Fire-tube immersion heater optimization program and field heater audit program

Energy Technology Data Exchange (ETDEWEB)

Croteau, P. [Petro-Canada, Calgary, AB (Canada)

2007-07-01

This presentation provided an overview of the top 5 priorities for emission reduction and eco-efficiency by the Petroleum Technology Alliance of Canada (PTAC). These included venting of methane emissions; fuel consumption in reciprocating engines; fuel consumption in fired heaters; flaring and incineration; and fugitive emissions. It described the common concern for many upstream operating companies as being energy consumption associated with immersion heaters. PTAC fire-tube heater and line heater studies were presented. Combustion efficiency was discussed in terms of excess air, fire-tube selection, heat flux rate, and reliability guidelines. Other topics included heat transfer and fire-tube design; burner selection; burner duty cycle; heater tune up inspection procedure; and insulation. Two other programs were also discussed, notably a Petro-Canada fire-tube immersion heater optimization program and the field audit program run by Natural Resources Canada. It was concluded that improved efficiency involves training; managing excess air in combustion; managing the burner duty cycle; striving for 82 per cent combustion efficiency; and providing adequate insulation to reduce energy demand. tabs., figs.

HV Switch Tube Development Program status report: April 28, