampling plans for in-service inspection of steam generator tubes

    International Nuclear Information System (INIS)

    Kurtz, R.J.; Heasler, P.G.; Baird, D.B.

    1994-02-01

    This report summarizes the results of three previous studies to evaluate and compare the effectiveness of sampling plans for steam generator tube inspections. An analytical evaluation and Monte Carlo simulation techniques were the methods used to evaluate sampling plan performance. To test the performance of candidate sampling plans under a variety of conditions, ranges of inspection system reliability were considered along with different distributions of tube degradation. Results from the eddy current reliability studies performed with the retired-from-service Surry 2A steam generator were utilized to guide the selection of appropriate probability of detection and flaw sizing models for use in the analysis. Different distributions of tube degradation were selected to span the range of conditions that might exist in operating steam generators. The principal means of evaluating sampling performance was to determine the effectiveness of the sampling plan for detecting and plugging defective tubes. A summary of key results from the eddy current reliability studies is presented. The analytical and Monte Carlo simulation analyses are discussed along with a synopsis of key results and conclusions

  10. Structural integrity assessments of steam generator tubes using the FAD methodology

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  11. Structural integrity assessments of steam generator tubes using the FAD methodology

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  12. Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

    International Nuclear Information System (INIS)

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

    2006-01-01

    It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

  13. Fuel rod end plug

    International Nuclear Information System (INIS)

    McGeary, R.K.; Bucher, G.D.

    1989-01-01

    This patent describes an end plug for welded disposition within the end of a tube. It comprises a circumferentially extending, axially oriented land surface, having a radial extent defined by means of a first predetermined dimension, for disposition within the end of the tube; a circumferentially extending, axially oriented land surface, having a radial extent defined by means of a second predetermined dimension which is greater than the first predetermined dimension, for disposition outside of the end of the tube. The second land surface being disposed upstream of the first land surface; an annularly extending, radially oriented shoulder portion, defined at the downstream end of the second land surface and having a radially inward depth which is greater than the difference defined between the first and second radial dimensions of the first and second land surfaces, for engaging the end of the tube in a butt contact fashion; and annular groove means defined between the upstream end of the first land surface and the shoulder portion of the end plug, for eliminating porosity defects normally developed within a weldment defined between the tube end and the end plug when the end plug is welded within the tube end, and including a conical surface which extends radially outwardly from the innermost radial depth extent of the shoulder portion to the upstream end of the first land surface

  14. 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.

  15. 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)

  16. Depressurization experiments on a plugged fibrous insulation in a horizontal pressure tube

    International Nuclear Information System (INIS)

    Lang, H.; Weise, H.J.; Ennen, P.

    1977-08-01

    Hot gas ducts for high-temperature reactors with a helium turbine are subject to additional operational loads not caused by the gas temperature. They include vibrations, caused by high gas velocities or by the sound fields emitted from the turbine, and stresses, originating from fast, short-time pressure changes. Such pressure changes occur as a rule if the generator coupled with the turbine has to be disconnected from the grid. In order to avoid no-load operation of the turbine a bypass between HP and LP side of the turbine is opened. As a consequence of this measure a sudden pressure drop occurs in the free flow cross-section causing differential pressures within the insulation. As the size of these differential pressures depends on the insulating material, the density of plugging, the kind of internals, and on the position and size of the depressurization borings, the pressure distributions in the insulation were measured on a test tube for the HP channel. (orig./RW) [de

  17. Research perspectives on the evaluation of steam generator tube integrity

    International Nuclear Information System (INIS)

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

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  20. Inhibition of Microbial Growth by Fatty Amine Catalysts from Polyurethane Foam Test Tube Plugs

    Science.gov (United States)

    Bach, John A.; Wnuk, Richard J.; Martin, Delano G.

    1975-01-01

    When polyurethane foam test tube plugs are autoclaved, they release volatile fatty amines that inhibit the growth of some microorganisms. The chemical structures of these amines were determined by the use of a gas chromatographmass spectrometer. They are catalysts used to produce the foam. The problem of contaminating growth media with toxic substances released from polymeric materials is discussed. PMID:1096816

  1. Eddy current technology for heat exchanger and steam generator tube inspection

    Energy Technology Data Exchange (ETDEWEB)

    Obrutsky, L.; Lepine, B.; Lu, J.; Cassidy, R.; Carter, J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2004-07-01

    A variety of degradation modes can affect the integrity of both heat exchanger (HX) and balance of plant tubing, resulting in expensive repairs, tube plugging or replacement of tube bundles. One key component for ensuring tube integrity is inspection and monitoring for detection and characterization of the degradation. In-service inspection of HX and balance of plant tubing is usually carried out using eddy current (EC) bobbin coils, which are adequate for the detection of volumetric degradations. However, detection and quantification of additional modes of degradation such as pitting, intergranular attack (IGA), axial cracking and circumferential cracking require specialized probes. The need for timely, reliable detection and characterization of these modes of degradation is especially critical in Nuclear Generating Stations. Transmit-receive single-pass array probes, developed by AECL, offer high defect detectability in conjunction with fast and reliable inspection capabilities. They have strong directional properties, permitting probe optimization for circumferential or axial crack detection. Compared to impedance probes, they offer improved performance in the presence of variable lift-off. This EC technology can help resolve critical detection issues at susceptible areas, such as the rolled-joint transitions at the tubesheet, U-bends and tube-support intersections. This paper provides an overview of the operating principles and the capabilities of advanced ET inspection technology available for HX tube inspection. Examples of recent application of this technology in Nuclear Generating Stations (NGSs) are discussed. (author)

  2. Eddy current technology for heat exchanger and steam generator tube inspection

    International Nuclear Information System (INIS)

    Obrutsky, L.; Lepine, B.; Lu, J.; Cassidy, R.; Carter, J.

    2004-01-01

    A variety of degradation modes can affect the integrity of both heat exchanger (HX) and balance of plant tubing, resulting in expensive repairs, tube plugging or replacement of tube bundles. One key component for ensuring tube integrity is inspection and monitoring for detection and characterization of the degradation. In-service inspection of HX and balance of plant tubing is usually carried out using eddy current (EC) bobbin coils, which are adequate for the detection of volumetric degradations. However, detection and quantification of additional modes of degradation such as pitting, intergranular attack (IGA), axial cracking and circumferential cracking require specialized probes. The need for timely, reliable detection and characterization of these modes of degradation is especially critical in Nuclear Generating Stations. Transmit-receive single-pass array probes, developed by AECL, offer high defect detectability in conjunction with fast and reliable inspection capabilities. They have strong directional properties, permitting probe optimization for circumferential or axial crack detection. Compared to impedance probes, they offer improved performance in the presence of variable lift-off. This EC technology can help resolve critical detection issues at susceptible areas, such as the rolled-joint transitions at the tubesheet, U-bends and tube-support intersections. This paper provides an overview of the operating principles and the capabilities of advanced ET inspection technology available for HX tube inspection. Examples of recent application of this technology in Nuclear Generating Stations (NGSs) are discussed. (author)

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

    International Nuclear Information System (INIS)

    Michaut, B.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    1986-06-01

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

  5. 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

  6. Evaluation of candidate photomultiplier tubes for the upgrade of the CDF end plug calorimeter

    International Nuclear Information System (INIS)

    Koska, W.; Delchamps, S.W.; Freeman, J.; Kinney, W.; Lewis, D.; Limon, P.; Strait, J.; Fiori, I.; Gallinaro, M.; Shen, Q.

    1994-01-01

    The Collider Detector at Fermilab is upgrading its end plug calorimeter from a gas detector system to one using scintillating tiles and wavelength shifting fibers. This tile-fiber calorimeter will be read out through 1,824 photomultiplier tubes. The performance requirements of the calorimeter require that the PMTs have good response to light in the 500 nm region, provide adequate amplification for signals from minimum ionizing particles yet provide linear response for peak anode currents up to 25 mA at a gain of 50,000, and fit into the restricted space at the rear of the plugs. This paper will describe the evaluation process used to determine the adequacy of the commercially available PMTs which appeared to meet these performance requirements

  7. Damage mechanisms and estimation of the frequency of leaks of steam generator tubes in German PWRs

    International Nuclear Information System (INIS)

    Reck, H.

    1992-01-01

    Operating experience of steam generator tubes in German PWRs has shown that so far there have only relatively few cases of damage been registered. The only steam generators with a high failure rate were exchanged in 1983. The material of the affected tubes was Inconel 600. The types of failure that occurred in the late 70's and early 80's were mainly wastage corrosion, which was thought to be the result of phosphate operating. After optimising the water chemistry and changing to ''high AVT'' operating, the failure rate decreased considerably. In total, about 973 of the 193335 tubes that were in operation were plugged because of wall-thinning or leaks. There have been 6 leaks, with the highest leakage volume being 40 liters per hour. 7 refs., 6 figs., 6 tabs

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

    International Nuclear Information System (INIS)

    2007-12-01

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

  9. 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

  10. 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)

  11. 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...

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

    International Nuclear Information System (INIS)

    Xiongzi, Li; Zhongxue, Gan; Lance, Fitzgibbons

    2001-01-01

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

  13. Development and application of an efficient method for performing modal analysis of steam generator tubes in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Huinam [Dept of Mechanical and Aerospace Engineering, Sunchon National University, Sunchon, 540-742 (Korea, Republic of); Boo, Myung-Hwan [Korea Hydro and Nuclear Power Company, Yuseong-Gu, Daejeon 305-343 (Korea, Republic of); Park, Chi-Yong [KEPCO Research Institute, Yuseong-Gu, Daejeon 305-380 (Korea, Republic of); Ryu, Ki-Wahn, E-mail: kwryu@chonbuk.ac.k [Department of Aerospace Engineering, Chonbuk National University, 664-14, Deogjin-Dong, Jeonju 561-756 (Korea, Republic of)

    2010-10-15

    with the FIV analysis program, it gives an amazing benefit, which makes the FIV analysis of all tubes possible within a limited time during the design or maintenance period. Using the developed program, the stability ratio regarding the fluid-elastic instability and the amplitude of vibration resulting from the turbulence flow excitation can be calculated for all tubes according the standard ASME Code, therefore, much more reliable design of the steam generator against the FIV related failures can be achieved. For an operating plant, there is a requirement that every single tube must be quantitatively checked whether wear would be more than 40% in thickness during the next operation period. If yes, the tube must be plugged to prevent severe failure including the tube wall penetration. In order to decide the tube plugging, which leads the plant performance degradation, the wear prediction is required. The wear prediction analysis needs wear data by ultrasonic testing and modal property data of tubes. PIAT-MODE will be a power tool enabling the wear prediction in a limited maintenance period by supplying modal data for all tubes.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  16. Steam generator tube integrity program

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  17. 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.

  18. Evaluating remote control and robotics actions in NPPs in an ALARA perspective: lessons from the evolution of steam generator tube plugging technique in France

    International Nuclear Information System (INIS)

    Lefaure, C.; Lochard, J.; Blain, A.

    1989-01-01

    Introducing remote tooling and robotics in NPPs raises many questions. Is it better to develop specific robotic tools for each type of operation, or to conceive more generic multipurpose tool carriers. Does the introduction of remote control and robotics always reduce human involvement and collective doses. What are the impacts on individual doses distribution for the most exposed operators. Under what conditions is robotics justified from an economical point of view. Even if robotics and remote tooling actions can reduce significantly individual and collective exposures, because of their high costs of development, they will be adopted only if simultaneously the potential for operational cost savings is also clearly demonstrated. Integrating operational cost savings in the cost evaluation of the ALARA procedure is important to compare and select among different robotics and remote control actions. This paper presents how to evaluate in a systematic way dose reductions, operational costs savings as well as the conditions under which robotics and remote control actions can effectively improve occupational radiation protection in NPPs. The demonstration is based on an analysis of the evolution of the steam generator tube plugging technique developed in French PWRs

  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. Forming a cohesive steam generator maintenance strategy

    International Nuclear Information System (INIS)

    Poudroux, G.

    1991-01-01

    In older nuclear plants, steam generator tube bundles are the most fragile part of the reactor coolant system. Steam generator tubes are subject to numerous types of loading, which can lead to severe degradation (corrosion and wear phenomena). Preventive actions, such as reactor coolant temperature reduction or increasing the plugging limit and their associated analyses, can increase steam generator service life. Beyond these preventive actions, the number of affected tubes and the different locations of the degradations that occur often make repair campaigns necessary. Framatome has developed and qualified a wide range of treatment and repair processes. They enable careful management of the repair campaigns, to avoid reaching the maximum steam generator tube plugging limit, while optimizing the costs. Most of the available repair techniques allow a large number of affected tubes to be treated. Here we look only at those techniques that should be taken into account when defining a maintenance strategy. (author)

  1. End plug for fuel rod and welding method therefor

    International Nuclear Information System (INIS)

    Yoneda, Hiroshi; Murakami, Kazuo; Oyama, Jun-ichi.

    1996-01-01

    An end plug of a fuel rod comprises a pressure-insertion portion having a diameter somewhat greater than the inner diameter of a fuel cladding tube and a welding portion having a diameter substantially the same as the outer diameter of the cladding tube. A V-shaped recess having an outer diameter smaller than the greatest outer diameter of the pressure-insertion portion is formed over the entire circumferential surface of the outer circumference of the connection portion of the pressure-insertion portion and the welding portion. The pressure-insertion portion of the end plug is inserted to the end of the cladding tube till the end of the cladding tube abuts against the inclined surface of the V-shaped recess. The abutting surfaces of the end plug and the cladding tube are subjected to resistance welding in this state. The inner portion bulged from the inclined surface of the V-shaped recess is filled in the recess in a molten state. Lowering of temperature of the cladding tube in the vicinity of the welded portion is decreased by γ heat during reactor operation. Accordingly, lowering of ductility of the cladding tube and degradation of material of the welded region due to segregation of hydrogen in the cladding tube can be suppressed. (I.N.)

  2. 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.)

  3. 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

  4. Dialysis grafts arterial plug: Retrieval using the tulip sheath device in vitro

    International Nuclear Information System (INIS)

    Sharafuddin, Melhem J.; Titus, Jack L.; Gu Xiaoping; Hunter, David W.; Amplatz, Kurt

    1997-01-01

    The 'arterial plug' is a resistant thrombus that frequently persists at the arterial anastomosis of clotted hemodialysis grafts following thrombolytic therapy. We studied the physical and morphological characteristics of the plug and determined the feasibility of transcatheter removal in vitro using the tulip compression thrombectomy system. Sixteen thrombus plugs were recovered during surgical thrombectomy of clotted human dialysis grafts. The physical and gross physical characteristics of all plugs were analyzed. Eight specimens were evaluated microscopically. Transcatheter compression thrombectomy of eight plugs was attempted in vitro. Each plug was embedded in a polyvinyl tube filled with newly clotted blood and connected to a flow circuit. First, balloon-assisted aspiration thrombectomy (BAT) of soft thrombus was performed, while sparing the distal-most segment containing the plug. The tulip sheath was then introduced facing the 'arterial end' of the tube. The thrombus segment containing the plug was pulled back into the tulip mesh using either a 3 Fr Fogarty balloon catheter or a self-expanding rake. The tulip was closed to compress and remove the trapped plug. Near-complete thrombectomy of soft clot was achieved in all tested tubes. Compression and retrieval of the entire arterial plug was successful in all except one, where only partial compression of the plug occurred, presumably due to fibrotic changes. No fragmentation or embolization occurred in the remaining procedures. Spongy consistency was noted in 94% of the specimens. Microscopic evaluation showed organized layered thrombus with compaction in five plugs. Transcatheter removal of a thrombus plug is feasible in vitro using the tulip compression-thrombectomy system

  5. 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

  6. 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)

  7. Obturation and holding back device for a leakproof closure plug for a steam generator tube

    International Nuclear Information System (INIS)

    Lenoble, R.

    1991-01-01

    This leak proof closure plug has a bolt with at least a transversal aperture perpendicular to its axis. In the aperture is jammed a bracking piece. When screwing the bolt in the expander, the extremity part of the bracking piece is jammed in the tapped hole of the expander [fr

  8. Small leak shutdown, location, and behavior in LMFBR steam generators

    International Nuclear Information System (INIS)

    Sandusky, D.W.

    1976-01-01

    The paper summarizes an experimental study of small leaks tested under LMFBR steam generator conditions. Defected tubes were exposed to flowing sodium and steam. The observed behavior of the defected tubes is reported along with test results of shutdown methods. Leak location methods were investigated. Methods were identified to open plugged defects for helium leak testing and detect plugged leaks by nondestructive testing

  9. PWR steam generator tubing sample library

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

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

  10. 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)

  11. 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

  12. A balanced strategy in managing steam generator thermal performance

    International Nuclear Information System (INIS)

    Hu, M. H.; Nelson, P. R.

    2009-01-01

    This paper presents a balanced strategy in managing thermal performance of steam generator designed to deliver rated megawatt thermal (MWt) and megawatt electric (MWe) power without loss with some amount of thermal margin. A steam generator (SG) is a boiling heat exchanger whose thermal performance may degrade because of steam pressure loss. In other words, steam pressure loss is an indicator of thermal performance degradation. Steam pressure loss is mainly a result of either 1) tube scale induced poor boiling or 2) tube plugging historically resulting from tubing corrosion, wear due to flow induced tube vibration or loose parts impact. Thermal performance degradation was historically due to tube plugging but more recently it is due to poor boiling caused by more bad than good constituents of feedwater impurities. The whole SG industry still concentrates solely on maintenance programs towards preventing causes for tube plugging and yet almost no programs on maintaining adequate boiling of fouled tubes. There can be an acceptable amount of tube scale that provides excellent boiling capacity without tubing corrosion, as operational experience has repeatedly demonstrated. Therefore, future maintenance has to come up balanced programs for allocating limited resources in both maintaining good boiling capacity and preventing tube plugging. This paper discusses also thermal performance degradation due to feedwater impurity induced blockage of tube support plate and thus subsequent water level oscillations, and how to mitigate them. This paper provides a predictive management of tube scale for maintaining adequate steam pressure and stable water level without loss in MWt/MWe or recovering from steam pressure loss or water level oscillations. This paper offers a balanced strategy in managing SG thermal performance to fulfill its mission. Such a strategy is even more important in view of the industry trend in pursuing extended power uprate as high as 20 percent

  13. Steam generators regulatory practices and issues in Spain

    International Nuclear Information System (INIS)

    Mendoza, C.; Castelao, C.; Ruiz-Colino, J.; Figueras, J.M.

    1997-01-01

    This paper presents the actual status of Spanish Steam Generator tubes, actions developed by PWR plant owners and submitted to CSN, and regulatory activities related to tube degradation mechanisms analysis; NDT tube inspection techniques; tube, tubesheet and TSPs integrity studies; tube plugging/repair criteria; preventive and corrective measures including whole SGs replacement; tube leak measurement methods and other operational aspects

  14. Steam generator operation and maintenance

    International Nuclear Information System (INIS)

    Lee, C.K.

    2004-01-01

    Corrosion of steam generator tube has resulted in the need for extensive repair and replacement of steam generators. Over the past two decades, steam generator problems in the United States were viewed to be one of the most significant contributor to lost generation in operating PWR plants. When the SGOG-I (Steam Generator Owners Groups) was formed in early 1977, denting was responsible for almost 90% of the tube plugging. By the end of 1982, this figure was reduced to less than 2%. During the existence of SGOG-II (from 1982 to 1986), IGA/SCC (lntergranular Attack/Stress Corrosion Cracking) in the tube sheet, primary side SCC, pitting, and fretting surfaced as the primary causes of tube degradation. Although significant process has been made with wastage and denting, the utilities experience shows that the percentage of reactors plugging tubes and the percentage of tubes being plugged each year has remained relatively constant. The diversity of the damage mechanisms means that no one solution is likely to resolve all problems. The task of maintaining steam generator integrity continues to be formidable and challenging. As the older problems were brought under control, many new problems emerged. SGOG-II (Steam Generator Owners Group program from 1982 to 1986) has focused on these problem areas such as tube stress corrosion cracking (SCC) and intergranular attack (IGA) in the open tube sheet crevice, primary side tube cracking, pitting in the lower span, and tube fretting in preheated section and anti-vibration bar (AVB) locations. Primary Water Stress Corrosion Cracking (PWSCC) in the tube to tubesheet roll transition has been a wide spread problem in the Recirculation Steam Generators (RSG) during this period. Although significant progress has been made in resolving this problem, considerable work still remains. One typical problem in the Once Through Steam Generator (OTSG) was the tube support plate broached hole fouling which affects the OTSG steam generating

  15. End plug welding of nuclear fuel elements-AFFF experience

    International Nuclear Information System (INIS)

    Bhatt, R.B.; Singh, S.; Aniruddha Kumar; Amit; Arun Kumar; Panakkal, J.P.; Kamath, H.S.

    2004-01-01

    Advanced Fuel Fabrication Facility is engaged in the fabrication of mixed oxide (U,Pu)O 2 fuel elements of various types of nuclear reactors. Fabrication of fuel elements involves pellet fabrication, stack making, stack loading and end plug welding. The requirement of helium bonding gas inside the fuel elements necessitates the top end plug welding to be carried out with helium as the shielding gas. The severity of the service conditions inside a nuclear reactor imposes strict quality control criteria, which demands for almost defect free welds. The top end plug welding being the last process step in fuel element fabrication, any rejection at this stage would lead to loss of effort prior to this step. Moreover, the job becomes all the more difficult with mixed oxide (MOX) as the entire fabrication work has to be carried out in glove box trains. In the case of weld rejection, accepted pellets are salvaged by cutting the clad tube. This is a difficult task and recovery of pellets is low (requiring scrap recovery operation) and also leads to active metallic waste generation. This paper discusses the experience gained at AFFF, in the past 12 years in the area of end plug welding for different types of MOX fuel elements

  16. 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

  17. Circumferential cracking of steam generator tubes

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    1997-04-01

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

  18. Principle plug design for deposition tunnels

    International Nuclear Information System (INIS)

    Haaramo, M.; Lehtonen, A.

    2009-06-01

    protection of reinforcement. The tightness and the thickness of the concrete cover of the steel bars are tools that can be used to improve the situation. To avoid shrinkage and high generation of hydration heat the minimum volume of the aggregate should be 70% and in addition to cement, some other cementing agent should be used. The other cementing agent could be blast furnace cinder, fly ash or silica. The amount of cement needed is smallest with blast furnace cinder. It can be difficult to find cinder of a type with a suitable chemical consistency for repository use. The construction of the plug must start as soon as the backfilling of the deposition tunnel is completed. The estimated construction time of a plug is 40 days, including the hardening time of the concrete. The front wall of the backfill can be realized as a vertical plane. During the construction of the plug and the hardening of concrete the pressure of the backfill and the water is controlled by leakage tubes. In short term the plug can be subjected to a pressure smaller than the total pressure to be developed. The period between the completion of the backfill and the completion of the plug can be shortened through careful preliminary works. (orig.)

  19. Report on US-Japan 1983 meetings on steam generators

    International Nuclear Information System (INIS)

    1984-04-01

    This is a report on a trip to Japan by personnel of the US Nuclear Regulatory Commission in 1983 to exchange information on steam generators of nuclear power plants. Steam generators of Japanese pressurized water reactors have experienced nearly all of the forms of degradation that have been experienced in US recirculating-type steam generators, except for denting and pitting. More tubes have been plugged per year of reactor operation in Japanese than in US steam generators, but much of the Japanese tube plugging is preventative rather than the result of leaks experienced. The number of leaks per reactor year is much smaller for Japanese than for US steam generators. No steam generators have been replaced in Japan while several have been replaced in the US. The Japanese experience may be related to their very stringent inspection and maintenance programs for steam generators

  20. Explosive plugging of nuclear heat exchangers

    International Nuclear Information System (INIS)

    Crossland, B.; Bahrani, A.S.; Townsley, W.J.

    1977-01-01

    Explosive welding is a well established process for cladding one metal on another or for welding tubes to tubeplates or lap welding, etc. Recently, the process has been adapted to plugging of heat exchangers in conventional and nuclear power plant, where it has already been accepted especially in situations where the access is difficult and remote from the site of plugging. The paper describes the explosive plugging techniques developed in the Department of Mechanical and Industrial Engineering of The Queen's University of Belfast for the reheater and superheater of the PFR, and for the reheater of the AGR. For the PFR a point charge system has been used which causes a spherical expansion of the plug, which gives two zones of welding. Initially for the much larger plug required for the AGR it was proposed to use a parallel stand-off welding set-up, but it proved difficult or impossible to avoid a crevice. Consequently, a rim charge set-up has been developed which gives a circular ring expansion of the plug with two zones of welding. Besides the problem of the design of the plug and explosive charge geometry it has also been necessary to consider the distortion of holes adjoining the hole in which a plug is welded. Bunging of adjoining holes in order to reduce the distortion has also been investigated

  1. Minimize corrosion degradation of steam generator tube materials

    International Nuclear Information System (INIS)

    Lu, Y.

    2006-01-01

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

  2. Reactor vessel sealing plug

    International Nuclear Information System (INIS)

    Dooley, R.A.

    1986-01-01

    This invention relates to an apparatus and method for sealing the cold leg nozzles of a nuclear reactor pressure vessel from a remote location during maintenance and inspection of associated steam generators and pumps while the pressure vessel and refueling canal are filled with water. The apparatus includes a sealing plug for mechanically sealing the cold leg nozzle from the inside of a reactor pressure vessel. The sealing plugs include a primary and a secondary O-ring. An installation tool is suspended within the reactor vessel and carries the sealing plug. The tool telescopes to insert the sealing plug within the cold leg nozzle, and to subsequently remove the plug. Hydraulic means are used to activate the sealing plug, and support means serve to suspend the installation tool within the reactor vessel during installation and removal of the sealing plug

  3. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

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

  4. Steam generator tube fitness-for-service guidelines

    International Nuclear Information System (INIS)

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

    1995-07-01

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

  5. Design, development and deployment of special sealing plug for 540 MWe PHWRs

    International Nuclear Information System (INIS)

    Sharma, G.; Roy, S.; Patel, R.J.

    2012-01-01

    The coolant channel in Pressurized Heavy Water Reactors is a pressure boundary component and is very important for reactor performance and reactor safety. Monitoring the condition of the pressure tube of each coolant channel on a periodic basis is very important. In-Service Inspection (ISI) of the coolant channels in water filled condition is done regularly for 220 MWe PHWR. For the same purpose BARC Channel Inspection System is developed for 540 MWe PHWR also. Special Sealing Plug has been developed to facilitate the channel inspection (in water filled condition) with all necessary safety features at par with normal sealing plug. Special Sealing Plug provides a 50 mm through hole for passage of drive tube of Inspection Head maintaining integrity of PHT. Lot of challenges were faced for developing the Special Sealing Plug and its associated tools. It was a first of its kind design. First ISI of TAPS-4 was conducted successfully using this plug along with associated tools in November 2011. This development has provided immense help to NPCIL in life management of 540 MWe PHWR coolant channels. (author)

  6. 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

  7. Ford C-Max plug-in hybrid; Ford C-Max mit Plug-in-Hybridtechnik

    Energy Technology Data Exchange (ETDEWEB)

    Schamel, Andreas; D' Annunzio, Julie; Iorio, Rob [Ford Motor Company, Dearborn, MI (United States); Schmitz, Peter [Ford-Forschungszentrum Aachen GmbH, Aachen (Germany)

    2013-03-01

    Ford provides consumers a broad choice of electrified vehicles globally, including full hybrids, plug-in hybrids and all-electric vehicles. The all-new 2013 model year C-Max Energi Plug-in Hybrid utilises the third generation of Ford hybrid technology. This article discusses the hybrid powersplit architecture and components, as well as the charging capability and human-machine interfaces, used in the C-Max Energi Plug-In Hybrid. (orig.)

  8. 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

  9. Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

    International Nuclear Information System (INIS)

    Larson, G.C.

    1980-01-01

    This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

  10. Effect of gravity on liquid plug transport through an airway bifurcation model.

    Science.gov (United States)

    Zheng, Y; Anderson, J C; Suresh, V; Grotberg, J B

    2005-10-01

    Many medical therapies require liquid plugs to be instilled into and delivered throughout the pulmonary airways. Improving these treatments requires a better understanding of how liquid distributes throughout these airways. In this study, gravitational and surface mechanisms determining the distribution of instilled liquids are examined experimentally using a bench-top model of a symmetrically bifurcating airway. A liquid plug was instilled into the parent tube and driven through the bifurcation by a syringe pump. The effect of gravity was adjusted by changing the roll angle (phi) and pitch angle (gamma) of the bifurcation (phi = gamma =0 deg was isogravitational). Phi determines the relative gravitational orientation of the two daughter tubes: when phi not equal to 0 deg, one daughter tube was lower (gravitationally favored) compared to the other. Gamma determines the component of gravity acting along the axial direction of the parent tube: when gamma not equal to 0 deg, a nonzero component of gravity acts along the axial direction of the parent tube. A splitting ratio Rs, is defined as the ratio of the liquid volume in the upper daughter to the lower just after plug splitting. We measured the splitting ratio, Rs, as a function of: the parent-tube capillary number (Cap); the Bond number (Bo); phi; gamma; and the presence of pre-existing plugs initially blocking either daughter tube. A critical capillary number (Cac) was found to exist below which no liquid entered the upper daughter (Rs = 0), and above which Rs increased and leveled off with Cap. Cac increased while Rs decreased with increasing phi, gamma, and Bo for blocked and unblocked cases at a given Cap > Ca,. Compared to the nonblockage cases, Rs decreased (increased) at a given Cap while Cac increased (decreased) with an upper (lower) liquid blockage. More liquid entered the unblocked daughter with a blockage in one daughter tube, and this effect was larger with larger gravity effect. A simple theoretical

  11. Safety significance of steam generator tube degradation mechanisms

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-07-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  17. 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

  18. Modeling and Simulation of U-tube Steam Generator

    Science.gov (United States)

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

    2018-03-01

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

  19. Technique employed to seal a tube leaking in a heat exchanger of the tube type by explosives with supporting means for the adjacent tubes

    International Nuclear Information System (INIS)

    Larson, G.C.

    1978-01-01

    This invention concerns the technique employed to seal a tube leaking in a heat exchanger of the tube and tube plate type by detonating metal plugs activated by an explosive and inserted in both ends of the tube. It refers in particular to an apparatus and process in which the deformation or distortion of the adjacent tubes and tube plate ties under the effect of the explosive forces is significantly reduced [fr

  20. 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

  1. Steam generator tube integrity program. Phase I report

    International Nuclear Information System (INIS)

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

    1979-09-01

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

  2. Experimental and numerical modeling of sulfur plugging in carbonate reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Abou-Kassem, J.H. [Chemical and Petroleum Engineering Department, UAE University, PO Box 17555, Al-Ain (United Arab Emirates)

    2000-05-01

    Sour gas, mainly in the form of hydrogen sulfide, is produced in large amounts from many oil and gas reservoirs in the United Arab Emirates. In addition to creating problems in production lines, the precipitation of elemental sulfur in vicinity of the wellbore is often reported to cause wellbore damage. While there have been several studies performed on the role of solid deposition in gas reservoirs, the role of sulfur deposition in oil reservoirs has not been investigated. This paper presents experimental results along with a comprehensive wellbore model that predicts sulfur precipitation as well as plugging. Two separate sets of experiments, one for a gas phase system and another for a crude oil system, were conducted to investigate the deposition of elemental sulfur in (linear) carbonate cores. The gas flow tests were conducted with elemental sulfur being carried with nitrogen through limestone cores. Changes in gas flow rate were monitored while the injection pressure was held constant. A series of experiments generated valuable data for plugging with elemental sulfur. X-ray diffraction tests provided evidence of sulfur deposition along the cores. The oil flow tests were carried out to observe sulfur precipitation and plugging in a carbonate core. The crude oil was de-asphalted before conducting these tests in order to isolate the effect of asphaltene plugging. Significant plugging was observed and was found to be dependent on flow rate and initial sulfur concentration. This information was used in a phenomenological model that was incorporated in the wellbore numerical model. The data for the numerical model were obtained from both test tube and oil flow experiments. By using a phenomenological model, the wellbore plugging was modeled with an excellent match (with experimental results)

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

    International Nuclear Information System (INIS)

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

    1997-03-01

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

  4. Ontario Hydro's operating experience with steam generators with specifics on Bruce A and Bruce B problems

    Energy Technology Data Exchange (ETDEWEB)

    Eatock, J W; Patterson, R W [Ontario Hydro, Toronto, ON (Canada); Dyck, R W [Ontario Hydro, Central Production Services Division, Toronto, ON (Canada)

    1991-04-01

    The performance of the steam generators in Ontario Hydro nuclear power stations is reviewed. This performance has generally been outstanding compared to world averages, with very low tube failure and plugging rates. Steam generator problems have made only minor contributions to Ontario Hydro nuclear station incapability factors. The mechanisms responsible for the the observed tube degradation and failures are described. The majority of the leaks have been due fatigue in the U-bend of the Bruce 'A' steam generators. There have been very few failures attributed to corrosion of the three tube materials used in Ontario Hydro steam generators. Recent performance has been deteriorating primarily due to deposit accumulation in the steam generators. Plugging of the broached holes in the upper support plates at Bruce 'A' has caused some derating of two units. Increases have been observed in the primary heat transport system reactor inlet temperature of several units. These increases may be attributed to steam generator tube surface fouling. In addition, several units have accumulated deep, hard sludge piles on the tube sheet, although little damage been observed. Recently some fretting of tubes has been observed at BNGSB in the U-bend support region. Remedial measures are being taken to address the current problems. Solutions are being evaluated to reduce the generation of corrosion products in the feedtrain and their subsequent transport to the steam generators. (author)

  5. Ontario Hydro's operating experience with steam generators with specifics on Bruce A and Bruce B problems

    International Nuclear Information System (INIS)

    Eatock, J.W.; Patterson, R.W.; Dyck, R.W.

    1991-01-01

    The performance of the steam generators in Ontario Hydro nuclear power stations is reviewed. This performance has generally been outstanding compared to world averages, with very low tube failure and plugging rates. Steam generator problems have made only minor contributions to Ontario Hydro nuclear station incapability factors. The mechanisms responsible for the the observed tube degradation and failures are described. The majority of the leaks have been due fatigue in the U-bend of the Bruce 'A' steam generators. There have been very few failures attributed to corrosion of the three tube materials used in Ontario Hydro steam generators. Recent performance has been deteriorating primarily due to deposit accumulation in the steam generators. Plugging of the broached holes in the upper support plates at Bruce 'A' has caused some derating of two units. Increases have been observed in the primary heat transport system reactor inlet temperature of several units. These increases may be attributed to steam generator tube surface fouling. In addition, several units have accumulated deep, hard sludge piles on the tube sheet, although little damage been observed. Recently some fretting of tubes has been observed at BNGSB in the U-bend support region. Remedial measures are being taken to address the current problems. Solutions are being evaluated to reduce the generation of corrosion products in the feedtrain and their subsequent transport to the steam generators. (author)

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  7. Steam generator tubing development for commercial fast breeder reactors

    International Nuclear Information System (INIS)

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

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Cepcek, S.

    1997-01-01

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

  9. Failure of fretted steam generator tubes under accident conditions

    International Nuclear Information System (INIS)

    Forrest, C.F.

    1996-10-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  11. Steam generator tube rupture (SGTR) scenarios

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  12. Tube sheet design for PFBR steam generator

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    International Nuclear Information System (INIS)

    DeFur, D.D.

    1975-04-01

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

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

    International Nuclear Information System (INIS)

    Denis, Jean.

    1977-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Xiaoqing Tian

    2015-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-08-01

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

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

    International Nuclear Information System (INIS)

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

    1988-08-01

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

  19. The Sealed Tube Neutron Generator

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  20. In service inspection for steam generator tubes

    International Nuclear Information System (INIS)

    Comby, R.; Eyrolles, Ph.

    1988-01-01

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

  1. Welding nuclear reactor fuel rod end plugs

    International Nuclear Information System (INIS)

    Yeo, D.

    1984-01-01

    Apparatus for applying a vacuum to a nuclear fuel rod cladding tube's interior through its open end while girth welding an inserted end plug to its other end. An airtight housing has an orifice with a seal which can hermetically engage the tube's open end. A vacuum hose has one end connected to the housing and the other end connected to a vacuum pump. A mechanized device which moves the housing to engage or disengage its seal with the tube's open end includes at least one arm having one end attached to the housing and the other end pivotally attached to a movable table; an arm rotating device to coaxially align the housing's orifice with the welding-positioned tube; and a table moving device to engage the seal of the coaxially aligned orifice with the tube's open end. (author)

  2. Evaluation of sampling schemes for in-service inspection of steam generator tubing

    International Nuclear Information System (INIS)

    Hanlen, R.C.

    1990-03-01

    This report is a follow-on of work initially sponsored by the US Nuclear Regulatory Commission (Bowen et al. 1989). The work presented here is funded by EPRI and is jointly sponsored by the Electric Power Research Institute (EPRI) and the US Nuclear Regulatory Commission (NRC). The goal of this research was to evaluate fourteen sampling schemes or plans. The main criterion used for evaluating plan performance was the effectiveness for sampling, detecting and plugging defective tubes. The performance criterion was evaluated across several choices of distributions of degraded/defective tubes, probability of detection (POD) curves and eddy-current sizing models. Conclusions from this study are dependent upon the tube defect distributions, sample size, and expansion rules considered. As degraded/defective tubes form ''clusters'' (i.e., maps 6A, 8A and 13A), the smaller sample sizes provide a capability of detecting and sizing defective tubes that approaches 100% inspection. When there is little or no clustering (i.e., maps 1A, 20 and 21), sample efficiency is approximately equal to the initial sample size taken. Thee is an indication (though not statistically significant) that the systematic sampling plans are better than the random sampling plans for equivalent initial sample size. There was no indication of an effect due to modifying the threshold value for the second stage expansion. The lack of an indication is likely due to the specific tube flaw sizes considered for the six tube maps. 1 ref., 11 figs., 19 tabs

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

    International Nuclear Information System (INIS)

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

    1985-01-01

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

  4. Method of repairing pressure tube type reactors

    International Nuclear Information System (INIS)

    Asada, Takashi.

    1983-01-01

    Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

  5. Method of start-up of rotary plug sealing devices in FBR type reactors

    International Nuclear Information System (INIS)

    Sakuragi, Masanori; Akita, Haruo

    1980-01-01

    Purpose: To rapidly and safely start-up the rotary plug sealing device by controlling to eliminate the pressure difference in the pressures of gases exerting on the liquid surfaces in the inner and the outer cylinders of a sealing alloy vessel in the rotary plug of a FBR type reactor. Method: In a case where an abnormal state results in the pressure difference of gases exerted on the liquid surfaces in the inner and the outer cylinders of a vessel charged with sealing alloy in a rotary plug and the sealing valve for the back-up gas supply tube is rapidly closed to seal the sealing portion, the pressure in the gas supply tube is controlled so that the pressure difference in the gases exerted on the liquid surfaces in the inner and outer cylinders while closing the sealing valve. Then, after conforming that the pressure is controlled to a predetermined level at which the pressure difference can be regarded to be zero, the sealing valve is gradually opened while regulating the pressure in the gas supply tube so as to maintain the pressure difference to a predetermined level. This prevents the occurrence of external disturbances upon opening of the sealing valve and enables rapid and safety start-up for the rotary plug sealing device. (Moriyama, K.)

  6. 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

  7. 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.

  8. 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.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  10. Impulse generation by detonation tubes

    Science.gov (United States)

    Cooper, Marcia Ann

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

  11. Neutron generator tube ion source control

    International Nuclear Information System (INIS)

    Bridges, J.R.

    1982-01-01

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

  12. 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)

  13. Corrosion induced clogging and plugging in water-cooled generator cooling circuit

    International Nuclear Information System (INIS)

    Park, B.G.; Hwang, I.S.; Rhee, I.H.; Kim, K.T.; Chung, H.S.

    2002-01-01

    Water-cooled electrical generators have been experienced corrosion-related problems that are restriction of flow through water strainers caused by collection of excessive amounts of copper corrosion products (''clogging''), and restriction of flow through the copper strands in the stator bars caused by growth or deposition of corrosion products on the walls of the hollow strands (''plugging''). These phenomena result in unscheduled shutdowns that would be a major concern because of the associated loss in generating capacity. Water-cooled generators are operated in one of two modes. They are cooled either with aerated water (dissolved oxygen >2 ppm) or with deaerated water (dissolved oxygen <50 ppb). Both modes maintain corrosion rates at satisfactorily low levels as long as the correct oxygen concentrations are maintained. However, it is generally believed that very much higher copper corrosion rates result at the intermediate oxygen concentrations of 100-1000 ppb. Clogging and plugging are thought to be associated with these intermediate concentrations, and many operators have suggested that the period of change from high-to-low or from low-to-high oxygen concentration is particularly damaging. In order to understand the detailed mechanism(s) of the copper oxide formation, release and deposition and to identify susceptible conditions in the domain of operating variables, a large-scale experiments are conducted using six hollow strands of full length connected with physico-chemically scaled generator cooling water circuit. To ensure a close simulation of thermal-hydraulic conditions in a generator stator, strands of the loop will be ohmically heated using AC power supply. Experiments is conducted to cover oxygen excursions in both high dissolved oxygen and low dissolved oxygen conditions that correspond to two representative operating condition at fields. A thermal upset condition is also simulated to examine the impact of thermal stress. During experiments

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  15. A Flue Gas Tube for Thermoelectric Generator

    DEFF Research Database (Denmark)

    2013-01-01

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

  16. 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)

  17. Fracture toughness determination in steam generator tubes

    International Nuclear Information System (INIS)

    Bergant M; Yawny, A; Perez Ipina, J

    2012-01-01

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

  18. North Anna Power Station - Unit 1: Overview of steam generator replacement project activities

    International Nuclear Information System (INIS)

    Gettler, M.W.; Bayer, R.K.; Lippard, D.W.

    1993-01-01

    The original steam generators at Virginia Electric and Power Company's (Virginia Power) North Anna Power Station (NAPS) Unit 1 have experienced corrosion-related degradation that require periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, continued tube degradation in the steam generators necessitated the removal from service of approximately 20.3 percent of the tubes by plugging, (18.6, 17.3, and 25.1 for steam generators A, B, and C, respectively). Additionally, the unit power was limited to 95 % during, its last cycle of operation. Projections of industry and Virginia Power experience indicated the possibility of mid-cycle inspections and reductions in unit power. Therefore, economic considerations led to the decision to repair the steam generators (i.e., replace the steam generator lower assemblies). Three new Model 51F Steam Generator lower assembly units were ordered from Westinghouse. Virginia Power contracted Bechtel Power Corporation to provide the engineering and construction support to repair the Unit 1 steam generators. On January 4, 1993, after an extended coastdown period, North Anna Unit 1 was brought off-line and the 110 day (breaker-to-breaker) Steam Generator Replacement Project (SGRP) outage began. As of this paper, the outage is still in progress

  19. Plug-in electric vehicles integrating fluctuating renewable electricity

    Energy Technology Data Exchange (ETDEWEB)

    Dallinger, David

    2013-11-01

    This paper examines a method to model plug-in electric vehicles as part of the power system and presents results for the contribution of plug-in electric vehicles to balance the fluctuating electricity generation of renewable energy sources. The scientific contribution includes: - A novel approach to characterizing fluctuating generation. This allows the detailed comparison of results from energy analysis and is the basis to describe the effect of electricity from renewable energy sources and plug-in electric vehicles on the power system. - The characterization of mobile storage, which includes the description of mobility behavior using probabilities and battery discharging costs. - The introduction of an agent-based simulation approach, coupling energy markets and distributed grids using a price-based mechanism design. - The description of an agent with specific driving behavior, battery discharging costs and optimization algorithm suitable for real plug-in vehicles and simulation models. - A case study for a 2030 scenario describing the contribution of plug-in electric vehicles to balance generation from renewable energy sources in California and Germany.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    Zong Hai; Wang Detai; Ding Yang

    2012-01-01

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

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

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kweon, Young Chul

    2002-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-06-15

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

  5. Development of end plug welding method in the fabrication of FBR fuel pins

    International Nuclear Information System (INIS)

    Ohtani, Seiji; Sawayama, Takeo; Tateishi, Yoshinori

    1977-01-01

    As a part of the development of the automatic and remote controlled fabrication of FBR fuel pins, welding of fuel pin end plugs has been examined. Cladding tubes and end plugs used for this experiment are made of SUS 316, and they are the components of fuel pins for the prototype fast breeder reactor (Monju) or the second core of Joyo (Joyo MK-II). The welding tests of cladding tubes and four kinds of end plugs were carried out by means of two techniques; tungsten inert gas welding and laser welding. It can be said that no considerable difference was observed in weld penetration, occurrence rate of weld defects and breaking strength between the tight fit and the loose fit plugs. The face-to-face fit welding requires the least welding heat input, but involves much difficulty in the control of weld penetration and bead zone diameter. The good concentrative property and high energy density of laser beam make the face of weld hollow due to the vaporization of weld metal. However, this problem can be easily solved by changing the shape of end plugs. Good results in the other characteristics of the weld also were obtained by this laser welding. Further experiment is needed in connection with the compatibility of weld metal with sodium and neutron irradiation before final judgement is made on the laser welding technique. (Nakai, Y.)

  6. Considerations in selecting tubing materials for CANDU steam generators

    International Nuclear Information System (INIS)

    Hemmings, R.L.

    1978-01-01

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

  7. Plug Power

    Energy Technology Data Exchange (ETDEWEB)

    Marsh, A. [Plug Power Inc., New York, NY (United States)

    2009-07-01

    This presentation described Plug Power's GenDrive hydrogen fuel cell unit that supplies the power needs for folk lift trucks used in high-throughput distribution and high-volume manufacturing operations. The system offers an alternative to lead acid batteries, providing maximum performance at all times during use. The system is particularly useful in the material handling industry, where the revenue generated is based on operator uptime and lift truck productivity. The use of the system allows customers to reduce operational costs and expand valuable floor space by eliminating batteries and associated recharging infrastructure. Fuel cell units also reduce the wear on truck motors. Truck operators can easily and safely refuel at hydrogen fueling stations in 1-5 minutes. GenDrive works with all major OEM lift trucks, making the transition seamless. Commercial customers are investing in this solution to improve their current operations. In 2008, Plug Power sold to Wal-Mart, Bridgestone Firestone and Nestle. Most notably, Central Grocers purchased 220 fuel cell units for a new greenfield distribution center. Plug Power currently has more than 380 systems in operation.

  8. Decontamination of Steam Generator tube using Abrasive Blasting Technology

    International Nuclear Information System (INIS)

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

    2010-01-01

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

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

    International Nuclear Information System (INIS)

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

    1988-06-01

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

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

    Directory of Open Access Journals (Sweden)

    Tian Xiaoqing

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-12-01

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

  13. Design of Monju steam generator - with regard to maintenance and repair

    International Nuclear Information System (INIS)

    Takahashi, T.; Yamagishi, Y.; Mukaibo, R.

    2002-01-01

    The steam generator design of 'Monju' started in 1968 and since then, extensive research and design work has been done. 'Monju' steam generator consists of one evaporator and one superheater to each of the three independent cooling systems. Ultrasonic and eddy current tubing inspection devices have been developed for maintenance. And for failed tubes, welding or explosive plugging is applied. Following the completed safety review and the coming design and construction licensing, 'Monju' is expected to reach criticality in fiscal year 1990. (author)

  14. Aspects of welding of zircaloy thin tube to end plugin the experimental welding facility of fuel element fabrication laboratory

    International Nuclear Information System (INIS)

    Shafy, M.; El-Hakim, E.

    1997-01-01

    The work was achieved within the scope of developing egyptian nuclear fuel fabrication laboratory in inshas. It showed the results of developing a welding facility for performing a qualified zircaloy-2 and 4 thin tubes to end weld joints. The welding chamber design was developed to get qualified weld for both PWR and CANDU fuel rod configurations. Experimental works for optimizing the welding parameters of tungsten inert gas (TIG) welding and electron beam (EB) welding processes were achieved. The ld penetration deeper than the wall tube thickness can be obtained for qualified end plug weld joints. It recommended to use steel compensating block for radiographic inspection of end plug weld joints. The predominate defects that can be expected in end plug weld joints, are lack of penetration and cavity. The microstructure of the fusion zone and heat affected zones are Widmanstaetten structure and its grain size is drastically sensible to the heat generation and removal of arc welding. 16 figs

  15. Operating experience gained during the copper oxide plugging incident in Koeberg unit 1 generator stator

    International Nuclear Information System (INIS)

    Mellor, S.P.; Matthee, F.W.

    2002-01-01

    In June 1999 Koeberg's unit 1 started to experience adverse operating conditions which were later ascribed to blockages in the hollow conductors of the generator stator. These blockages were attributed to copper oxide plugs which developed progressively during the following year and culminated in reduced power operation. Many attempts were made to address the plugging by implementing various off-line and on-line cleaning processes. Subsequent to a successful on-line cleaning operation, the unit was returned to full power and the chemistry regime for the stator cooling water system was changed to allow for operation at an elevated pH. This paper discusses Koeberg's experience with copper oxide blockages, describes the initial indications of the problem and the impact on the operating parameters. The remainder of the paper focuses on the actions taken to address the deteriorating situation and the different cleaning methods implemented to remove the copper oxide deposits. The paper concludes with the current status of the unit 1 generator stator and the lessons learned during the resolution of this problem. (authors)

  16. Operating experience gained during the copper oxide plugging incident in Koeberg unit 1 generator stator

    Energy Technology Data Exchange (ETDEWEB)

    Mellor, S.P.; Matthee, F.W. [ESKOM, Koeberg Nuclear Power Station (South Africa)

    2002-07-01

    In June 1999 Koeberg's unit 1 started to experience adverse operating conditions which were later ascribed to blockages in the hollow conductors of the generator stator. These blockages were attributed to copper oxide plugs which developed progressively during the following year and culminated in reduced power operation. Many attempts were made to address the plugging by implementing various off-line and on-line cleaning processes. Subsequent to a successful on-line cleaning operation, the unit was returned to full power and the chemistry regime for the stator cooling water system was changed to allow for operation at an elevated pH. This paper discusses Koeberg's experience with copper oxide blockages, describes the initial indications of the problem and the impact on the operating parameters. The remainder of the paper focuses on the actions taken to address the deteriorating situation and the different cleaning methods implemented to remove the copper oxide deposits. The paper concludes with the current status of the unit 1 generator stator and the lessons learned during the resolution of this problem. (authors)

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-02-15

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

  19. Environmental Benefits of Using Wind Generation to Power Plug-In Hybrid Electric Vehicles

    Directory of Open Access Journals (Sweden)

    Mahdi Hajian

    2011-08-01

    Full Text Available As alternatives to conventional vehicles, Plug-in Hybrid Electric Vehicles (PHEVs running off electricity stored in batteries could decrease oil consumption and reduce carbon emissions. By using electricity derived from clean energy sources, even greater environmental benefits are obtainable. This study examines the potential benefits arising from the widespread adoption of PHEVs in light of Alberta’s growing interest in wind power. It also investigates PHEVs’ capacity to mitigate natural fluctuations in wind power generation.

  20. Fretting wear damage of steam generator tubes and its prediction modeling

    International Nuclear Information System (INIS)

    Che Honglong; Lei Mingkai

    2013-01-01

    The steam generator is the key equipment used for the energy transition in nuclear power plant. Since the high-temperature and high-pressure fluid flows with high speed, the steam generator tubes will be excited and vibrate, leading to the tremendous fretting wear problem on the tubes, sometimes even leading to tube cracking. This paper introduces typical fretting wear cases, the result of corresponding simulation wear experiment and damage mechanism which combining mechanical wear and erosion-corrosion. Work rate model could give a reasonable life prediction about the steam generator tube, and this predictive model has been used in nuclear power plant safety assessment. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Parrish, K.R.

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    Parrish, K.R.

    1995-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-12-15

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

  6. Anatomy Education for the YouTube Generation

    Science.gov (United States)

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

    2016-01-01

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

  7. Estimating probable flaw distributions in PWR steam generator tubes

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  8. Troubles detected during regular inspection of No.1 plant in Oi Power Station, Kansai Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1990-01-01

    No. 1 plant in Oi Power Station, Kansai Electric Power Co., Inc. is a PWR plant with rated output of 1175 MW, and its regular inspection is carried out since August 14, 1989. When eddy current flaw detection inspection was carried out on the total number (11426 except already plugged tubes) of the heating tubes of steam generators, significant indication was observed in tube supporting plate part of 279 tubes, at the boundary of tube plate expanded part of 34 tubes, and in the tube plate expanded part of 99 tubes, 411 heating tubes in total (all on high temperature side). Consequently, it was decided to repair 367 tubes using sleeves, and to plug other 44 tubes. Besides, among the heating tubes plugged in the past, it was decided to remove plugs from 161 tubes, and by repairing them with sleeves, to use them again. Total number of heating tubes 13552 (3388 tubes x 4 steam generators), Number of plugged tubes 2009 (decrease by 117 this time), Ratio of plugging 14.8%. (K.I.)

  9. An Overview of the HomePlug AV2 Technology

    Directory of Open Access Journals (Sweden)

    Larry Yonge

    2013-01-01

    Full Text Available HomePlug AV2 is the solution identified by the HomePlug Alliance to achieve the improved data rate performance required by the new generation of multimedia applications without the need to install extra wires. Developed by industry-leading participants in the HomePlug AV Technical Working Group, the HomePlug AV2 technology provides Gigabit-class connection speeds over the existing AC wires within home. It is designed to meet the market demands for the full set of future in-home networking connectivity. Moreover, HomePlug AV2 guarantees backward interoperability with other HomePlug systems. In this paper, the HomePlug AV2 system architecture is introduced and the technical details of the key features at both the PHY and MAC layers are described. The HomePlug AV2 performance is assessed, through simulations reproducing real home scenarios.

  10. Heat exchanger tubing materials for CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Taylor, G.F.

    1977-07-01

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

  11. Improvements in welding parameters for a new design of zircaloy-4 tube-end plug joints

    International Nuclear Information System (INIS)

    Martinez, R.L.; Fernandez, L.; Corso, H.L.; Ausas, J; Santisteban, J.R.

    2010-01-01

    This work presents the experimental results for the characterization of welds using a new design for zircaloy-4 tube-end plug joints, applicable to the production of fuel elements for the Atucha I Nuclear Plant. Test specimens were prepared following the new joint design and were welded using orbital welding equipment. Hydrogen content was measured in the different welding areas, and corrosion tests, and mechanical and microstructural descriptions were carried out, obtaining values that meet the current production standards. We reported previously that test samples welded in equipment with a smaller camera showed some relatively high hydrogen levels, together with alterations in the welded zone in the corrosion tests. Given these results, new tests were undertaken to optimize the welding parameters, being very careful with the purity of the welding atmosphere and in the handling of the samples. The intensity of the welding current was increased slightly to obtain better penetration of the material, without significantly increasing the heat input. The traction resistance values improved, reducing the hydrogen content to well below the maximum allowed by the standards (25 ppm) in all the welding zones and obtaining satisfactory results in the corrosion tests

  12. Shielding plug device

    International Nuclear Information System (INIS)

    Orii, Shoichi; Hasegawa, Satoshi; Makishima, Kenji.

    1976-01-01

    Object: To reduce the size of and extend the life of a revolving bearing and facilitate the laying of driving cables and duct lines, this being accomplished by providing plug raising means of a fast breeder on a stationary plug mounting base so as to prevent the shearing force of sodium from acting upon the revolving bearing. Structure: The shield plug means comprises a stationary plug secured to the open end of the reactor container, a rotary plug rotatable with respect to the stationary plug, an annular base formed on top of the stationary plug so as to cover the rotary plug, a bearing secured to the rotary plug edge lower face and upper and lower locking plates. At the time of the rotation of the rotary plug, the upper locking plate is withdrawn, the stationary plug is raised to release the seal structure, and the lower locking plate is inserted between the bearing and stationary plug. In this way, smooth rotation of the rotary plug can be obtained. (Horiuchi, T.)

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

    International Nuclear Information System (INIS)

    Karwoski, K.J.

    2009-01-01

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

  14. 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)

  15. Status of Siemens steam generator design and measures to assure continuous long-term reliable operation

    International Nuclear Information System (INIS)

    Hoch, G.

    1999-01-01

    Operating pressurized water reactors with U-tube steam generators have encountered difficulties with either one or a combination of inadequate material selection, poor design or manufacturing and an insufficient water chemistry control which resulted in excessive tube degradation. In contrast to the above mentioned problems, steam generators from Siemens/KWU are proving by operating experience that all measures undertaken at the design stage as well as during the operating and maintenance phase were effective enough to counteract any tube corrosion phenomena or other steam generator related problem. An Integrated Service Concept has been developed, applied and wherever necessary improved in order to ensure reliable steam generator operation. The performance of the steam generators is updated continuously, evaluated and implemented in lifetime databases. The main indicator for steam generator integrity are the results of the eddy current testing of the steam generator tubes. Tubes with indications are rated with lifetime threshold values and if necessary plugged, based on individual assessment criteria.(author)

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

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  17. Flaw analysis in steam generator tube

    International Nuclear Information System (INIS)

    Hutin, J.P.; Billon, F.

    1985-08-01

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

  18. An Overview of the HomePlug AV2 Technology

    OpenAIRE

    Yonge, Larry; Abad, Jose; Afkhamie, Kaywan; Guerrieri, Lorenzo; Katar, Srinivas; Lioe, Hidayat; Pagani, Pascal; Riva, Raffaele; Schneider, Daniel M.; Schwager, Andreas

    2013-01-01

    HomePlug AV2 is the solution identified by the HomePlug Alliance to achieve the improved data rate performance required by the new generation of multimedia applications without the need to install extra wires. Developed by industry-leading participants in the HomePlug AV Technical Working Group, the HomePlug AV2 technology provides Gigabit-class connection speeds over the existing AC wires within home. It is designed to meet the market demands for the full set of future in-home networking co...

  19. 2D fringe probing of liquid film dynamics of a plug bubble in a micropipe

    International Nuclear Information System (INIS)

    Ji, H F; Qiu, H H

    2009-01-01

    An extended film thickness measurement method that can obtain the liquid film thickness profile of the whole plug bubble in a capillary tube simultaneously is presented. The approach is based on a 2D spatial fringe scattering technique, where the spatial frequency of scattered fringes is a function of liquid film thickness along the micropipe. Laser sheets are used instead of the laser beams, and the width of the laser sheets can be selected to cover the whole measurement field. Capillary tubes, with inner diameters of 1.0 mm and 0.3 mm, and lengths of 125 mm and 65 mm, are used. A gas plug bubble, 2.5–20 mm long, is introduced and moves through the testing part of the tube, which is filled with water as the working fluid. The interference fringes produced by two incident laser sheets are scattered from the interface between gas and water, and captured by a high-speed camera at a speed of up to 2000 frames s −1 . The experimental results show that the improved method can obtain the liquid film thickness profile at the different times and can be used to analyze the status of the plug bubble movement in a micropipe

  20. Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor

    International Nuclear Information System (INIS)

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

    1981-01-01

    Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt

  1. Stress analysis of steam generator row-1 tubes

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    Dillon, R.L.

    1981-01-01

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

  3. 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)

  4. Maintenance implications of critical components in ITER CXRS upper port plug design

    International Nuclear Information System (INIS)

    Koning, Jarich; Jaspers, Roger; Doornink, Jan; Ouwehand, Bernard; Klinkhamer, Friso; Snijders, Bart; Sadakov, Sergey; Heemskerk, Cock

    2009-01-01

    Already in the early phase of a design for ITER, the maintenance aspects should be taken into account, since they might have serious implications. This paper presents the arguments in support of the case for the maintainability of the design, notably if this maintenance is to be performed by advanced remote methods. This structure is compliant to the evolving maintenance strategy of ITER. Initial results of a Failure Mode Effects and Criticality Analysis (FMECA) and a development risk analysis for the ITER upper port plug no. 3, housing the Charge Exchange Recombination Spectroscopy (CXRS) diagnostic, are employed for the definition of the maintenance strategy. The CXRS upper port plug is essentially an optical system which transfers visible light from the plasma into a fiber bundle. The most critical component in this path is the first mirror (M1) whose reflectivity degrades during operation due to deposition and/or erosion dominated effects. Amongst other measures to mitigate these effects, the strategy is to allow for a replacement of this mirror. Therefore it is mounted on a retractable central tube. The main purpose of this tube is to make frequent replacements possible without hindering operation. The maintenance method in terms of time, geometry and spare part policy has a large impact on cost of the system and time usage in the hot cell. Replacement of the tube under vacuum and magnetic field seems infeasible due to the operational risk involved. The preferred solution is to have a spare tube available which is replaced in parallel with other maintenance operations on the vessel, as to avoid any interference in the hot cell with the shutdown scheduling. This avoids having to refurbish a full port plug and also allows for a more frequent replacement of M1, as we can replace the mirror anytime the vacuum vessel is vented, estimated to be once a year.

  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. A Special Application Coiled Tubing Applied Plug for Geothermal Well Casing Remediation

    International Nuclear Information System (INIS)

    Knudsen, S.D.; Sattler, A.R.; Staller, G.E.

    1999-01-01

    Casing deformation in wells is a common problem in many geothermal fields. Casing remediation is necessary to keep wells in production and occasionally, to even enter the well for an approved plug and abandonment procedure. The costly alternative to casing remediation is to incur the expense of drilling a new well to maintain production or drilling a well to intersect a badly damaged well below the deformation for abandonment purposes. The U.S. Department of Energy and the Geothermal Drilling Organization sponsor research and development work at Sandia National Laboratories in an effort to reduce these remediation expenditures. Sandia, in cooperation with Halliburton Energy Services, has developed a low cost, commercially available, bridge-plug-type packer for use in geothermal well environments. This report documents the development and testing of this tool for use in casing remediation work

  7. Internal ultrasonic testing of steam generator tubes

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  8. Review on Automotive Power Generation System on Plug-in Hybrid Electric Vehicles & Electric Vehicles

    Directory of Open Access Journals (Sweden)

    Leong Yap Wee

    2016-01-01

    Full Text Available Regenerative braking is a function to recharge power bank on the Plug-in electric vehicles (PHEV and electric vehicles (EV. The weakness of this system is, it can only perform its function when the vehicle is slowing down or by stepping the brake foot pedal. In other words, the electricity recharging system is inconsistent, non-continuous and geography dependent. To overcome the weakness of the regenerative braking system, it is suggested that to apply another generator which is going to be parallel with the regenerative braking system so that continuous charging can be achieved. Since the ironless electricity generator has a less counter electromotive force (CEMF comparing to an ironcored electricity generator and no cogging torque. Applying the ironless electricity generator parallel to the regenerative braking system is seen one of the options which creates sustainable charging system compared to cored electricity generator.

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

    International Nuclear Information System (INIS)

    Ruloff, G.

    1990-01-01

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

  10. Single-tank X-ray generator

    International Nuclear Information System (INIS)

    Petersen, K.

    1979-01-01

    The X-ray tube and the high-tension generator are provided with insulators which together form a high-tension plug connection. By this means the generator casing gets a compact shape. The anode is cooled in a closed cycle by means of oil or water, the cooling pipes being spirally wound around the high-tension generator in the interspace towards the casing. The cooling pipes themselves are cooled in their turn by the air stream of a fan in the casing. Intake and exhaust air flows through slots into and out of the casing (Macrotank G 2000 for shipyards). (DG) [de

  11. Mechanical and thermo-mechanical analyses of the tapered plug for plugging of deposition tunnels. A feasibility study

    Energy Technology Data Exchange (ETDEWEB)

    Faelth, Billy (Clay Technology AB, Lund (Sweden)); Gatter, Patrik (Vattenfall Power Consultant AB, Stockholm (Sweden))

    2009-09-15

    This report presents results from a study that was carried out in order to examine the applicability of the tapered plug concept for plugging of deposition tunnels in the deep repository for spent nuclear fuel. The report presents results from mechanical and thermo-mechanical models of the tapered plug. The models were analyzed with 3DEC. The models included a portion of a deposition tunnel and its intersection with a main tunnel. In the deposition tunnel, a tapered concrete plug was installed. The plug was subjected to the combined load from the swelling backfill material and from pore pressure inside the deposition tunnel. The thermo-mechanical effects due to the heat generation in the spent fuel were also included in the analyses. Generic material parameter values for the concrete were used. The following items were studied: - Stresses and displacements in the plug. - Shear stresses and shear displacements in the rock-concrete interface. - Stress additions in the rock due to the loads. The sensitivity of the results to changes of constitutive models, to changes of the plug geometry and to pore water pressure in the rock-concrete interface was examined. The results indicate that the displacements in the plug will be within reasonable ranges but the stresses may locally be high enough that they exceed acceptable levels. However, they can be reduced by choice of advantageous plug geometry and by having a good rock-concrete bond. The results also show that the stress additions in the rock due to the thermal load may yield stresses that locally exceed the spalling strength of the rock. At most locations, however, the rock stresses will amount at lower levels. It was concluded that, with choice of an appropriate design, the tapered plug seems to be an applicable concept for plugging of deposition tunnels. It was also concluded that further studies of the tapered plug concept should use material properties parameter values for low-pH concrete. Further, they should also

  12. Fatigue cracking on a steam generator tube

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  13. Pulse tube coolers for Meteosat third generation

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  14. Anatomy education for the YouTube generation.

    Science.gov (United States)

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  16. The steam generator repair project of the Donald C. Cook Nuclear Plant, Unit 2

    International Nuclear Information System (INIS)

    White, J.D.

    1993-01-01

    Donald C. Cook Nuclear Plant Unit 2 is part of a two unit nuclear complex located in southwestern Michigan and owned and operated by the Indiana Michigan Power Company. The Cook Nuclear Plant is a pressurized water reactor (PWR) plant with four Westinghouse Series 51 steam generators housed in an ice condenser containment. This paper describes the program undertaken by Indiana Michigan Power and the American Electric Power Service Corporation (AEPSC) to repair the Unit 2 steam generators. (Both Indiana Michigan Power and AEPSC arc subsidiaries of American Electric Power Company, Incorporated (AEP). AEPSC provides management and technical support services to Indiana Michigan Power and the other AEP operating companies.) Eddy current examinations, in a series of refueling and forced outages between November 1983 and July 1986 resulted in 763 (5.6%) plugged tubes. In order to maintain adequate reactor core cooling, a limit of 10% is placed on the allowable percentage of steam generator tubes that can be removed from service by plugging. Additionally, sections of tubes were removed for metallurgical analysis and confirmed that the degradation was due to intergranular stress corrosion cracking. In developing the decision on how to repair the steam generators, four alternative actions were considered for addressing these problems: retubing in place, sleeving, operating at 80% reactor power to lower temperature and thus reduce the rate of corrosion, replacing steam generator lower assemblies

  17. Steam-Generator Integrity Program/Steam-Generator Group Project

    International Nuclear Information System (INIS)

    1982-10-01

    The Steam Generator Integrity Program (SGIP) is a comprehensive effort addressing issues of nondestructive test (NDT) reliability, inservice inspection (ISI) requirements, and tube plugging criteria for PWR steam generators. In addition, the program has interactive research tasks relating primary side decontamination, secondary side cleaning, and proposed repair techniques to nondestructive inspectability and primary system integrity. The program has acquired a service degraded PWR steam generator for research purposes. This past year a research facility, the Steam Generator Examination Facility (SGEF), specifically designed for nondestructive and destructive examination tasks of the SGIP was completed. The Surry generator previously transported to the Hanford Reservation was then inserted into the SGEF. Nondestructive characterization of the generator from both primary and secondary sides has been initiated. Decontamination of the channelhead cold leg side was conducted. Radioactive field maps were established in the steam generator, at the generator surface and in the SGEF

  18. 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

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

    International Nuclear Information System (INIS)

    Pigeon, M.; David, B.

    1983-06-01

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

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

    International Nuclear Information System (INIS)

    Gay, N.; Granger, S.

    1992-11-01

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

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

    International Nuclear Information System (INIS)

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

    2004-10-01

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

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  3. 1984 NRC Branch annual report: Steam Generator Group Project

    International Nuclear Information System (INIS)

    Clark, R.A.; Bickford, R.L.; Birks, A.S.

    1985-02-01

    Program objectives are to develop validated models, based on experimental data, for prediction of margins-to-failure under burst and collapse pressures of steam generator tubing found to be service-degraded by eddy current inservice inspection. This required the research contractor, Pacific Northwest Laboratory, to obtain, house and prepare an out-of-service degraded steam generator for a confirmatory research program. This research includes: (1) validation of current and developmental nondestructive examination techniques; (2) optimization of inservice inspection procedures, sampling plan and inspection period; (3) validation of tube integrity predictive models; (4) validation of stress corrosion cracking predictive models; (5) development/optimization of tube plugging criteria; and (6) evaluation of proposed chemical cleaning and decontamination processes/procedures with respect to near-term integrity and long-term effects on corrosion, degradation and safety

  4. Trouble found in regular inspection of No.1 plant in Ikata Power Station, Shikoku Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1989-01-01

    Since May 2, 1989, the regular inspection of No.1 plant which is a PWR plant with the rated output of 566 MW in Ikata Power Station, Shikoku Electric Power Co., Inc. has been carried out, and eddy current flaw detection inspection was conducted on the total 6585 heating tubes of steam generators except already plugged tubes. As the result, significant indication was observed in 12 heating tubes at the expanded part of the high temperature side tube plates. As to the cause, similarly to those observed in the same plant in the past, it is considered that the residual stress caused by expanding at the time of the manufacture and the internal pressure stress during the operation were superposed, and stress corrosion cracking occurred. It was decided that these 12 defective tubes are plugged. State of plugging in steam generators. Number of total heating tubes: 6776=3388 tubes x 2 steam generators. Number of plugged tubes: 203 including the increase of 12 this time. Ratio of plugging: 3.0 %. Heating tubes: Inconel 600 tubes of φ22.7 mm x 1.27 mm thickness. (K.I.)

  5. Indian Point 2 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    Dayan, A.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Cheng Weili; Finnie, I.

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    Malagola, P.; Vassal, J.M.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

    Yang Ruichang; Liu Ruolei; Liu Jinggong; Qin Shiwei

    2008-01-01

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

  9. Staged fracturing of horizontal shale gas wells with temporary plugging by sand filling

    Directory of Open Access Journals (Sweden)

    Xing Liang

    2017-03-01

    Full Text Available Due to downhole complexities, shale-gas horizontal well fracturing in the Sichuan Basin suffered from casing deformation and failure to apply the technique of cable-conveyed perforation bridge plug. In view of these problems, a new technique of staged volume fracturing with temporary plugging by sand filling is employed. Based on theoretical analyses and field tests, a design of optimized parameters of coiled tubing-conveyed multi-cluster sand-blasting perforation and temporary plugging by sand filling was proposed. It was applied in the horizontal Well ZJ-1 in which casing deformation occurred. The following results are achieved in field operations. First, this technique enables selective staged fracturing in horizontal sections. Second, this technique can realize massive staged fracturing credibly without mechanical plugging, with the operating efficiency equivalent to the conventional bridge plug staged fracturing. Third, full-hole is preserved after fracturing, thus it is possible to directly conduct an open flow test without time consumption of a wiper trip. The staged volume fracturing with temporary plugging by sand filling facilitated the 14-stage fracturing in Well ZJ-1, with similar SRV to that achieved by conventional bridge plug staged fracturing and higher gas yield than neighboring wells on the same well pad. Thus, a new and effective technique is presented in multi-cluster staged volume fracturing of shale gas horizontal wells.

  10. An optimal design for millimeter-wide facture plugging zone

    Directory of Open Access Journals (Sweden)

    Yili Kang

    2015-01-01

    Full Text Available Lost circulation control in millimeter-wide fractures has been a challenge in well drilling all the time. Low pressure-bearing capacity of a plugging zone will result in excessive consumption of lost circulation materials (LCMs and extra down time. In this study, laboratory experiments were conducted on the plugging of millimeter-wide fractures to evaluate the plugging effects of different types of LCM including rigid granules, elastic particles and fiber. Maximum plugging pressure, total loss volume before sealing and plugging time were taken as the evaluation index of the LCM plugging effect. According to the experimental results, the synergistic plugging mechanisms of different LCM combinations were also analyzed. Experimental results showed that the total loss volume of the plugging zone formed by rigid and elastic particle combination was generally greater than 400 mL, and the maximum plugging pressure of the plugging zone formed by elastic particle and fiber combination was generally less than 6 MPa. In contrast, the plugging zone formed by the combination of the three types of LCMs has the maximum plugging pressure of up to 13 MPa and total loss volume before sealing of 75 mL. In the synergistic plugging process, rigid granules form a frame with high pressure-bearing capacity in the narrower parts of the fractures; elastic particles generate elastic force through elastic deformation to increase the friction between a fracture and a plugging zone to make the plugging zone more stable; fibers filling in the pore space between the particles increase the tightness and integrity of the plugging zone. The experimental results can provide guidance for the optimal design of LCMs used in the field.

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

    International Nuclear Information System (INIS)

    Saniie, J.; Nagle, D.T.

    1987-01-01

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

  12. 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.

  13. 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

  14. 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)

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

    International Nuclear Information System (INIS)

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

    1978-03-01

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

  16. 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.

  17. Method to detect steam generator tube leakage

    International Nuclear Information System (INIS)

    Watabe, Kiyomi

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1995-01-01

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

  19. 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

  20. Flow induced pulsations generated in corrugated tubes

    NARCIS (Netherlands)

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

    2008-01-01

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

  1. Flow induced vibration studies on PFBR control plug components

    Energy Technology Data Exchange (ETDEWEB)

    Prakash, V., E-mail: prakash@igcar.gov.in [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu (India); Kumar, P. Anup; Anandaraj, M.; Thirumalai, M.; Anandbabu, C.; Rajan, K.K. [Fast Reactor Technology Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamilnadu (India)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Flow induced vibration studies on Prototype Fast Breeder Reactor control plug model carried out. Black-Right-Pointing-Pointer Velocity similitude was followed for the study. Black-Right-Pointing-Pointer Frequencies and amplitude of vibrations of various control plug components measured. Black-Right-Pointing-Pointer Overall values of vibration are well within permissible limits. - Abstract: The construction of Prototype Fast Breeder Reactor (PFBR), a 500 MWe liquid sodium cooled reactor, is in progress at Kalpakkam in India. Control plug (CP) is located right above the core subassemblies in the hot pool. Control plug is an important component as many of the critical reactor parameters are sensed and controlled by the components housed in the control plug assembly. In PFBR primary circuit, components are basically thin walled, slender shells with diameter to thickness ratio ranging from 100 to 650. These components are prone to flow induced vibrations. The existence of free liquid (sodium) surfaces, which is the source of sloshing phenomenon and the operation of primary sodium pump in the primary pool are other potential sources of vibration of reactor components. Control plug is a hollow cylindrical shell structure and provides passages and support for 12 absorber rod drive mechanisms (ARDM) which consists of 9 control and safety rods and 3 diverse safety rods, 210 thermo wells to measure the sodium temperature at the exit of various fuel subassemblies, three failed fuel localization modules (FFLM) and acoustic detectors. It consists of a core cover plate (CCP), which forms the bottom end, two intermediate supports plate, i.e. lower stay plate (LSP) and upper stay plate (USP) and an outer shell. The CCP is located at a distance of 1.3 m from the core top. With such a gap, there will be long free hanging length of the thermocouple sleeves, Delayed neutron detector (DND) sampling tubes and ARDM shroud tubes and hence they are

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

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1987-01-01

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

  3. 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.

  4. 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

  5. Duplex-tube sodium-indication steam generator

    International Nuclear Information System (INIS)

    Matal, O.; Martoch, J.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-02-01

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

  7. 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

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

    International Nuclear Information System (INIS)

    Roach, W.H.

    1985-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  10. 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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  13. Vascular plugs - A key companion to Interventionists - 'Just Plug it'.

    Science.gov (United States)

    Ramakrishnan, Sivasubramanian

    2015-01-01

    Vascular plugs are ideally suited to close extra-cardiac, high flowing vascular communications. The family of vascular plugs has expanded. Vascular plugs in general have a lower profile and the newer variants can be delivered even through a diagnostic catheter. These features make them versatile and easy to use. The Amplatzer vascular plugs are also used for closing intracardiac defects including coronary arterio-venous fistula and paravalvular leakage in an off-label fashion. In this review, the features of currently available vascular plugs are reviewed along with tips and tricks of using them in the cardiac catheterization laboratory. Copyright © 2015. Published by Elsevier B.V.

  14. Trouble found during regular inspection of No.3 plant in Mihama Power Station, Kansai Electric Power Co., Inc

    International Nuclear Information System (INIS)

    1990-01-01

    No.3 plant in Mihama Power Station, Kansai Electric Power Co., Inc. is a PWR type plant with the rated output of 826 MWe. Its regular inspection has been carried out since September 11, 1989, and eddy current flaw detection inspection was carried out on the total number of steam generator heating tubes (9997 tubes except already plugged tubes). As the result, significant indication was observed in 24 tubes in the expanded parts in tube plates, and in 36 tubes at the boundary of the expanded parts (all on high temperature side), in total in 60 tubes. Consequently, it was decided to plug these 60 defective heating tubes. The heating tubes are those made of Inconel 600, having 22.2 mm outside diameter and 1.27 mm wall thickness. The total number of heating tubes in 10164 (3388 tubes x 3 steam generators), the number of plugged tubes is 227, and the ratio of plugging is 2.2 %. (K.I.)

  15. Eddy current testing of steam generator tubes

    International Nuclear Information System (INIS)

    Neumaier, P.

    1981-01-01

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

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

    International Nuclear Information System (INIS)

    Griesbach, T.; Cipolla, R.

    1985-01-01

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

  17. Preparation and Evaluation of Biodegradable Scleral Plug Containing Curcumin in Rabbit Eye.

    Science.gov (United States)

    Zhang, Jun; Sun, Haiyan; Zhou, Nalei; Zhang, Bin; Ma, Jingxue

    2017-12-01

    To test whether biodegradable curcumin-loaded scleral plug is a promising choice for treating posterior ocular diseases, the study investigated the in vitro release profile of the scleral plug and its safety in vivo. Scleral plugs containing 0.5 mg, 1.0 mg and 1.5 mg curcumin were synthesized by a compression-sintering method. These scleral plugs were placed in tubes containing balanced salt solution (BSS) buffer, which was replaced by fresh buffer daily. The curcumin concentration in the removed aliquot was tested daily for 14 days using high-performance liquid chromatography (HPLC). In the study, 44 rabbits were randomly divided into four groups: control, 0.5 mg, 1.0 mg and 1.5 mg curcumin groups. The scleral plug was trans-scleral fixed in the right eye of the rabbits in the three curcumin-treated groups. The control rabbits only received sclerotomy. The treated rabbit eyes were examined by a slit-lamp biomicroscope, an indirect ophthalmoscope and electroretinogram (ERG), and subjected to histological analysis. The concentration of the 1.5 mg curcumin-loaded scleral plug was higher than 15 μg/ml for consecutive 14 days in vitro. The in vivo experiments revealed intraocular pressure, a-wave and b-wave amplitudes of ERG, and conjunctival reaction degree were not significantly different between the four groups. Retinal structure was normal in the curcumin-treated groups. The sclerotomy wound healed after the plug was completely degraded. Anterior chamber reaction or complications were not observed. The study suggests that curcumin-loaded scleral plug could sustain high concentration of curcumin in vitro and is safe in vivo. It might be a promising alternative choice for the treatment of posterior ocular diseases.

  18. 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)

  19. Viewing device of a steam generator tube-plate

    International Nuclear Information System (INIS)

    Denis, J.; Poirier, D.

    1984-01-01

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

  20. Effect of heat transfer tube leak on dynamic characteristic of steam generator

    International Nuclear Information System (INIS)

    Sun Baozhi; Shi Jianxin; Li Na; Zheng Lusong; Liu Shanghua; Lei Yu

    2015-01-01

    Taking the steam generator of Daya Bay Nuclear Power Station as the research object, one-dimensional dynamic model of the steam generator based on drift flux theory and leak model of heat transfer tube were established. Steady simulation of steam generator under different conditions was carried out. Based on verifying the drift flux model and leak model of heat transfer tube, the effect of leak location and flow rate under different conditions on steam generator's key parameters was studied. The results show that the drift flux model and leak model can reflect the law of key parameter change accurately such as vapor mass fraction and steam pressure under different leak cases. The variation of the parameters is most apparent when the leak is at the entrance of boiling section and vapor mass fraction varies from 0.261 to 0.163 when leakage accounts for 5% of coolant flow rate. The successful prediction of the effect of heat transfer tube leak on dynamic characteristics of the steam generator based on drift flux theory supplies some references for monitoring and taking precautionary measures to prevent heat transfer tube leak accident. (authors)

  1. Heat exchanging tube behaviour in steam generators of pressurized water reactors

    International Nuclear Information System (INIS)

    Pastor, D.; Oertel, K.

    1979-01-01

    Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

  2. 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)

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

    International Nuclear Information System (INIS)

    Zhang Fuyuan

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  5. Design Concept of Array ECT Sensor for Steam Generator Tubing Inspection

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chan Hee; Lee, Tae Hun; Yoo, Hyun Ju [Korea Hydro and Nuclear Power Co. Ltd. CRI, Daejeon (Korea, Republic of)

    2015-05-15

    The eddy current testing, which is one of the nondestructive examination methods, is widely used for the inspection of heat exchangers including steam generator tubing in the nuclear power plant. It uses electromagnetic induction to detect flaws in conductive materials. Two types of eddy current probes are conventionally used for the inspection of steam generator tubing according to the main purpose. One is the bobbin probe technology and the other is the rotating probe. During the inspection, they have restrictions for the flaw detection or the inspection speed. An array probe can be alternative to the bobbin and rotating probes. The design concept of array coils with high sensitivity is described in this paper. It is expected that the eddy current testing using this type of array sensors may provide high detectability and resolution for flaws in steam generator tubing. Eddy current technology has some barriers for the inspection of steam generator tubing in the nuclear power plant. Bobbin probes offer poor circumferential crack detection and rotating probes are time and money consuming due to the mechanical rotation. Array probe inspection technique can replace bobbin and rotating probe techniques due to its sensitivity for flaw detection and inspection speed. In general, circular-shaped coils are considered in an array eddy current probe.

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

    International Nuclear Information System (INIS)

    Hayes, J.K.

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

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

    Science.gov (United States)

    Verbeke; Leung; Vujic

    2000-10-01

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

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  13. Hideout in steam generator tube deposits

    International Nuclear Information System (INIS)

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

    1998-05-01

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

  14. Clinical application of continent anal plug in bedridden patients with intractable diarrhea.

    Science.gov (United States)

    Kim, J; Shim, M C; Choi, B Y; Ahn, S H; Jang, S H; Shin, H J

    2001-08-01

    Some patients bedridden from various causes such as stroke or spinal cord injury experience poor control of bowel movement. This causes fecal leakage and diarrhea, increases the risk of perianal excoriation and bed sores, and is a burden on caregivers. To evaluate the efficacy of fecal evacuation and the prevention and treatment of skin complications in intractable diarrhea patients using a new device. A continent anal plug (US Patent No. 5 569 216) comprises an inner balloon surrounded by an outer balloon, both of which are mounted on a silicone tube containing a pair of air passages and an enema fluid inlet. The tube is secured in place in the rectum by the inflatable outer balloon and is designed to drain fecal matter through a thin collapsible hose situated in the anal canal. Thirty-two patients (21 male; median age 61 (range, 28-76) years) were evaluated after fully informed consent. Median duration was 12 (range, 3-37) days. The continent anal plug evacuated efficiently in those patients with loose or watery stools who only required irrigation once daily or not at all. Skin excoriations improved in three to seven days. Minimal leakage was seen around the anus. There was no anorectal mucosal injury noted over 37 days. The continent anal plug is an efficient method of treating patients with loss of bowel control and incontinence because it enables controlled fecal evacuation and helps reduce skin complications without causing anorectal mucosal injury.

  15. Oscillation of an isolated liquid plug inside a dry capillary

    Science.gov (United States)

    Srinivasan, Vyas; Kumar, Siddhartha; Asfer, Mohammed; Khandekar, Sameer

    2017-11-01

    The present work reports an experimental study on the dynamics of partially wetting isolated liquid plug (DI water), which is made to oscillate inside a square, glass capillary tube (1 mm × 1 mm; 60 mm length). The liquid plug is made to oscillate pneumatically at two different frequencies (0.25 and 0.35 Hz), using a cam-follower mechanism. Bright field imaging is used to visualize the three-phase contact line behavior, while, micro-Particle Imaging Velocimetry (PIV) apparatus is used to discern the nature of flow inside the oscillating liquid plug. During a cycle, due to the partial wetting nature of DI water, the three-phase contact line at the menisci gets pinned at the extreme end of each stroke, where the dynamic apparent contact angle gets drastically altered before the initiation of the next stroke. The difference between the apparent contact angle of the front and rear meniscus are seen to be a function of the oscillating frequency; the difference increasing with increasing frequency. The flow inside the liquid plug reveals unique non-Poiseuille flow features near the meniscus, due to free-slip boundary condition, which leads to formation of distinct vortex pairs behind it. The vortices too change their direction during each stroke of the oscillation, eventually leading to an alternating recirculation pattern inside the plug. The results clearly indicate that improved mathematical models are required for predicting transport parameters in such flows, which are important in engineering systems such as pulsating heat pipes, lab-on-chip devices and PEM fuel cells.

  16. 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.

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  18. Tube tightness survey during Phenix steam generator operation

    International Nuclear Information System (INIS)

    Cambillard, E.

    1976-01-01

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

  19. Experimental and numerical modeling of sulfur plugging in a carbonate oil reservoir

    Energy Technology Data Exchange (ETDEWEB)

    Al-Awadhy, F. [ADMA-OPCO, Abudhabi (United Arab Emirates); Kocabas, I.; Abou-Kassem, J.H. [UAE University, Al Ain (United Arab Emirates); Islam, M.R. [Dalhousie University, Halifax, NS (United States)

    2005-01-15

    Many oil and gas reservoirs in the United Arab Emirates produce large amounts of sour gas, mainly in the form of hydrogen sulfide. In addition to creating problems in the production line, wellbore damage is often reported due to the precipitation of elemental sulfur in the vicinity of the wellbore. While there have been several studies performed on the role of solid deposition in a gas reservoir, the role of sulfur deposition in oil reservoirs has not been investigated. This article presents experimental results along with a comprehensive wellbore model that predicts sulfur precipitation as well as plugging. The experiments were conducted in a core (linear) system. Both analytical and numerical modelings were performed in a linear coordinate system. Data for the numerical model was obtained from both test tube and coreflood experiments. By using a phenomenological model, the wellbore plugging was modeled with an excellent match (with experimental results). The crude oil was de-asphalted prior to conducting the experiment in order to isolate the effect of asphaltene plugging. A series of coreflood tests was carried out to observe sulfur precipitation and plugging in a carbonate rock. Significant plugging was observed and was found to be dependent on flow rate and initial sulfur concentration. This information was used in the phenomenological model and can be incorporated in the wellbore numerical model. (author)

  20. Drill pipe bridge plug

    International Nuclear Information System (INIS)

    Winslow, D.W.; Brisco, D.P.

    1991-01-01

    This patent describes a method of stopping flow of fluid up through a pipe bore of a pipe string in a well. It comprises: lowering a bridge plug apparatus on a work string into the pipe string to a position where the pipe bore is to be closed; communicating the pipe bore below a packer of the bridge plug apparatus through the bridge plug apparatus with a low pressure zone above the packer to permit the fluid to flow up through the bridge plug apparatus; engaging the bridge plug apparatus with an internal upset of the pipe string; while the fluid is flowing up through the bridge plug apparatus, pulling upward on the work string and the bridge plug apparatus and thereby sealing the packer against the pipe bore; isolating the pipe bore below the packer from the low pressure zone above the packer and thereby stopping flow of the fluid up through the pipe bore; disconnecting the work string from the bridge plug apparatus; and maintaining the bridge plug apparatus in engagement with the internal upset and sealed against the pipe bore due to an upward pressure differential applied to the bridge plug apparatus by the fluid contained therebelow

  1. 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

  2. High temperature technological heat exchangers and steam generators with helical coil assembly tube bundle

    International Nuclear Information System (INIS)

    Korotaev, O.J.; Mizonov, N.V.; Nikolaevsky, V.B.; Nazarov, E.K.

    1990-01-01

    Analysis of thermal hydraulics characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the basis of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given. The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the following types: straight tubes, field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conjunction with forced and natural mixing in connecting lines. Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers and steam generators within the wide range of operating conditions without additional expenses on the research and development work

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  4. Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

    International Nuclear Information System (INIS)

    Mohany, A.; Feenstra, P.; Janzen, V.P.; Richard, R.

    2009-01-01

    Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

  5. 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

  6. Evaluating Steam Generator Tubing Corrosion through Shutdown Nickel and Cobalt Releases

    International Nuclear Information System (INIS)

    Marks, Chuck; Little, Mike; Krull, Peter; Dennis Hussey; Kenny Epperson

    2012-09-01

    During power operation in PWRs, steam generator tubing corrodes. In PWRs with nickel alloy steam generator tubing this leads to the release of nickel into the coolant. While not structurally significant, this process leads to corrosion product deposition on the fuel surfaces that can threaten fuel integrity, provide a site for boron precipitation, and, through activation and subsequent release, lead to increased out-of-core radiation fields. During shutdown, decreases in temperature and pH and an increase in the oxidation potential lead to dissolution of some corrosion products from the core. This work evaluated the masses of corrosion products released during shutdown as a proxy for steam generator tubing corrosion rates. The masses were evaluated for trends with time (e.g., the number of cycles) and for the influence of design and operating features such as tubing manufacturer, plant design (e.g., three loop versus four loop), and operating chemistry program. This project utilized the EPRI PWR Chemistry Monitoring and Assessment database. Data from over 20 units, many over several cycles, were assessed. The focus was on corrosion product release from Alloy 690TT tubing and all data were from units that had replaced steam generators. Data were analyzed using models developed from corrosion rate test data reported in the literature with a heavy reliance on data from the EDF BOREAL testing. The most striking result of this analysis was a clear division between plants that exhibited corrosion with a falling rate (i.e., following an exponential decay as has been observed, for example, in the BOREAL testing) and those that showed a constant corrosion rate, sustained for many outages. This difference appears to be most closely correlated with the manufacturer of the tubing. Within the two distinct plant groups (decaying corrosion rate and constant corrosion rate), details of the trends were evaluated for correlation with zinc addition history, plant type, and operating

  7. The Development of Expansion Plug Wedge Test for Clad Tubing Structure Mechanical Property Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jy-An John [ORNL; Jiang, Hao [ORNL

    2016-01-12

    To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at the Oak Ridge National Laboratory (ORNL) and is described fully in US Patent Application 20060070455, “Expanded plug method for developing circumferential mechanical properties of tubular materials.” This method is designed for testing fuel rod cladding ductility in a hot cell using an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of the specimen’s strain. It was also found that cladding strength could be determined from the test results.

  8. Plug-in vehicles and the future of road infrastructure funding in the United States

    International Nuclear Information System (INIS)

    Dumortier, Jerome; Kent, Matthew W.; Payton, Seth B.

    2016-01-01

    In the United States, road infrastructure funding is declining due to an increase in fuel efficiency and the non-adjustment of fuel taxes to inflation. Legislation to tax plug-in vehicles has been proposed or implemented in several states. Those propositions are contrary to policies to promote fuel efficient vehicles. This paper assesses (1) the magnitude of the decline in federal fuel tax revenue caused by plug-in vehicles and (2) quantifies the revenue that could be generated from a federal plug-in vehicle registration fee. We find that the contribution of plug-in vehicles to the decline of the federal fuel tax revenue is at most 1.6% and the majority of the shortfall can be attributed to the non-adjustment of the fuel tax rate and the increase in vehicle fuel efficiency by 2040. An additional tax of $50–$200 per plug-in vehicle per year in the reference case would generate $188–$745 million in 2040 which represents an increase of 1.69–6.71% in federal fuel tax revenue compared to no tax. The lesson for policy makers is that plug-in vehicles do not contribute significantly to the funding shortfall in the short- and medium-run and a supplemental tax would generate a small percentage of additional revenue. - Highlights: •Fees on plug-in cars are proposed or implemented to collect foregone fuel taxes. •Plug-in cars are responsible for a very small percentage of declining tax revenue. •An additional tax on plug-in cars does not stop the decline in fuel tax revenue. •Adjusting fuel taxes to inflation is a more effective tool to increase tax revenue.

  9. Application of laser-based profilometry to tubing in power generating utilities

    Science.gov (United States)

    Doyle, James L.

    1995-05-01

    Over the past several years lasers have been employed in an ever widening number of applications in an incredibly diverse set of markets. In the area of nondestructive testing, however, laser-based systems have only recently made inroads into the commercial markets. About ten years ago QUEST Integrated, Inc., began working with the U.S. Navy to adapt the principal of laser triangulation to solve a serious maintenance related problem. The internal surfaces of marine boiler tubes were experiencing pitting and corrosion which had resulted in catastrophic shipboard failures. At that time, conventional visual methods only allowed operators to inspect the first eighteen inches of the tube using a rigid borescope. If any pits were located, a mechanical stylus mechanism was used to obtain an approximate depth measurement of the pit. The condition of the balance of the tube was then extrapolated based on this extremely limited amount of information. Often the worst pitting was found in the bends of the tube, which could not be inspected by the visual method. Finally, a catastrophic boiler failure on an aircraft carrier resulted in the initiation of a search by the U.S. Navy for a better solution. Quest was contracted to develop an articulated probe which could negotiate the full length of a boiler tube with multiple bends, and generate a complete digital map of the inside surface. A key requirement of this probe would be rapid and quantitative measurement of internal features such as ID pits and corrosion. In 1987 QUEST delivered the first laser- optic tube inspection system to the U.S. Navy for use in marine boiler tubes. The Laser Optic Tube Inspection System (LOTISTM) was immediately put to use and paid for itself many times over in reduced maintenance costs. Over the next six years several generations of LOTIS were developed for the U.S. Navy, each one providing more capabilities, improved inspection speeds, and more user friendly operator interface. Today, LOTIS is

  10. Random excitation forces in tube bundles subjected to two-phase cross-flow

    International Nuclear Information System (INIS)

    Taylor, C.E.; Pettigrew, M.J.; Currie, I.G.

    1996-01-01

    Data from two experimental programs have been analyzed to determine the characteristics of the random excitation forces associated with two-phase cross-flow in tube bundles. Large-scale air-water flow loops in France and Canada were used to generate the data. Tests were carried out on cantilevered, clamped-pinned, and clamped-clamped tubes in normal-square, parallel-triangular, and normal-triangular configurations. Either strain gages or force transducers were used to measure the vibration response of a centrally located tube as the tue array was subjected to a wide range of void fractions and flow rates. Power spectra were analyzed to determine the effect of parameters such as tube diameter, frequency, flow rate, void fraction, and flow regime on the random excitation forces. Normalized expressions for the excitation force power spectra were found to be flow-regime dependent. In the churn flow regime, flow rate and void fraction had very little effect on the magnitude of the excitation forces. In the bubble-plug flow regime, the excitation forces increased rapidly with flow rate and void fraction

  11. Retained Herrick Plug

    Directory of Open Access Journals (Sweden)

    Justin B. Hellman

    2018-05-01

    Full Text Available A 79-year-old female with a history of keratoconjunctivitis sicca presented with several years of epiphora of both eyes. Thirteen years earlier, intracanalicular Herrick lacrimal plugs (Lacrimedics, Eastsound, WA, USA had been placed in both eyes to treat her dry eye syndrome. After 13 years the patient felt the epiphora was intolerable and underwent endoscopic dacryocystorhinostomy (DCR of the left, then the right side. Intraoperatively, during the right endoscopic DCR, a Herrick lacrimal plug was found in the common canaliculus into the lacrimal sac. Postoperatively, the patient did well with improved epiphora. The Herrick plug is designed to be intracanalicular, and this case illustrates that the plug can migrate and be retained for many years. Collared punctal plugs have a lower risk of this type of complication.

  12. Experimental facility design for study of fretting in steam generator tubes

    International Nuclear Information System (INIS)

    Balbiani, J.P.; Bergant, M.; Yawny, A.

    2012-01-01

    The design of an experimental facility for fretting wear testing of steam generator tubes under pressurized water up to 340 o C, is presented. The main component of the device consists in an autoclave which permits to recreate steam generator operating conditions. CAD CATIA V5R18, CAE ABAQUS and ASME Sec. VII Div. 1 (Rules for Construction of Pressure Vessels) were used along the design process. The design of the autoclave included the pressure vessel itself and the necessary flanges and nozzles. In addition, an axial dynamic sealing system was designed to allow for actuation from outside the pressure boundary. Complementary, typical tube - support contact conditions were analyzed and the principal variables affecting their mutual interaction determined. In addition, a simple device which allows performing fretting wear testing on steam generator tubes in air at room temperature was fabricated and the feasibility of a quantitative assessment of different aspects related with the fretting induced damage was explored. Characterization techniques available at Centro Atomico Bariloche, like light microscopy, scanning electron microscopy (SEM), energy dispersive analysis of X-ray (EDAX) and surface damage analysis by optic profilometry were shown to be appropriate for this aim. The designed facility will allow evaluating fretting damage of tubes - support combinations that might be used on the steam generator of the prototype reactor CAREM-25. It is also expected it could be applied to characterize fretting severity in other applications (nuclear fuel elements) (author)

  13. Steam generator operating experience: Update for 1984-1986

    International Nuclear Information System (INIS)

    Frank, L.; Stokley, J.

    1988-06-01

    This report summarizes operational events and degradation mechanisms affecting pressurized water reactor steam generator integrity, provides updated inspection results reported in 1984, 1985, and 1986, and highlights both prevalent problem areas and advances in improved equipment test practices, preventive measures, repair techniques, and replacement procedures. It describes equipment design features of the three major suppliers and discusses 68 plants in detail. Steam generator degradation mechanisms include intergranular stress corrosion cracking, primary water stress corrosion cracking, pitting, intergranular attack, and vibration wear that effects tube integrity and causes leakage. Plugging, sleeving heat treatment, peening, chemical cleaning, and steam generator replacements are described and regulatory instruments and inspection guidelines for nondestructive evaluations and girth weld cracking are discusses. The report concludes that although degradation mechanisms are generally understood, the elimination of unscheduled plant shutdowns and costly repairs resulting from leaking tubes has not been achieved. Highlights of steam generator research and unresolved safety issues are discussed. 21 refs., 8 tabs

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

    International Nuclear Information System (INIS)

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

    1984-10-01

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

  15. Stress relief treatment of Alloy 600 steam generator tubing

    International Nuclear Information System (INIS)

    Rooyen, D. van; Cragnolino, C.

    1994-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-07-01

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

  17. Simulating Porous Magnetite Layer Deposited on Alloy 690TT Steam Generator Tubes.

    Science.gov (United States)

    Jeon, Soon-Hyeok; Son, Yeong-Ho; Choi, Won-Ik; Song, Geun Dong; Hur, Do Haeng

    2018-01-02

    In nuclear power plants, the main corrosion product that is deposited on the outside of steam generator tubes is porous magnetite. The objective of this study was to simulate porous magnetite that is deposited on thermally treated (TT) Alloy 690 steam generator tubes. A magnetite layer was electrodeposited on an Alloy 690TT substrate in an Fe(III)-triethanolamine solution. After electrodeposition, the dense magnetite layer was immersed to simulate porous magnetite deposits in alkaline solution for 50 days at room temperature. The dense morphology of the magnetite layer was changed to a porous structure by reductive dissolution reaction. The simulated porous magnetite layer was compared with flakes of steam generator tubes, which were collected from the secondary water system of a real nuclear power plant during sludge lancing. Possible nuclear research applications using simulated porous magnetite specimens are also proposed.

  18. 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)

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

    International Nuclear Information System (INIS)

    Meurgey, P.; Baumaire, A.

    1990-01-01

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

  20. 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

  1. 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.

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  3. 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

  4. Nickel electroplating as a remedy to steam generator tubing PWSCC

    International Nuclear Information System (INIS)

    Michaut, B.; Steltzlen, F.; Sala, B.; Laire, Ch.; Stubbe, J.

    1993-01-01

    Nickel plating appears to be a versatile process, as the application field, even if always used against PWSCC, is different from plant-to-plant. Its usage has been from a purely preventive action on tubes without defects, to a corrective action on through-wall cracked and leaking tubes. As a background for the large scale on-site operations of Doel 2 in 1990 (345 tubes) and Tihange 2 in 1992 (600 tubes), studies on four points are outlined, i.e. corrosion tests, stress measurements, sulfamate bath quality control, and in-service inspection. In conclusion, it appears that the nickel plating technique, following a case-by-case study, can often be a convenient remedy against Alloy 600 stress corrosion problems. New applications, in locations other than the steam generator field are under consideration

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

    International Nuclear Information System (INIS)

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

    1996-07-01

    The objective of the Improved Eddy-Current ISI for Steam Generator Tubing program is to upgrade and validate eddy-current inspections, including probes, instrumentation, and data processing techniques for inservice inspection of new, used, and repaired steam generator tubes; to improve defect detection, classification, and characterization as affected by diameter and thickness variations, denting, probe wobble, tube sheet, tube supports, copper and sludge deposits, even when defect types and other variables occur in combination; to transfer this advanced technology to NRC's mobile NDE laboratory and staff. This report describes the design of specialized high-speed 16-coil eddy-current array probes. Both pancake and reflection coils are considered. Test results from inspections using the probes in working steam generators are given. Computer programs developed for probe calculations are also supplied

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

    International Nuclear Information System (INIS)

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

    1978-02-01

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

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

    International Nuclear Information System (INIS)

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

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

  8. 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

  9. Ultrastructural features of Mimulus aurantiacus (Scrophulariaceae) pollen tubes in vivo.

    Science.gov (United States)

    Ekici, Nuran; Dane, Feruzan; Olgun, Göksel

    2009-03-01

    The aim of this study is to give information on ultrastructure of in vivo pollen tubes of Mimulus aurantiacus which were collected from the Botanical Garden of the University of California at Berkeley. Materials were prepared according to electron microscopy methods and examined under Zeiss electron microscope. Four zones were examined in the pollen tubes of Mimulus aurantiacus. Apical zone: Mitochondria, smooth endoplasmic reticulum, rough endoplasmic reticulum, dictyosomes and secretory vesicles were observed. Subapical zone: This area contained abundant rough endoplasmic reticulum and occasionally some smooth endoplasmic reticulum. The polysomes, mitochondria, proplastids that contain starch, small vacuoles and a few lipid bodies were detected. Nuclear zone: Both generative and vegetative cell nuclei lie in this zone. The vegetative cell nucleus was large and long. Rough endoplasmic reticulum, mitochondria, ribosomes, dictyosomes, and amyloplasts that are rich of starch were observed. Vacuolation and plug formation zone: Cytoplasm of the tubes was full of large vacuoles. Few organelles such as mitochondria, dictyosome and rough endoplasmic reticulum were detected along their periphery.

  10. Freezing controlled penetration of molten metals flowing through stainless steel tubes

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Spencer, B.W.; Vetter, D.L.; Wesel, R.H.

    1985-01-01

    The freezing controlled penetration potential of molten metals flowing within stainless steel structure is important to the safety assessment of hypothetical severe accidents in liquid metal reactors. A series of scoping experiments has been performed in which molten stainless steel and nickel at various initial temperatures and driving pressures were injected downward and upward into 6.4 and 3.3 mm ID stainless steel tubes filled with argon gas and initially at room temperature. In all tests, there was no evidence that the solid tube wall was wetted by the molten metals. The penetration phenomena are markedly different for downward versus upward injections. The dependency upon tube orientation is explained in terms of the absence of wetting. Complete plugs were formed in all experiments halting the continued injection of melt. Calculations with a fluid dynamics/heat transfer computer code show that the injected masses limited by plugging are consistent with freezing through the growth of a stable solidified layer (crust) of metal upon the solid tube wall. 23 refs., 5 figs., 2 tabs

  11. 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.

  12. Repairing a steam generator tube by inserting a sheath

    International Nuclear Information System (INIS)

    Gaudin, J.P.

    1986-01-01

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

  13. Polysaccharides and bacterial plugging. Final report, 1992--1993

    Energy Technology Data Exchange (ETDEWEB)

    Fogler, H.S.

    1995-02-01

    In situ core plugging experiments and transport experiments, using the model bacteria Leuconostoc m., have been conducted. Results demonstrated that cellular polysaccharide production increases cell distribution in porous media and caused an overall decrease in media permeability. Further, a parallel core plugging experiment was conducted and showed the feasibility of this system to divert injection fluid from high permeability zones into low permeability zones within porous media as is needed for profile modification. To implement this type of application, however, controlled placement of cells and rates of polymer production are needed. Therefore, kinetic studies were performed. A kinetic model was subsequently developed for Leuconostoc m. bacteria. This model is based on data generated from batch growth experiments and allows for the prediction of saccharide utilization, cell generation, and dextran production. These predictions can be used to develop injection strategies for field implementation. Transport and in situ growth micromodel experiments have shown how dextran allow cells to remain as clusters after cell division which enhanced cell capture and retention in porous media. Additional Damkohler experiments have been performed to determine the effects of the nutrient injection rate and nutrient concentration on the rate of porous media plugging. As shown experimentally and as predicted by a model for in situ growth, an increase in nutrient concentration and/or its injection rate will result in a faster rate of porous media plugging. Through continuum model simulations, it has been shown that the initial cell profiles play a key role on the core plugging rate. Controlling the location of the inoculating cells is thus another key factor in using bacteria for profile modification.

  14. Strategic management of steam generators

    International Nuclear Information System (INIS)

    Hernalsteen, P.; Berthe, J.

    1991-01-01

    This paper addresses the general approach followed in Belgium for managing any kind of generic defect affecting a Steam Generator tubebundle. This involves the successive steps of: problem detection, dedicated sample monitoring, implementation of preventive methods, development of specific plugging criteria, dedicated 100% inspection, implementation of repair methods, adjusted sample monitoring and repair versus replacement strategy. These steps are illustrated by the particular case of Primary Water Stress Corrosion Cracking in tube roll transitions, which is presently the main problem for two Belgian units Doele-3 and Tihange-2. (author)

  15. Leak on a steam generator tube: in-depth analysis

    International Nuclear Information System (INIS)

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

    2015-01-01

    A circumferential through crack was observed on a steam generator tube of the unit 2 of the Fessenheim plant. Destructive tests showed that the crack was due to cycle fatigue combined with the presence of inter-granular corrosion zones. An in-depth analysis based on simulations shows that the combination of 5 elements caused the crack. First, a specific position of the anti-vibration bar near this tube, secondly, a local presence of fouling, these 2 first elements led to an increase of the tube vibratory level. Thirdly, the 600 MA alloy used is known to be susceptible to corrosion. Fourthly, the trapping of chemical species on the secondary circuit side due to the presence of interstices on the crosspiece and fifthly, the presence of spots where inter-granular corrosion developed. (A.C.)

  16. Optimization of Peripheral Finned-Tube Evaporators Using Entropy Generation Minimization

    OpenAIRE

    Pussoli, Bruno; Barbosa Jr., Jader; da Silva, Luciana; Kaviany, Massoud

    2012-01-01

    The peripheral finned-tube (PFT) is a new geometry for enhanced air-side heat transfer under moisture condensate blockage (evaporators). It consists of individual hexagonal (peripheral) fin arrangements with radial fins whose bases are attached to the tubes and tips are interconnected with the peripheral fins. In this paper, experimentally validated semi-empirical models for the air-side heat transfer and pressure drop are combined with the entropy generation minimization theory to determine ...

  17. Microcontroller based automation system for end plug welding of test fuel pins in solgel facility

    International Nuclear Information System (INIS)

    Prabhakar Rao, J.; Srinivas, K.C.; Prabhu, T.V.; Ravi, N.

    2010-01-01

    A microcontroller based stepper motor control and driver Unit for 'XY' positioning system is designed and developed to perform the 'pick-place' of fuel tube to pre-determined coordinates. This Unit provides a fine movement of the fuel tube to get perfect position for welding. The Graphical User Interface software running on PC displays the absolute position of the XY system and provides all the required control buttons to achieve the accurate positioning. The welding of clad tube with end plug is carried out in a high precision welding fixture by operating it remotely. This paper discusses about the Hardware and Software features and implementation of the instrumentation. (author)

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro

    1994-11-01

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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  1. Evaluation of ECT reliability for axial ODSCC in steam generator tubes

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  2. Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

  3. Predicted wear on the tube outside surface due to foreign object in the secondary side of steam generator

    International Nuclear Information System (INIS)

    Kim, Hyung Nam; Cho, Nam Cheoul

    2012-01-01

    It is necessary to evaluate the effects of foreign objects on steam generator tubes and to use this information to take appropriate safety precautions to prevent nuclear accidents. Foreign objects may include loose parts from the feed water system and items lost by workers during o/h, and may flow into the secondary side of steam generators during operation. A foreign object could damage steam generator tube walls if there is relative motion between the tube and the foreign object. This is especially true for foreign objects that land on the tube sheet because the velocity of cross flow, which creates a contact force between the tube and foreign object, is relatively high there. During steam generator overhauls, foreign objects are detected by non destructive methods such as the visual test and/or the eddy current test. Confirmed foreign objects should be removed for nuclear safety. The Foreign Object Search and Retrieval System (FOSAR) can be used to remove foreign objects from the steam generators with a square tube array. However, the FOSAR cannot be used (or can be used in only a very restricted area, such as the outside of the tube bundle) in the steam generators with a triangular tube array. In order to continue nuclear power plant operations without removing foreign objects, the integrity of the steam generator tube must be verified. This paper introduces a practical method developed to evaluate the effects of foreign objects detected on tube sheets in the secondary sides of steam generators

  4. Steam generator tubes rupture probability estimation - study of the axially cracked tube case

    International Nuclear Information System (INIS)

    Mavko, B.; Cizelj, L.; Roussel, G.

    1992-01-01

    The objective of the present study is to estimate the probability of a steam generator tube rupture due to the unstable propagation of axial through-wall cracks during a hypothetical accident. For this purpose the probabilistic fracture mechanics model was developed taking into account statistical distributions of influencing parameters. A numerical example considering a typical steam generator seriously affected by axial stress corrosion cracking in the roll transition area, is presented; it indicates the change of rupture probability with different assumptions focusing mostly on tubesheet reinforcing factor, crack propagation rate and crack detection probability. 8 refs., 4 figs., 4 tabs

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-04-01

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

  7. The plug-based nanovolume Microcapillary Protein Crystallization System (MPCS)

    International Nuclear Information System (INIS)

    Gerdts, Cory J.; Elliott, Mark; Lovell, Scott; Mixon, Mark B.; Napuli, Alberto J.; Staker, Bart L.; Nollert, Peter; Stewart, Lance

    2008-01-01

    The Microcapillary Protein Crystallization System (MPCS) is a new protein-crystallization technology used to generate nanolitre-sized crystallization experiments for crystal screening and optimization. Using the MPCS, diffraction-ready crystals were grown in the plastic MPCS CrystalCard and were used to solve the structure of methionine-R-sulfoxide reductase. The Microcapillary Protein Crystallization System (MPCS) embodies a new semi-automated plug-based crystallization technology which enables nanolitre-volume screening of crystallization conditions in a plasticware format that allows crystals to be easily removed for traditional cryoprotection and X-ray diffraction data collection. Protein crystals grown in these plastic devices can be directly subjected to in situ X-ray diffraction studies. The MPCS integrates the formulation of crystallization cocktails with the preparation of the crystallization experiments. Within microfluidic Teflon tubing or the microfluidic circuitry of a plastic CrystalCard, ∼10–20 nl volume droplets are generated, each representing a microbatch-style crystallization experiment with a different chemical composition. The entire protein sample is utilized in crystallization experiments. Sparse-matrix screening and chemical gradient screening can be combined in one comprehensive ‘hybrid’ crystallization trial. The technology lends itself well to optimization by high-granularity gradient screening using optimization reagents such as precipitation agents, ligands or cryoprotectants

  8. Installation of the backfill and plug test

    International Nuclear Information System (INIS)

    Gunnarsson, D.; Borgesson, L.; Hokmark, H.; Hohannesson, L.E.; Sanden, T.

    2003-01-01

    and 2 in the floor of the tunnel, were installed to measure the mechanical properties of the backfill after saturation. The water pressure in the rock is measured in 75 sections in bore holes. Microorganisms were placed in both backfill materials to investigate if they can multiply under the existing conditions. All cables and tubes were led through watertight seals through bore holes to a neighbouring tunnel. The plug is designed to resist water and swelling pressures that may develop. The design includes a 1.5 m deep slot and an 0-ring of highly compacted bentonite in order to cut off the excavation disturbed zone. The installation as a whole worked well. The bulk average dry density was estimated to be between 1650 and 1700 kg/m 3 for the 30/70 backfill material and the average measured dry density of the 0/100 material was 2170 kg/m 3 . The casting of the plug went according to plan. After the installation of the test the water saturation of the backfill started. The water saturation is expected to be completed in the beginning of 2003 and then the subsequent flow testing will start. (authors)

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-04-01

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

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

    Directory of Open Access Journals (Sweden)

    Lukjanov Alexander V.

    2014-12-01

    Full Text Available Boiler construction is one of the major industries of any state. The aim is to determine the effect of the turbulator on the intensity of heat transfer in the convective part of the fire-tube heat generator of domestic production. The improvement of convective heating surfaces is one of the ways to increase the energy efficiency of the fire-tube heat generator. Since model of the process of heat transfer of gas flow in the convective tubes is multifactorial and does not have clear analytical solution at present, the study of process above is carried out using the experimental method. The results of applying the flow turbulator as a broken tape in the fire-tube heat generator of KV-GM type are presented. On their basis it can be concluded about increasing of heat transfer in convective part of the unit. The use of efficient, reliable, easy to manufacture, relatively inexpensive turbulator in domestic fire-tube heat generators will allow to increase their energy conversion efficiency and reduce fuel consumption, which will have a positive economic effect.

  12. Improved in-service inspection program for management of degradation in steam generator tubing

    International Nuclear Information System (INIS)

    Kurtz, R.; Heasler, P.; Muscara, J.

    1992-01-01

    This paper presents an overview of significant results from NRC-sponsored research on steam generator tube integrity and inspection. Burst test results are described along with empirical models to relate flaw geometry and size to tube burst pressure. Results of round robin examinations of a retired-from-service steam generator to determine eddy current inspection reliability are presented. An evaluation and comparison of various sampling plans for in-service inspection of steam generators is discussed. Finally, performance demonstration qualification efforts for eddy current inspection systems are described

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

    International Nuclear Information System (INIS)

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

    1992-10-01

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

  14. 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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  16. Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

    International Nuclear Information System (INIS)

    1980-03-01

    The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

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

    International Nuclear Information System (INIS)

    Choi, E.I.; Jansen, D.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    1998-03-01

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

  19. Limits to the Recognizability of Flaws in Non-Destructive Testing Steam-Generator Tubes for Nuclear-Power Plants

    International Nuclear Information System (INIS)

    Kuhlmann, A.; Adamsky, F.-J.

    1965-01-01

    In the Federal Republic of Germany there are nuclear reactors under construction with steam generators inside the reactor pressure-vessel. As a result design repairs of steam- generator tubes are very difficult and cause large shut-down times of the nuclear-power plant. It is known that numerous troubles in operating conventional power plants are results of steam-generator tube damages. Because of the high total costs of these reactors it. is necessary to construct the steam generators especially in such a manner that the load factor of the power plant is as high as possible. The Technischer Überwachungs-Verein Rheinland was charged to supervise and to test fabrication and construction of the steam generators to see that this part of the plant was as free of defects as possible. The experience gained during this work is of interest for manufacture and construction of steam generators for nuclear-power plants in general. This paper deals with the efficiency limits of non-destructive testing steam-generator tubes. The following tests performed will be discussed in detail: (a) Automatic ultrasonic testing of the straight tubes in the production facility; (b) Combined ultrasonic and radiographic testing of the bent tubes and tube weldings; (c) Other non-destructive tests. (author) [fr

  20. Friction Pull Plug and Material Configuration for Anti-Chatter Friction Pull Plug Weld

    Science.gov (United States)

    Littell, Justin Anderson (Inventor)

    2016-01-01

    A friction pull plug is provided for use in forming a plug weld in a hole in a material. The friction pull plug includes a shank and a series of three frustoconical sections. The relative sizes of the sections assure that a central one of the sections defines the initial contact point between the hole's sides. The angle defined by the central one of the sections reduces or eliminates chatter as the plug is pulled into the hole.

  1. Rotary plug

    International Nuclear Information System (INIS)

    Yamada, Keiji.

    1979-01-01

    Purpose: In a rotating plug of a reactor using a liquid metal as a coolant as in the case of a fbr type reactor, to prevent the ingress of Na vapor into the sliding surface of the rotating plug. Constitution: A rotating plug comprising a large disc covering the upper part of a reactor pressure vessel containing therein a liquid metal and forming surfaces for mounting a shielding device and various other devices, and at least one or more of small discs provided rotatably and eccentrically within said large disc, which is characterized in that an elastic member consisting of bellows and a sealing is interposed between said large disc and said small discs. (Nakamura, S.)

  2. Stability of single-phase natural circulation with inverted U-tube steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sanders, J.

    1988-08-01

    For natural circulation it is shown that parallel flow in the tubes of an inverted U-tube stream generator can be, at certain power levels, unstable. A mathematical model, based on one-dimensional Oberbeck-Boussinesq equations, shows that stability can be attained if in some tubes the water flows backward, opposite to the normal flow direction. The results are compared to measurements obtained from the natural circulation test A2-77A in the LOBI-MOD2 integral system test facility.

  3. Structural integrity investigations of feeder pipe ice plugging procedures

    International Nuclear Information System (INIS)

    Flaman, M.T.; Shah, N.N.

    1985-03-01

    A procedure involving the use of a liquid nitrogen cooled heat exchanger to form internal ice plugs in feeder pipes is routinely used in nuclear generating stations. The use of this procedure has caused concerns with regard to the safety of station maintenance personnel, and in regard to the integrity of the feeder pipes. This report describes the results of laboratory stress and pressure measurements which were performed on a feeder pipe section during ice plugging operations to investigate these concerns. From the results of this study, and from the results of previous studies of material behaviour at low temperatures, it has been determined that the ice plugging procedure can be performed on feeder pipes in a safe and effective manner

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

    International Nuclear Information System (INIS)

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

    1994-07-01

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

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

    International Nuclear Information System (INIS)

    Park, Jae Seok

    2007-02-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-31

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  8. Double-plug seismic connector

    International Nuclear Information System (INIS)

    Annoot, I.R.

    1984-01-01

    The double-plug seismic electric connector comprises an elongated insulating body. A male connector plug is at one end of the body and a female connector plug is at the opposite end of the body. Each plug has a pair of male and female contacts. A pair of spaced axial conductors is embedded within the insulating body for interconnecting the opposite male and female contacts. The inner end of a double-wire cable is embedded within the insulating body and each wire inner end is connected to one of the conductors

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  10. Development of a gradient tube method for examining microbial ...

    African Journals Online (AJOL)

    2013-06-24

    Jun 24, 2013 ... as brackish and salt marshes and these biofilms may be an important component ... the bottom of a sterile 15 mℓ glass test tube, a 5 mℓ molten agarose plug ... B. Field-scale floating sulphur biofilm reactor developed to enable.

  11. Hideout of sea water impurities in steam generator tube deposits: laboratory and field studies

    International Nuclear Information System (INIS)

    Balakrishnan, P.V.; Turner, C.W.; Thompson, R.; Sawochka, S.

    1996-01-01

    Sea water impurities hide out within thin (∼10 μm) deposits on steam generator tubes, as demonstrated by both laboratory studies using segments of fouled steam generator tubes pulled in 1992 from Crystal River-3 nuclear power station and field hideout return studies performed during recent plant shutdowns. Laboratory tests performed at 279 o C (534 o F) and heat fluxes ranging from 35 to 114 kW/m 2 (11,100 - 36,150 Btu/h.ft 2 ), conditions typical of the lower tubesheet to the first support plate region of a once-through steam generator, showed that impurity hideout can occur in thin free-span tube deposits. The extent of hideout increased with increasing heat flux. Soluble species, such as sodium and chloride ions, returned promptly to the bulk water from the deposits when the heat flux was turned off, whereas less soluble species, such as calcium sulfate and magnesium hydroxide, returned more slowly. Recent field hideout return studies performed at Crystal River-3 where the water level in the steam generators was maintained below the first tube support plate during the shutdown, thus wetting only the thin deposits in the free span and the small sludge pile, corroborate the laboratory findings, showing that hideout does indeed occur in the free-span regions of the tubes. These findings suggest that hideout within tube deposits has to be accounted for in the calculation of crevice chemistry from hideout return studies and in controlling the bulk chemistry using the molar ratio criterion. (author). 3 refs., 4 tabs., 3 figs

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

    International Nuclear Information System (INIS)

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

    1977-06-01

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

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

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1986-01-01

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

  14. Plug pattern optimization for gamma knife radiosurgery treatment planning

    International Nuclear Information System (INIS)

    Zhang Pengpeng; Wu, Jackie; Dean, David; Xing Lei; Xue Jinyue; Maciunas, Robert; Sibata, Claudio

    2003-01-01

    Purpose: To develop a novel dose optimization algorithm for improving the sparing of critical structures during gamma knife radiosurgery by shaping the plug pattern of each individual shot. Method and Materials: We first use a geometric information (medial axis) aided guided evolutionary simulated annealing (GESA) optimization algorithm to determine the number of shots and isocenter location, size, and weight of each shot. Then we create a plug quality score system that checks the dose contribution to the volume of interest by each plug in the treatment plan. A positive score implies that the corresponding source could be open to improve tumor coverage, whereas a negative score means the source could be blocked for the purpose of sparing normal and critical structures. The plug pattern is then optimized via the GESA algorithm that is integrated with this score system. Weight and position of each shot are also tuned in this procedure. Results: An acoustic tumor case is used to evaluate our algorithm. Compared to the treatment plan generated without plug patterns, adding an optimized plug pattern into the treatment planning process boosts tumor coverage index from 95.1% to 97.2%, reduces RTOG conformity index from 1.279 to 1.167, lowers Paddick's index from 1.34 to 1.20, and trims the critical structure receiving more than 30% maximum dose from 16 mm 3 to 6 mm 3 . Conclusions: Automated GESA-based plug pattern optimization of gamma knife radiosurgery frees the treatment planning team from the manual forward planning procedure and provides an optimal treatment plan

  15. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  18. Calculation of relative tube/tube support plate displacements in steam generators under accident condition loads using non-linear dynamic analysis methodologies

    International Nuclear Information System (INIS)

    Smith, R.E.; Waisman, R.; Hu, M.H.; Frick, T.M.

    1995-01-01

    A non-linear analysis has been performed to determine relative motions between tubes and tube support plates (TSP) during a steam line break (SLB) event for steam generators. The SLB event results in blowdown of steam and water out of the steam generator. The fluid blowdown generates pressure drops across the TSPS, resulting in out-of-plane motion. The SLB induced pressure loads are calculated with a computer program that uses a drift-flux modeling of the two-phase flow. In order to determine the relative tube/TSP motions, a nonlinear dynamic time-history analysis is performed using a structural model that considers all of the significant component members relative to the tube support system. The dynamic response of the structure to the pressure loads is calculated using a special purpose computer program. This program links the various substructures at common degrees of freedom into a combined mass and stiffness matrix. The program accounts for structural non-linearities, including potential tube and TSP interaction at any given tube position. The program also accounts for structural damping as part of the dynamic response. Incorporating all of the above effects, the equations of motion are solved to give TSP displacements at the reduced set of DOF. Using the displacement results from the dynamic analysis, plate stresses are then calculated using the detailed component models. Displacements form the dynamic analysis are imposed as boundary conditions at the DOF locations, and the finite element program then solves for the overall distorted geometry. Calculations are also performed to assure that assumptions regarding elastic response of the various structural members and support points are valid

  19. Performance and Economics of Catalytic Glow Plugs and Shields in Direct Injection Natural Gas Engines for the Next Generation Natural Gas Vehicle Program: Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Mello, J. P.; Bezaire, D.; Sriramulu, S.; Weber, R.

    2003-08-01

    Subcontractor report details work done by TIAX and Westport to test and perform cost analysis for catalytic glow plugs and shields for direct-injection natural gas engines for the Next Generation Natural Gas Vehicle Program.

  20. On the evaluation of lifetime of evaporative tubes of once-through steam generators at steam-generating surface temperature oscillations in the burnout region

    International Nuclear Information System (INIS)

    Vorob'ev, V.A.; Loshchinin, V.M.; Remizov, O.V.

    1978-01-01

    Suggested is a method for evaluation of a stressed state of evaporation tubes of once-through steam generators at temperature oscillations in the burnout region. Calculated is the amplitude of steam-generating surface temperature oscillations in the burnout region depending on the frequency of a liquid-steam boundary transfer and on this basis determined are thermal stresses in a tube wall. Knowing a fatigue curve gives the possibility to evaluate a heat transfer tube lifetime

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

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

    OpenAIRE

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

    2014-01-01

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

  3. Wear behavior of 2-1/4 Cr-1Mo tubing against alloy 718 tube-support material in sodium-cooled steam generators

    International Nuclear Information System (INIS)

    Wilson, W.L.

    1983-05-01

    A series of prototypic steam generator 2-1/4 Cr-1 Mo tube/alloy 718 tube support plate wear tests were conducted in direct support of the Westinghouse Nuclear Components Division -- Breeder Reactor Components Project Large Scale steam Generator design. The initial objective was to verify the acceptable wear behavior of softer, ''over-aged'' alloy 718 support plate material. For all interfaces under all test conditions, resultant wear damage was adhesive in nature with varying amounts of 2-1/4 Cr-1 Mo tube material being adhesively transferred to the alloy 718 tube supports. Maximum tube wear depths exceeded the initially established design allowable limit of 127 μm (.005 in.) at 17 of the 18 interfaces tested. A decrease in contact stresses produced acceptable tube wear depths below a readjusted maximum design allowable value of 381 μm (.015 in.). Additional conservatisms associated with the simulation of a 40-year lifetime of rubbing in a one-week laboratory test provided further confidence that the 381 μm maximum tube wear allowance would not be exceeded in service. Softer, ''over-aged'' alloy 718 material was found to produce slightly less wear damage on 2-1/4 Cr-1 Mo tubing than fully age hardened material. Also, air formed oxide films on the alloy 718 reduced initial tube wear and delayed the onset of adhesive surface damage. However, at high surface stress levels, these films were not sufficiently stable to provide adequate long term protection from adhesive wear. The results of the present work and those of previous test programs suggest that the successful in-sodium tribological performance of 2-1/4 Cr-1 Mo/alloy 718 rubbing couples is dependent upon the presence of lubricative surface films, such as oxides and/or surface reaction or deposition products. 11 refs., 13 figs., 4 tabs

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

    International Nuclear Information System (INIS)

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

    1991-11-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lorenz, H; Herberg, G

    1975-07-01

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

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  8. Assessment of the integrity of degraded steam generator tube by the use of heterogeneous finite element method

    International Nuclear Information System (INIS)

    Duan, X.; Kozluk, M.; Pagan, S.; Mills, B.

    2006-01-01

    Steam generator tubes at Ontario Power Generation (OPG) have been experiencing a variety of degradations such as pitting, fretting wear, erosion-corrosion, thinning and denting. To assist with steam generator life cycle management, OPG has developed Fitness-For-Service Guidelines (FFSG) for steam generator tubes. The FFSG are intended to provide standard acceptance criteria and evaluation procedures for assessing the condition of steam generator tubes for structural integrity, operational leak rate, and consequential leakage during an upset or abnormal event. Based on inspection results in conjunction with representative, postulated distributions of flaws in the un-inspected tubes, the FFSG provide an acceptable method of satisfying the intent of CSA-N285.4 and justifying the continued operation of degraded steam generator tubes. Some non-mandatory empirical axial and circumferential flaw models are also provided in the FFSG for structural integrity assessments. The test data from the OPG Steam Generator Tube Test Program (SGTTP) showed that the FFSG axial flaw model is conservative for a wide range of defect morphologies. A defect-specific axial flaw model was proposed for lattice-bar fret defects in I800 tubes by utilizing the SGTTP database of extensive test results. A defect-specific flaw model for outer diameter (OD) pitting and inner diameter (ID) intergranular attack in Monel 400 tubes was also developed using the SGTTP test data. More tests have been scheduled to support the development of defect specific models for axial flaws (OD cracks or ID laps) in Monel 400 and to supplement the database for Monel 400 pits. This paper explores the use of simulated testing for use in developing defect specific flaw models to reduce the amount of expensive tests. The Heterogeneous Finite Element Model (HFEM) has been developed and successfully applied to predict the failure behaviour of ductile metals under various deformation modes, i.e. plane stress, plane strain and

  9. Demineralized water flow cancelling experiments with ice plug into high diameter horizontal tube (300 nominal diameter)

    International Nuclear Information System (INIS)

    Gyongyosi, T.; Valeca, S.; Corbescu, B.; Puiu, D.; Panaitescu, V. N.

    2015-01-01

    The isolation with ice plug of a high diameter horizontal pipeline section is a specific technique for repairs activities/ replacements of components owning to thermo-hydraulic installations working with liquid agents. The application of such technique don.t assumes stopping of the entire system. The ice plugging inside of the pipeline assumes using of a special device and of an own specific technology for application. The paper contains a brief description of the experimental technological facilities used, followed by setting off the main moments in evolution of two experimental tests carried out on the test section with 300 mm nominal diameter for demineralized water and, finally, by a brief results analysis and some conclusions. The paper is dedicated to the specialists working in the research and technological engineering. (authors)

  10. Repair technique for steam generator tubes using electroforming

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  11. Repair technique for steam generator tubes using electroforming

    International Nuclear Information System (INIS)

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

    2001-07-01

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

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

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

    International Nuclear Information System (INIS)

    Cui Zhenghua; Jia Dounan; Chen Xuejun; Yu Erjun

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  15. Mode Selection for Axial Flaw Detection in Steam Generator Tube Using Ultrasonic Guided Wave

    International Nuclear Information System (INIS)

    Yoon, Byung Sik; Yang, Seung Han; Guon, Ki Il; Kim, Yong Sik

    2009-01-01

    The eddy current testing method is mainly used to inspect steam generator tube during in-service inspection period. But the general problem of assessing the structural integrity of the steam generator tube using eddy current inspection is rather complex due to the presence of noise and interference signal under various conditions. However, ultrasonic testing as a nondestructive testing tool has become quite popular and effective for the flaw detection and material characterization. Currently, ultrasonic guided wave is emerging technique in power industry because of its various merits. But most of previous studies are focused on detection of circumferential oriented flaws. In this study, the steam generator tube of nuclear power plant was selected to detect axially oriented flaws and investigate guided wave mode identification. The longitudinal wave mode is generated using piezoelectric transducer frequency from 0.5 MHz, 1.0 MHz, 2.25MHz and 5MHz. Dispersion based STFT algorithm is used as mode identification tool

  16. Aspergillus fumigatus colonization of punctal plugs.

    Science.gov (United States)

    Tabbara, Khalid F

    2007-01-01

    Punctal plugs are used in patients with dry eye syndrome to preserve the tears. In this report, I present two cases of Aspergillus fumigatus colonization of punctal plugs. Observational series of two cases. Approval was obtained from the institutional review board. Two men aged 29 and 31 years developed black spots inside the hole of punctal plug, which looked like eyeliner deposits. The deposits inside the hole of the plug in each patient were removed and cultured. Cultures of the two punctal plugs black deposits grew A fumigatus. Bacterial cultures were negative. Colonization of the punctal plug hole with A fumigatus was observed in two cases. It is recommended that punctal plugs be removed in patients undergoing refractive or intraocular procedures or in patients who are receiving topical corticosteroids. Current punctal plugs should be redesigned to avoid the presence of an inserter hole.

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

    International Nuclear Information System (INIS)

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

    1978-01-01

    Means and method are described for energizing and regulating a neutron generator tube having a target, an ion source and a replenisher. It providing a negative high voltage to the target and monitoring the target current. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordence with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current

  18. Steam generator group project: Task 13 final report: Nondestructive examination validation

    International Nuclear Information System (INIS)

    Bradley, E.R.; Doctor, P.G.; Ferris, R.H.; Buchanan, J.A.

    1988-08-01

    The Steam Generator Group Project (SGGP) was a multi-task effort using the retired-from-service Surry 2A pressurized water reactor steam generator as a test bed to investigate the reliability and effectiveness of in-service nondestructive eddy current (EC) inspection equipment and procedures. The information developed provided the technical basis for recommendations for improved in- service inspection and tube plugging criteria of steam generators. This report describes the results and analysis from Task 13--NDE Validation. The primary objective of Task 13 was to validate the EC inspection to detect and size tube defects. Additional objectives were to assess the nature and severity of tube degradation from all regions of the generator and to measure the remaining integrity of degraded specimens by burst testing. More than 550 specimens were removed from the generator and included in the validation studies. The bases for selecting the specimens and the methods and procedures used for specimen removal from the generator are reported. Results from metallurgical examinations of these specimens are presented and discussed. These examinations include visual inspection of all specimens to locate and identify tube degradation, metallographic examination of selected specimens to establish defect severity and burst testing of selected specimens to establish the remaining integrity of service-degraded tubes. Statistical analysis of the combined metallurgical and EC data to determine the probability of detection (POD) and sizing accuracy are reported along with a discussion of the factors which influenced the EC results. Finally, listings of the metallurgical and corresponding EC data bases are given. 12 refs., 141 figs., 24 tabs

  19. 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

  20. Borehole plugging experiment in OL-KR24 at Olkiluoto, Finland

    International Nuclear Information System (INIS)

    Rautio, T.

    2006-04-01

    Sealing of investigation boreholes has been studied by Svensk Koernbroenslehantering AB (SKB) and Posiva Oy (Posiva) as a part of final disposal research. The proposed principle is that investigation boreholes drilled at a site must not act as a continuous flow path for groundwater but be sealed to become as tight as the surrounding rock. As a part of the investigations SKB and Posiva started the third phase of the joint project 'Cleaning and sealing of investigation boreholes' in 2005. One of the sub-projects was the plugging experiment in borehole OL-KR24 at Olkiluoto. The aim of the experiment was to test all main procedures of borehole sealing concept in practise in a deep borehole. Borehole KR24 was drilled to the depth of 551.11 metres in 2003 and it was located inside the shaft profile in Onkalo. From the surface to the depth of about 120 m the borehole diameter is 98 mm. The rest of the borehole is 75.7 mm in diameter. The borehole is vertical and the inclination is quite accurately 90 degrees. The plugging experiment in borehole OL-KR24 consisted of four main activities: (1) cleaning of the borehole, (2) characterization of the borehole (3) selective stabilisation of the borehole, and (4) emplacement of plugs. The comprehensive cleaning of the borehole was to be done in the first stage to provide the basis for other activities. The aim of characterization was to study the borehole in order to determine the sections for selective stabilisation and the locations for plugs. The characterization phase consisted of caliper measurements, dummy probing and optical borehole imaging (OBI). The aim of selective stabilisation was to show that selected borehole sections can be stabilised using new techniques and methods. One borehole section was reamed from Ω 76 mm to Ω 98 mm. The reamed borehole section should have been filled with sufficient amount of cement-based material to achieve a stable 'concrete tube' after redrilling, but due to encountered problems and

  1. Non-destructive evaluation of stream generator tubes and pressure tubes from the PHWR reactors, using the rotating magnetic field method

    International Nuclear Information System (INIS)

    Premel, D.; Placko, D.; Grimberg, R.; Savin, A.

    2001-01-01

    This work presents a new type of eddy current transducer with a rotating magnetic field devoted to the inspection of steam generator tubes and pressure tubes from the PHWR reactors. A theoretical model has been developed that permits the calculations of the emf induced in the reception coils in the presence of the copper or magnetite deposits, anti-vibration railing and garter springs. (authors)

  2. Steam generator issues in the United States

    International Nuclear Information System (INIS)

    Strosnider, J.R.

    1997-01-01

    Alloy 600 steam generator tubes in the US have exhibited degradation mechanisms similar to those observed in other countries. Effective programs have been implemented to address several degradation mechanisms including: wastage; mechanical wear; pitting; and fatigue. These degradation mechanisms are fairly well understood as indicated by the ability to effectively mitigate/manage them. Stress corrosion cracking (SCC) is the dominant degradation mechanism in the US. SCC poses significant inspection and management challenges to the industry and the regulators. The paper also addresses issues of research into SCC, inspection programs, plugging, repair strategies, water chemistry, and regulatory control. Emerging issues in the US include: parent tube cracking at sleeve joints; detection and repair of circumferential cracks; free span cracking; inspection and cracking of dented regions; and severe accident analysis

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

    International Nuclear Information System (INIS)

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

    2009-11-01

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

  4. 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.)

  5. Experiences from the design and construction of plug II in the Prototype Repository

    Energy Technology Data Exchange (ETDEWEB)

    Dahlstroem, Lars-Olof (NCC Teknik (Sweden))

    2009-12-15

    factory and was ready for installation when arrival on site. Parallel to the installation of the reinforcement bars, the cooling system was installed together with measurement gauges and cables. The plug was casted with Self Compacting Concrete (SCC) mixed at the factory and continuously delivered on site. Mixing and Casting SCC is more complex than standard concrete and requires experienced and accurate personnel and a well developed control program. It is important to use correct procedures and well known ingredients when mixing SCC. After a controlled curing, the plug was cooled down to a temperature about 10 degrees below the ambient temperature in order to facilitate better penetration and filling of grout during contact grouting process. Contact grouting is performed through pre-installed grouting tubes. Mechanical measurements have been performed in order to monitor and verify the function and the behaviour of the plug. Stresses and deformations are not perfectly correlated between the calculated and the measured. However, the measurements show that the plug behaves as expected and that the entire plug is compressed, which was the purpose of its design

  6. Experiences from the design and construction of plug II in the Prototype Repository

    International Nuclear Information System (INIS)

    Dahlstroem, Lars-Olof

    2009-12-01

    installation when arrival on site. Parallel to the installation of the reinforcement bars, the cooling system was installed together with measurement gauges and cables. The plug was casted with Self Compacting Concrete (SCC) mixed at the factory and continuously delivered on site. Mixing and Casting SCC is more complex than standard concrete and requires experienced and accurate personnel and a well developed control program. It is important to use correct procedures and well known ingredients when mixing SCC. After a controlled curing, the plug was cooled down to a temperature about 10 degrees below the ambient temperature in order to facilitate better penetration and filling of grout during contact grouting process. Contact grouting is performed through pre-installed grouting tubes. Mechanical measurements have been performed in order to monitor and verify the function and the behaviour of the plug. Stresses and deformations are not perfectly correlated between the calculated and the measured. However, the measurements show that the plug behaves as expected and that the entire plug is compressed, which was the purpose of its design

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

    International Nuclear Information System (INIS)

    Hare, M.G.

    1976-11-01

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

  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. 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

  10. 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

  11. 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)

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

    International Nuclear Information System (INIS)

    1977-01-01

    A means and method for energizing and regulating a neutron generator tube is described. It has a target, an ion source and a replenisher. A negative high voltage is applied to the target and the target current monitored. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordance with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current. (C.F.)

  13. 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.

  14. 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

  15. Utilization of a sealed-tube neutron generator for training and research

    International Nuclear Information System (INIS)

    Jonah, S.A.

    2000-01-01

    The development of a program in nuclear science and technology in Nigeria began in 1976 with the establishment of two research centers, namely, the Centre for Energy Research and Training, (CERT), Zaria and the Centre for Energy Research and Development (CERD), Ile-Ife. The choice of Neutron Activation Analysis (NAA) technique as a very effective method of training scientists in basic and applied nuclear research led to the purchase of two KAMAN A-711 Neutron Generators for the two research centers. At CERT, the neutron generator (code named ZARABUNG-1) was successfully installed and the first 14 MeV neutrons were produced through the technical assistance of the International Atomic Energy Agency (IAEA) in 1988. In 1991, a new tube-head was purchased and installed due to the expiration of the old tube. Following the completion of its permanent site, the neutron generator was re-located from the old site and re-installed at the permanent site of CERT in 1995. (author)

  16. Neutron generator tube ion source control apparatus

    International Nuclear Information System (INIS)

    Bridges, J.R.

    1982-01-01

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

  17. In vitro sealing of iatrogenic fetal membrane defects by a collagen plug imbued with fibrinogen and plasma.

    Science.gov (United States)

    Engels, A C; Hoylaerts, M F; Endo, M; Loyen, S; Verbist, G; Manodoro, S; DeKoninck, P; Richter, J; Deprest, J A

    2013-02-01

    We aimed to demonstrate local thrombin generation by fetal membranes, as well as its ability to generate fibrin from fibrinogen concentrate. Furthermore, we aimed to investigate the efficacy of collagen plugs, soaked with plasma and fibrinogen, to seal iatrogenic fetal membrane defects. Thrombin generation by homogenized fetal membranes was measured by calibrated automated thrombography. To identify the coagulation caused by an iatrogenic membrane defect, we analyzed fibrin formation by optical densitometry, upon various concentrations of fibrinogen. The ability of a collagen plug soaked with fibrinogen and plasma was tested in an ex vivo model for its ability to seal an iatrogenic fetal membrane defect. Fetal membrane homogenates potently induced thrombin generation in amniotic fluid and diluted plasma. Upon the addition of fibrinogen concentrate, potent fibrin formation was triggered. Measured by densiometry, fibrin formation was optimal at 1250 µg/mL fibrinogen in combination with 4% plasma. A collagen plug soaked with fibrinogen and plasma sealed an iatrogenic membrane defect about 35% better than collagen plugs without these additives (P = 0.037). These in vitro experiments suggest that the addition of fibrinogen and plasma may enhance the sealing efficacy of collagen plugs in closing iatrogenic fetal membrane defects. © 2013 John Wiley & Sons, Ltd.

  18. Ultrasonic wall thickness gauging for ferritic steam generator tubing as an in-service inspection tool

    International Nuclear Information System (INIS)

    Haesen, W.M.J.; Tromp, Th.J.

    1980-01-01

    In-service inspection of LWR steam generators is more or less a standard routine operation. The situation can be very different for LMFBRs. For the SNR 300 (Kalkar Power Station) the situation is different because the steam generators have ferritic tubing. The tube walls are comparatively thick, 2 to 4.5 mm. During inservice examinations the steam generators will be drained on both sides, however on the sodium side a sodium film will be present. Furthermore the SNR 300 will have two types of steam generator. A straight tube design and a helical coil design will be used. Both types consist of a evaporator and superheater. The steam generators are of course not radioactive. It is obvious that in this case the eddy current (EC) technique is not an enviable inservice inspection tool. Basically EC is a surface flaw detection technique. Only the saturation magnetisation method will improve the EC technique sufficiently for ferritic material. However the 'in bore examination' with the saturation technique was, in case of the SNR 300 steam generator tubing, considered impossible since the inner diameters are fairly small. Furthermore sodium traces may influence the EC method. Although multifrequency methods can solve this problem, EC is not considered as a useful tool for examining ferritic tubing. Another method is to employ the 'stray flux' method which is under development with the TNO organization in Holland. The EC and stray flux method do have one drawback, these methods do not detect gradual changes in wall thickness. Ultrasonic examinations will be used in the SNR 300 as the main inspection tool for the steam generators. In this paper the reasons why ultrasonic examination was selected are explained. The results of the development work on this subject are discussed

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

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  20. Support vector regression model based predictive control of water level of U-tube steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Kavaklioglu, Kadir, E-mail: kadir.kavaklioglu@pau.edu.tr

    2014-10-15

    Highlights: • Water level of U-tube steam generators was controlled in a model predictive fashion. • Models for steam generator water level were built using support vector regression. • Cost function minimization for future optimal controls was performed by using the steepest descent method. • The results indicated the feasibility of the proposed method. - Abstract: A predictive control algorithm using support vector regression based models was proposed for controlling the water level of U-tube steam generators of pressurized water reactors. Steam generator data were obtained using a transfer function model of U-tube steam generators. Support vector regression based models were built using a time series type model structure for five different operating powers. Feedwater flow controls were calculated by minimizing a cost function that includes the level error, the feedwater change and the mismatch between feedwater and steam flow rates. Proposed algorithm was applied for a scenario consisting of a level setpoint change and a steam flow disturbance. The results showed that steam generator level can be controlled at all powers effectively by the proposed method.

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-02-01

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

  3. Design and development of face seal type sealing plug for advanced heavy water reactor

    International Nuclear Information System (INIS)

    Bansal, S.; Bhattacharyya, S.; Patel, R.J.; Agrawal, R.G.; Vaze, K.K.

    2005-09-01

    Advanced Heavy Water Reactor is a vertical pressure tube type reactor having light water as its coolant and heavy water as moderator. Sealing plug is required to close the pressure boundary of main heat transport system of the reactor by preventing escape of light water/steam From the coolant channel. There are 452 coolant channels in the reactor located in square lattice pitch. Sealing plug is located at the top of each coolant channel (in the top end fitting). Top end fitting is having a stepped bore to create a sealing face. Sealing plug is held through its expanded jaws in a specially provided groove of the end fitting. The plug was designed and prototypes were manufactured considering its functional importance, intricate design and precision machining requirements. Sealing plug consists of about 20 components mostly made up of precipitation hardening stainless steel, which is suitable for water environment and meets other requirements of strength and resistance to wear and galling. Seal disc is a critical component of the sealing plug as it is the pressure-retaining component. It is a circular disc with protruded stem. One face of the seal disc is nickel plated in the peripheral area that creates the sealing by abutting against the sealing face provided in the end fitting. The typical shape and profile of seal disc provides flexibility and allows elastic deformation to assist in locking of sealing plug and creating adequate seating force for effective sealing. Design and development aspects of the sealing plug have been detailed out in this report. Also results of stress analysis and experimental studies for seal disc have been mentioned in the report. Stress analysis and experimental testing was required for the seal disc because high stresses are developed due to its exposure to high pressure and temperature environment of Main Heat Transport system. Hot testing was carried out to simulate the reactor-simulated condition. The performance was found to be

  4. Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes; Revised September 3, 2003

    International Nuclear Information System (INIS)

    Rochau, Gary E.; Caffey, Thurlow W.H.; Bahram Nassersharif; Garcia, Gabe V.; Jedlicka, Russell P.

    2003-01-01

    OAK B204 Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003. 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 was researched for detection of defects using a continuous-wave radar method within metal tubing. The technique is 100% volumetric, and may find smaller defects, more rapidly, and less expensively than present methods. The project described in this report was a joint development effort between Sandia National Laboratories (SNL) and New Mexico State University (NMSU) funded by the US Department of Energy. The goal of the project was to research, design, and develop a new concept utilizing a continuous wave radar to detect defects inside metallic tubes and in particular nuclear plant steam generator tubing. The project was divided into four parallel tracks: computational modeling, experimental prototyping, thermo-mechanical design, and signal detection and analysis

  5. Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003

    Energy Technology Data Exchange (ETDEWEB)

    Gary E. Rochau and Thurlow W.H. Caffey, Sandia National Laboratories, Albuquerque, NM 87185-0740; Bahram Nassersharif and Gabe V. Garcia, Department of Mechanical Engineering, New Mexico State University, Las Cruces, NM 88003-8001; Russell P. Jedlicka, Klipsch School of Electrical and Computer Engineering, New Mexico State University, Las Cruces, NM 88003-8001

    2003-05-01

    OAK B204 Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003. 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 was researched for detection of defects using a continuous-wave radar method within metal tubing. The technique is 100% volumetric, and may find smaller defects, more rapidly, and less expensively than present methods. The project described in this report was a joint development effort between Sandia National Laboratories (SNL) and New Mexico State University (NMSU) funded by the US Department of Energy. The goal of the project was to research, design, and develop a new concept utilizing a continuous wave radar to detect defects inside metallic tubes and in particular nuclear plant steam generator tubing. The project was divided into four parallel tracks: computational modeling, experimental prototyping, thermo-mechanical design, and signal detection and analysis.

  6. Proceedings of steam generator sludge deposition in recirculating and once through steam generator upper tube bundle and support plates

    International Nuclear Information System (INIS)

    Baker, R.L.; Harvego, E.A.

    1992-01-01

    The development of remedial measures of shot peening have given nuclear utilities viable measures to address primary water stress corrosion cracking to extend steam generator life. The nuclear utility industry is now faced with potential replacement of steam generators in nuclear power plants due to stress corrosion cracking and intergranular attach in crevice locations on the secondary side of steam generators at tube support plates and at the crevice at the top of the tube sheet. Significant work has been done on developing and understanding of the effects of sludge buildup on the corrosion process at these locations. This session was envisioned to provide a forum for the development of an understanding of the mechanisms which control the transport and deposition of sludge on the secondary side of steam generators. It is hoped that this information will aid utilities in monitoring the progression of fouling of these crevices by further knowledge in where to look for the onset of support plate crevice fouling. An understanding of the progression of fouling from upper tube support plates to those lower in the steam generator where higher temperatures cause the corrosion process to initiate first can aid the nuclear utility industry in developing remedial measures for this condition and in providing a forewarning of when to apply such remedial measures

  7. ENVIRONMENTAL TECHNOLOGY VERIFICATION REPORT: RESIDENTIAL ELECTRIC POWER GENERATION USING THE PLUG POWER SU1 FUEL CELL SYSTEM

    Science.gov (United States)

    The Environmental Technology Verification report discusses the technology and performance of the Plug Power SU1 Fuel Cell System manufactured by Plug Power. The SU1 is a proton exchange membrane fuel cell that requires hydrogen (H2) as fuel. H2 is generally not available, so the ...

  8. Preliminary Plugging tests in Narrow Sodium Channels by Sodium and Carbon Dioxide reaction

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sun Hee; Wi, Myung-Hwan; Min, Jae Hong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    This report is on the investigation of the physical/chemical phenomena that a slow loss of CO{sub 2} inventory into sodium after the sodium-CO{sub 2} boundary failure in PCHEs in realistic operating conditions. The first phenomenon is potential channel plugging inside the narrow PCHE channel. Unlike a conventional shell and- tube type HXs, failures in a PCHE are expected to be small cracks. If the faulted channel is blocked, it may have a positive function for plant safety because the pressure boundary would automatically recover due to this self-plugging. The other one is damage propagation on pressure boundary, which is referred to as potential wastage with combined corrosion/erosion effect. Physical/chemical phenomena that a slow loss of CO{sub 2} inventory into sodium after the sodium-CO{sub 2} boundary failure in printed circuit heat exchangers (PCHEs) were investigated. Our preliminary experimental results of plugging show that sodium flow immediately stopped as CO{sub 2} was injected through the nozzle at 300-400 .deg. C in 3 mm sodium channels, whereas sodium flow stopped about 60 min after CO{sub 2} injection in 5 mm sodium channels.

  9. Sealing efficiency of an argillite-bentonite plug subjected to gas pressure, in the context of deep underground nuclear waste storage

    International Nuclear Information System (INIS)

    Liu, Jiang-Feng

    2013-01-01

    In France, the deep underground nuclear waste repository consists of a natural barrier (in an argillaceous rock named argillite), associated to artificial barriers, including plugs of swelling clay (bentonite)-sand for tunnel sealing purposes. The main objective of this thesis is to assess the sealing efficiency of the bentonite-sand plug in contact with argillite, in presence of both water and gas pressures. To assess the sealing ability of partially water-saturated bentonite/sand plugs, their gas permeability is measured under varying confining pressure (up to 12 MPa). It is observed that tightness to gas is achieved under confinement greater than 9 MPa for saturation levels of at least 86-91%. We than assess the sealing efficiency of the bentonite-sand plug placed in a tube of argillite or of Plexiglas-aluminium (with a smooth or a rough interface). The presence of pressurized gas affects the effective swelling pressure at values P gas from 4 MPa. Continuous gas breakthrough of fully water-saturated bentonite-sand plugs is obtained for gas pressures on the order of full swelling pressure (7-8 MPa), whenever the plug is applied along a smooth interface. Whenever a rough interface is used in contact with the bentonite-sand plug, a gas pressure significantly greater than its swelling pressure is needed for gas to pass continuously. Gas breakthrough tests show that the interface between plug/argillite or the argillite itself are two preferential pathways for gas migration, when the assembly is fully saturated. (author)

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

    International Nuclear Information System (INIS)

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

    1999-12-01

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

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

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  12. Anti vibration bars replacement in Vandellos II steam generators

    International Nuclear Information System (INIS)

    Vinyes, R.; Leal, R.

    1994-01-01

    C.N. Vandellos II is equipped with three steam generators Westinghouse model F. The number of tubes is 5626 each SG and the material Inconel 600TT. During the first inservice inspection, in 1989, tube wall thickness reductions were observed due to fretting in zones of contact with the tubes anti vibration bars. In the 2 nd shutdown for refueling (1990) all the tubes subject to this type of degradation were inspected by eddy currents, occurring a significative increase in number of tubes affected as well as the quantity of plugged tubes for that reason. Additionally, Westinghouse performed visual inspection and dimensional control of gaps in the tube bundles. Taking in account the results, the replacement with AVBs of new design was decided. AVBs new design is more complex than the original due to the combination of flexible and expandable bars in order to eliminate gaps between tubes an bars an assure proper bundle support. Given that the installation has to be done under water for shielding, all unions are bolted so that no welding is required. Each one of the bars, 333 per SG, is attached to a support structure consisting in 6 retaining plates and 4 bridge plates. (Author)

  13. 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

  14. Impact of plug-in electric vehicles on voltage unbalance in ...

    African Journals Online (AJOL)

    Plug-in electric vehicle (PEV) will soon be connected to residential distribution networks. ... generation units which transfer the energy stored in their battery into grid. ... electric vehicles on voltage imbalance in distribution system is presented. ... and other types of distribution generator such as solar photovoltaic and wind ...

  15. 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

  16. 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

  17. Numerical investigation of heat transfer and entropy generation of laminar flow in helical tubes with various cross sections

    International Nuclear Information System (INIS)

    Kurnia, Jundika C.; Sasmito, Agus P.; Shamim, Tariq; Mujumdar, Arun S.

    2016-01-01

    Highlights: • Heat transfers of helical coiled tube with several cross section profiles are evaluated. • Helical tubes offer higher heat transfer and lower entropy generation. • Square cross-section generates the highest entropy, followed by ellipse and circular. • Study could serve as a guideline in designing an efficient helical tube heat exchanger. - Abstract: This study evaluates heat transfer performance and entropy generation of laminar flow in coiled tubes with various cross-sections geometries i.e. circular, ellipse and square, relatives to the straight tubes of similar cross-sections. A computational fluid dynamics model is developed and validated against empirical correlations. Good agreement is obtained within range of Reynolds and Dean numbers considered. Effect of geometry, wall temperature, Reynolds number and heating/cooling mode were examined. To evaluate the heat transfer performance of the coiled tube configurations, a parameter referred as Figure of Merit (FoM) is defined as the ratio heat transfer rate to the required pumping power. In addition, exergy analysis is carried out to examine the inefficiency of the coiled tube configurations. The results indicate that coiled tubes provide higher heat transfer rate. In addition, it was found to be more efficient as reflected by lower entropy generation as compared to straight tubes. Among the studied cross-section, square cross-section generates the highest entropy, followed by ellipse and circular counterpart. Entropy production from heat transfer contribution is two order-of-magnitude higher than that of entropy contribution from viscous dissipation. Cooling case produces slightly higher entropy than heating counterpart. Finally, this study can provide practical guideline to design more efficient coiled heat exchanger.

  18. Core barrel plug

    International Nuclear Information System (INIS)

    Tolino, R.W.; Hopkins, R.J.; Congleton, R.L.; Popalis, C.H.

    1986-01-01

    A plug is described for preventing flow through a port in a core barrel of a pressurized water nuclear reactor which consists of: a substantially cylindrical body formed with a cylindrical portion and a flange and defining a tapered leading open end with the other end being closed by an end plug attached to the flange, the body defining a bore therein extending from the open end to the end plug with the bore having a smaller diameter near the open end than near the end plug, the cylindrical portion having a lip near the open end and being formed with longitudinal slots extending from the open end toward the flange and extending entirely through the thickness of the cylindrical portion, the cylindrical portion having a circumferential first groove on the outer surface thereof located near the forwardmost portion of the cylindrical portion but not in the section of the cylindrical portion that has the slots therein, and a plurality of circumferential second grooves on the outer surface thereof located in the section of the cylindrical portion that has the slots therein, the first and second grooves establishing a seal between the cylindrical portion and the inside surface of the port when the cylindrical portion is expanded, and the flange and the end plug having a passageway defined therein for introducing a fluid into the body; a metal ring disposed in each of the second grooves; a mandrel slidably disposed and captured in the body and capable of being moved toward the open end of the body when the fluid is introduced through the passageway, thereby causing the cylindrical portion to be expanded into contact with the inside surface of the port; and a locking mechanism disposed in the end plug for preventing inadvertent movement of the mandrel

  19. 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

  20. Contribution of statistic to the industrial problems of maintenance at EDF; Apport de la statistique a la problematique industrielle de la maintenance chez EDF

    Energy Technology Data Exchange (ETDEWEB)

    Cagnac, A.; Hervouet, C.; Bonnevialle, A.M.; Remy, E. [Electricite de France (EDF-RD), 78 - Chatou (France); Gay, N. [Electricite de France (EDF DPN), 93 - Saint-Denis (France)

    2006-07-01

    EDF/R and D was requested to take part in a program launched by EDF/DPN. The objective is to deal with the ASN request (Nuclear Safety Authorities), aiming at evaluating the feasibility of zero-leakage of the Steam Generators tube bundle during Primary Hydraulic Test (PHT) which will be carried out within the next decennial visits. A way to satisfy this objective is to plug the SG's tubes that can potentially contribute to the leakage. However, it is necessary to choose carefully the SG tubes to plug, in order to both, reduce significantly the SG's leak-flow during the test and limit the number of plugged tubes, which would imply an anticipation of the steam generator replacement. Within this framework, EDF/R and D/MRI carried out a preliminary statistical analysis from the feedback experience of the hydraulic tests carried out on units at the time of their first and second decennial visits. The goal was to identify, based on the control databases, the characteristics of the most contributing tubes to the leakage. The model obtained makes it possible to predict whether a steam generator will be leaky or not at the time of the hydraulic test, according to its state of cracking. The model provides a probability of considering the steam generator as leaky according to the sum of crack-lengths higher than a given threshold. This model was practically applied for the first time in 2005 in order to optimise the choice of the tubes to plug before the primary hydraulic test of the steam generators of a nuclear unit. Due to this optimised plugging, the leak-flows in PHT were sufficiently low and ensured the provisional life duration of these steam generators. (authors)

  1. GRUVAL for ET inspection of the steam generator tubes

    International Nuclear Information System (INIS)

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

    2013-01-01

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

  2. Estimation of a tube diameter in a ‘church window’ condenser based on entropy generation minimization

    Directory of Open Access Journals (Sweden)

    Laskowski Rafał

    2015-09-01

    Full Text Available The internal diameter of a tube in a ‘church window’ condenser was estimated using an entropy generation minimization approach. The adopted model took into account the entropy generation due to heat transfer and flow resistance from the cooling-water side. Calculations were performed considering two equations for the flow resistance coefficient for four different roughness values of a condenser tube. Following the analysis, the internal diameter of the tube was obtained in the range of 17.5 mm to 20 mm (the current internal diameter of the condenser tube is 22 mm. The calculated diameter depends on and is positively related to the roughness assumed in the model.

  3. Theoretical-experimental assessment of the variables affecting fretting of Atucha I nuclear power plant utility steam generators tubes

    International Nuclear Information System (INIS)

    Kulichevsky, Raul M.

    1995-01-01

    Fretting wear of Steam Generator tubes caused by flow induced vibrations generates uncertainty on their integrity. The knowledge of the controlling variables of the wear process may give a criterion to evaluate the tubes residual life. Information on vibratory response and dynamic interaction between tubes and their supports are prerequisites for understanding the relationship between fretting wear and tube vibration. Experimental results of the vibratory response of an Atucha-I nuclear power plant type U-tube, the influence of tube/support clearance on this response and a study of tube/support dynamic interaction, which allow the verification of a finite element model of this type of tubes, are presented in this work. Also wear results for the Incoloy 800/DIN 1.4550 austenitic stainless steel pair of materials and a first evaluation of the wear constant of this pair are presented. (author)

  4. Removable pipeline plug

    International Nuclear Information System (INIS)

    Vassalotti, M.; Anastasi, F.

    1984-01-01

    A removable plugging device for a pipeline, and particularly for pressure testing a steam pipeline in a boiling water reactor, wherein an inflatable annular sealing member seals off the pipeline and characterized by radially movable shoes for holding the plug in place, each shoe being pivotally mounted for self-adjusting engagement with even an out-of-round pipeline interior

  5. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    International Nuclear Information System (INIS)

    Diercks, D.R.

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

  6. Proceedings of the CNRA/CSNI workshop on steam generator tube integrity in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R. [Argonne National Lab., IL (United States)

    1997-02-01

    The objective of the workshop was to provide a working forum for the exchange of information by contributing experts on current issues related to PWR steam generator tube integrity. One hundred persons from 15 countries attended the workshop, including 36 from regulatory and nuclear policy agencies, 28 from research and development laboratories, 18 from nuclear vendors and consulting firms, and 18 from electrical utilities. The workshop opened with a plenary session; the first part of the session covered international steam generator regulatory practices and issues, featuring speakers from regulatory bodies in Belgium, France, Japan, Spain, and the US. In Part 2 of the plenary session, comprehensive technical overviews on steam generator tubing degradation, inspection, and integrity were presented by authorities in these fields from the US, France, and Belgium. Parallel working sessions on the second and third days of the workshop then developed findings and recommendations in the areas of (1) tubing degradation, (2) tubing inspection, (3) tubing integrity, (4) preventative and corrective measures, and (5) operational aspects and risk analysis. On the final day of the workshop, the working-session facilitators presented summaries of their sessions to the workshop attendees. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

  7. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

    1981-01-01

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

  10. 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

  11. The PISC programme on defective steam generator tubes inspection summary report

    International Nuclear Information System (INIS)

    Birac, C.; Comby, R.; Maciga, G.; Zanella, G.; Perez Prat, J.; Estorff, U. von

    1995-01-01

    The PISC III Actions are intended to extend the results and methodologies of the previous PISC exercises, i.e. the validation of the capabilities of the various examination techniques when used on real defects in real components under realistic conditions of inspection. The objective of this action is relatively close to that of the heavy structures programmes: the experimental evaluation of the performance of test procedures used for steam generator tubes in nuclear power plants during in-service or pre-service inspections. The exercise is a capability exercise consisting of Round Robin Tests on individual tubes including calibration, training and blind test tubes. In this paper the main conclusions from the RRT conducted in the framework of Action 5 will be presented and discussed. (author). 7 refs, 4 figs, 2 tabs

  12. Direct solar steam generation inside evacuated tube absorber

    Directory of Open Access Journals (Sweden)

    Khaled M. Bataineh

    2016-12-01

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

  13. Borehole Plugging Program. Plugging of ERDA No. 10 drill hole

    International Nuclear Information System (INIS)

    Gulick, C.W. Jr.

    1979-06-01

    A requirement exists to plug exploratory drill holes located in the proposed Waste Isolation Pilot Plant area of Southeastern New Mexico. Sandia Laboratories, in cooperation with the US Army Corps of Engineers, Waterways Experiment Station, Concrete Laboratory, developed pumpable and durable cement grouts. These grouts were successfully used to plug an existing drill hole in the area. Results of this project are presented, along with comments and conclusions

  14. 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

  15. 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

  16. Thermal-hydraulic tests of steam-generator tube-support-plate crevices. Volume 2. Appendixes I through S. Final report

    International Nuclear Information System (INIS)

    Cassell, D.S.; Vroom, D.W.

    1983-01-01

    A test program was conducted to determine for selected steam generator tube supports the thermal/hydraulic conditions at the inception of dryout as indicated by a tube wall temperature excursion, to determine the pressure drop across the supports, and to obtain photographic documentation of the flow upstream and downstream of the supports. A multi-tube steam generator model was used and testing performed over the range of typcal PWR steam generator operating conditions. These appendices contain information on instrumentation calibration, test model and loop calibration, error analysis, test model thermal-hydraulic analyses, index of lab materials and log sheets, index of two-phase flow still photographs, index of high speed movies and video, test data printouts, test model and loop fabrication drawings, procedure for silver brazing tubewall thermocouples, and procedure for esablishing tube-tube support line contact

  17. Mixture for plugging absorption zones

    Energy Technology Data Exchange (ETDEWEB)

    Sitinkov, G V; Kovalenko, N G; Makarov, L V; Zinnatulchin, Ts Kh

    1981-01-17

    A mixture is proposed for plugging absorption zones. The mixture contains synthetic polymer and a solvent. So as to increase the penetrability of the mixture through a reduction in its viscosity and an increase in insulation properties, the compound contains either Capron or Neilon as the synthetic polyamide resin polmyer, and concentrated chloride as the solvent. The mixture is prepared in a special AzINMASh-30 unit (acid cart). After the mixture has been produced, it is injected into the borehole by means of an acid cart pump. So as to prevent coaggulation at the point when the mixture in injected into the stratum through tubes, the mixture is placed betwen chemically inert fluids, for example, a clay mortar. The inert and compressed fluids are injected by means of a cementing unit. The entire process of production and application of the mixture is simple and fully automated through the use of well-known equipment.

  18. Methodology for the investigation of ignition near hot surfaces in a high-pressure shock tube

    Science.gov (United States)

    Niegemann, P.; Fikri, M.; Wlokas, I.; Röder, M.; Schulz, C.

    2018-05-01

    Autoignition of fuel/air mixtures is a determining process in internal combustion engines. Ignition can start either homogeneously in the gas phase after compression or in the vicinity of hot surfaces. While ignition properties of commercial fuels are conventionally described by a single quantity (octane number), it is known that some fuels have a varying propensity to the two processes. We present a new experimental concept that generates well-controlled temperature inhomogeneities in the shock-heated gases of a high-pressure shock tube. A shock-heated reactive mixture is brought into contact with a heated silicon nitride ceramic glow plug. The glow-plug temperature can be set up to 1200 K, higher than the post-reflected-shock gas temperatures (650-1050 K). High-repetition-rate chemiluminescence imaging is used to localize the onset of ignition in the vicinity of the hot surface. In experiments with ethanol, the results show that in most cases under shock-heated conditions, the ignition begins inhomogeneously in the vicinity of the glow plug and is favored because of the high wall temperature. Additionally, the interaction of geometry, external heating, and gas-dynamic effects was investigated by numerical simulations of the shock wave in a non-reactive flow.

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

    International Nuclear Information System (INIS)

    Garcia, Gabe V.

    2005-01-01

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

  20. New process for the relief of mechanically induced stresses in steam generator tubes

    International Nuclear Information System (INIS)

    Joyeux, J.P.

    1980-01-01

    Heat exchangers include very generally a set of tubes assembled in 'U-type' exchangers or in 'pass-through' exchangers. The tubes are introduced in holes drilled in the tube sheet plate, welded at their extremity and expanded to insure the necessary tightness. The steam generators built by FRAMATOME belong to the U-type and include, depending upon the nominal power of the plant, about three or five thousand inconel tubes. This material has been selected for its resistance to corrosion action at high temperatures. But one drawback of inconel is that residual stress lowers considerably this resistance to corrosion; so it is very important to apply manufacturing processes involving a residual stress level as low as possible. A new process, which involves 'kiss' rolling, is described. (author)

  1. 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

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

    International Nuclear Information System (INIS)

    Gaudin, J.P.

    1986-01-01

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

  3. Stagnations of increasing trends in negative pressure with repeated cavitation in water/metal Berthelot tubes as a result of mechanical sealing

    International Nuclear Information System (INIS)

    Hiro, Kazuki; Ohde, Yoshihito; Tanzawa, Yasutoshi

    2003-01-01

    To investigate effects of mechanical sealing on negative pressures in water/metal tube Berthelot systems, trends in negative pressure are observed through runs of temperature cycles below 90 deg. C in two systems made of metals having small amounts of gas inclusions. The first system is a pre-degassed all-stainless-steel tube/plug system. The steel is a special product for vacuum engineering. The second is the same tube sealed with plugs made of silver solidified one-dimensionally in a vacuum furnace. A new type of trend, stagnation for intermediate cycles is found in both systems so long as sealing distortion of each plug is small in amount. The stagnation period for the first system is longer than that for the second one. A metallurgical mechanism of a gas-being-replenished crevice model is proposed: distorted parts of metals undergo heat-treatment during runs of temperature cycles, and the heat-treatment enhances the rates of impurity gas transports to crevices on the metal surface where cavitation occurs, and the transport causes the stagnation for cycles during which the rates are still high

  4. Eddy-current tests on operational evaluation of steam generator tubes in nuclear power plants

    International Nuclear Information System (INIS)

    Lopez, Luiz Antonio Negro Martin; Ting, Daniel Kao Sun

    2000-01-01

    This paper presents a worldwide research on the technical and economical impacts due to failure in tube bundles of nuclear power plant steam generators. An Eddy current non destructive test using Foucault currents for the inspection and failure detection on the tubes, and also the main type of defects. The paper also presents the signals generated by a Zetec MIZ-40 test equipment. This paper also presents a brief description of an automatic system for data analysis which is under development by using a fuzzy logic and artificial intelligence

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

    International Nuclear Information System (INIS)

    1980-01-01

    A specification is given for an energizing and regulating circuit for a gas filled neutron generator tube consisting of a target, an ion source and a replenisher, the circuit consisting of a power supply to provide a negative high voltage to the target and a target current corresponding to the neutron output of the tube, a constant current source, and control means connected to the power supply and to the constant current source, the control means being responsive to the target current to provide a portion of the constant current to the replenisher substantially to regulate the neutron output of the tube. (author)

  6. Development of a novel miniature detonation-driven shock tube assembly that uses in situ generated oxyhydrogen mixture

    International Nuclear Information System (INIS)

    Janardhanraj, S.; Jagadeesh, G.

    2016-01-01

    A novel concept to generate miniature shockwaves in a safe, repeatable, and controllable manner in laboratory confinements using an in situ oxyhydrogen generator has been proposed and demonstrated. This method proves to be more advantageous than existing methods because there is flexibility to vary strength of the shockwave, there is no need for storage of high pressure gases, and there is minimal waste disposal. The required amount of oxyhydrogen mixture is generated using alkaline electrolysis that produces hydrogen and oxygen gases in stoichiometric quantity. The rate of oxyhydrogen mixture production for the newly designed oxyhydrogen generator is found to be around 8 ml/s experimentally. The oxyhydrogen generator is connected to the driver section of a specially designed 10 mm square miniature shock tube assembly. A numerical code that uses CANTERA software package is used to predict the properties of the driver gas in the miniature shock tube. This prediction along with the 1-D shock tube theory is used to calculate the properties of the generated shockwave and matches reasonably well with the experimentally obtained values for oxyhydrogen mixture fill pressures less than 2.5 bars. The miniature shock tube employs a modified tri-clover clamp assembly to facilitate quick changing of diaphragm and replaces the more cumbersome nut and bolt system of fastening components. The versatile nature of oxyhydrogen detonation-driven miniature shock tube opens up new horizons for shockwave-assisted interdisciplinary applications.

  7. Development of a novel miniature detonation-driven shock tube assembly that uses in situ generated oxyhydrogen mixture

    Energy Technology Data Exchange (ETDEWEB)

    Janardhanraj, S.; Jagadeesh, G., E-mail: jaggie@aero.iisc.ernet.in [Department of Aerospace Engineering, Indian Institute of Science, Bangalore 560012 (India)

    2016-08-15

    A novel concept to generate miniature shockwaves in a safe, repeatable, and controllable manner in laboratory confinements using an in situ oxyhydrogen generator has been proposed and demonstrated. This method proves to be more advantageous than existing methods because there is flexibility to vary strength of the shockwave, there is no need for storage of high pressure gases, and there is minimal waste disposal. The required amount of oxyhydrogen mixture is generated using alkaline electrolysis that produces hydrogen and oxygen gases in stoichiometric quantity. The rate of oxyhydrogen mixture production for the newly designed oxyhydrogen generator is found to be around 8 ml/s experimentally. The oxyhydrogen generator is connected to the driver section of a specially designed 10 mm square miniature shock tube assembly. A numerical code that uses CANTERA software package is used to predict the properties of the driver gas in the miniature shock tube. This prediction along with the 1-D shock tube theory is used to calculate the properties of the generated shockwave and matches reasonably well with the experimentally obtained values for oxyhydrogen mixture fill pressures less than 2.5 bars. The miniature shock tube employs a modified tri-clover clamp assembly to facilitate quick changing of diaphragm and replaces the more cumbersome nut and bolt system of fastening components. The versatile nature of oxyhydrogen detonation-driven miniature shock tube opens up new horizons for shockwave-assisted interdisciplinary applications.

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

    International Nuclear Information System (INIS)

    Pierson, E.; Stubbe, J.

    1993-01-01

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

  9. Tube-support response to tube-denting evaluation. Volume 1. Final report

    International Nuclear Information System (INIS)

    Anderson, P.L.; Hall, J.F.; Shah, P.K.; Wills, R.L.

    1983-05-01

    The response of the tube supports is one of the important considerations of tube denting in a steam generator. Investigations have indicated that damaged tube supports have the potential to distort and damage tubes. This investigation considers the response to tube denting of the Combustion Engineering type tube supports. Drilled support plates and eggcrate tube supports are tested in a model steam generator in which tube denting is induced. The experimental data is used to verify and refine analytical predictor models developed using finite element techniques. It was found that analytical models underpredicted the deformations of the tube supports and appropriate modifications to enhance the predictive capability are identified. Non-destructive examination methods are evaluated for application to operating steam generators. It was found that the standard eddy current and profilometry techniques are acceptable methods for determining tube deformations, but these techniques are not adequate to assess tube support damage. Radiography is judged to be the best available means of determining the extent and progression of damage in tube supports

  10. Probabilistic methodology for assessing steam generator tube inspection - Phase II: User's manual for CANTIA Version 1.1

    International Nuclear Information System (INIS)

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

    1997-03-01

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

  11. 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)

  12. 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).

  13. 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

  14. Operational control and maintenance integrity of typical and atypical coil tube steam generating systems

    Energy Technology Data Exchange (ETDEWEB)

    Beardwood, E.S.

    1999-07-01

    Coil tube steam generators are low water volume to boiler horsepower (bhp) rating, rapid steaming units which occupy substantially less space per boiler horsepower than equivalent conventional tire tube and water tube boilers. These units can be retrofitted into existing steam systems with relative ease and are more efficient than the generators they replace. During the early 1970's they became a popular choice for steam generation in commercial, institutional and light to medium industrial applications. Although these boiler designs do not require skilled or certified operators, an appreciation for a number of the operational conditions that result in lower unscheduled maintenance, increased reliability and availability cycles would be beneficial to facility owners, managers, and operators. Conditions which afford lower operating and maintenance costs will be discussed from a practical point of view. An overview of boiler design and operation is also included. Pitfalls are provided for operational and idle conditions. Water treatment application, as well as steam system operations not conducive to maintaining long term system integrity; with resolutions, will be addressed.

  15. Evaluation of a steam generator tube repair process using an explosive expansion techniuqe at TMI-1

    International Nuclear Information System (INIS)

    Rajan, J.; Shook, T.A.; Leonard, L.

    1983-01-01

    After a planned shutdown of Unit No. 1 at Three Mile Island, cracks were discovered in the primary side of steam generator tubes in the vicinity of the upper surface of the upper tubesheet. The nature of these cracks was later characterized as intergranular stress corrosion. The licensee, General Public Utilities Nuclear (GPUN), proposed to form a new tube-to-tubesheet seal below the cracks using a repair process wherein a detonating cord and polyethylene cartridge assembly inserted into the tube explosively expand the tube against the tubesheet. The explosive expansion process has had numerous applications over the years in the initial fabrication of heat exchanger tube-to-tubesheet assemblies and in repair processes using sleeving. However, this is the first use of this process in a steam generator to expand a previously rolled tube and to form a new seal between it and the tubesheet below a defective region in the tube. The seal obtained between the tube and tubesheet depends on the magnitude of explosive energy released in the detonating process. In this application, it is desired to obtain a mechanical bond rather than a metallurgical welding of the tube and tubesheet. A number of critical variables must be taken into account in order to obtain a successful mechanical seal. These include the explosive power of the detonating cord, the number of expansion shots used, the length of tube which is expanded, cartridge and tube diameters, the diameter of the tubesheet hole, the materials of the tube and tubesheet, and the condition of the surfaces at the time of repair. (orig./GL)

  16. CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, B.; Zhang, H.; Han, B.; Yang, B.W. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology; Mo, S.J.; Ren, H.B.; Qin, J.M.; Zuo, C.P. [China Nuclear Power Design Co. Ltd., ShenZhen (China)

    2016-07-15

    The integrity and thermal hydraulic characteristics of steam generator are of great concern in the nuclear industry. The tube support plates (TSP), one of the most important components of the steam generator, not only support the heat transfer tubes, but also affect the flow dynamic and thermal hydraulic characteristics of the secondary-side flow inside the steam generator. Different working conditions, ranging from single-phase adiabatic condition to two-phase high-void boiling condition, are simulated and analyzed. Calculated void fraction, under simple geometry, agrees well with the experiment data whilst the simulated heat transfer coefficient is tremendously close to the empirical correlation. Temperature, void fraction, and velocity distributions in different locations show reasonable distribution. The simulation results indicate that TSP can enhance the heat transfer in the secondary side of the steam generator. On the top of TSP, with the increase in cross-section flow area, the back-flow phenomenon occurs, which might lead to the contamination of precipitation.

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

    International Nuclear Information System (INIS)

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

    1993-01-01

    Various methods are being used to expand heat transfer tubes into the thick tubesheets of nuclear steam generators. The residual stresses in the as-expanded tubes and methods for reducing these stresses are important because of the role which residual stresses play in stress corrosion cracking and stress assisted corrosion of the tubing. Of the various expansion processes, the hydraulic expansion process is most amenable to analytical study. This paper presents results on the residual stresses and strains in hydraulically expanded tubes and the tubesheet as computed by two different finite element codes with three different finite element models and by a theoretical incremental analysis method. The calculations include a sensitivity analysis to assess the effects of the expansion variables and the effect of stress relief heat treatments. (orig.)

  18. Preparation and Characterization of Chemical Plugs Based on Selected Hanford Waste Simulants

    International Nuclear Information System (INIS)

    Mattigod, Shas V.; Wellman, Dawn M.; Parker, Kent E.; Cordova, Elsa A.; Gunderson, Katie M.; Baum, Steven R.; Crum, Jarrod V.; Poloski, Adam P.

    2008-01-01

    This report presents the results of preparation and characterization of chemical plugs based on selected Hanford Site waste simulants. Included are the results of chemical plug bench testing conducted in support of the M1/M6 Flow Loop Chemical Plugging/Unplugging Test (TP-RPP-WTP-495 Rev A). These results support the proposed plug simulants for the chemical plugging/ unplugging tests. Based on the available simulant data, a set of simulants was identified that would likely result in chemical plugs. The three types of chemical plugs that were generated and tested in this task consisted of: 1. Aluminum hydroxide (NAH), 2. Sodium aluminosilicate (NAS), and 3. Sodium aluminum phosphate (NAP). While both solvents, namely 2 molar (2 M) nitric acid (HNO3) and 2 M sodium hydroxide (NaOH) at 60 C, used in these tests were effective in dissolving the chemical plugs, the 2 M nitric acid was significantly more effective in dissolving the NAH and NAS plugs. The caustic was only slightly more effecting at dissolving the NAP plug. In the bench-scale dissolution tests, hot (60 C) 2 M nitric acid was the most effective solvent in that it completely dissolved both NAH and NAS chemical plugs much faster (1.5 - 2 x) than 2 M sodium hydroxide. So unless there are operational benefits for the use of caustic verses nitric acid, 2 M nitric acid heated to 60 C should be the solvent of choice for dissolving these chemical plugs. Flow-loop testing was planned to identify a combination of parameters such as pressure, flush solution, composition, and temperature that would effectively dissolve and flush each type of chemical plug from preformed chemical plugs in 3-inch-diameter and 4-feet-long pipe sections. However, based on a review of the results of the bench-top tests and technical discussions, the Waste Treatment Plant (WTP) Research and Technology (R and T), Engineering and Mechanical Systems (EMS), and Operations concluded that flow-loop testing of the chemically plugged pipe sections

  19. IAEA activities on steam generator life management

    International Nuclear Information System (INIS)

    Gueorguiev, B.; Lyssakov, V.; Trampus, P.

    2002-01-01

    steam generators, the relevant degradation mechanisms on heat exchanger tubes, to survey of national approaches on determining heat exchanger tube structural integrity and safety margins (plugging criteria) and to verify nondestructive inspection results and establish correlation between non-destructive testing (NDT) results and degradation. The co-ordinated research was started in 2001. The paper intends to present the concept and some interim results of the work. (author)

  20. Plug-In Hybrid Electric Vehicle Basics | NREL

    Science.gov (United States)

    Plug-In Hybrid Electric Vehicle Basics Plug-In Hybrid Electric Vehicle Basics Imagine being able to one that's in a standard hybrid electric vehicle. The larger battery pack allows plug-in hybrids to between fill-ups) that's very similar to the range of a conventional vehicle. A plug-in hybrid vehicle's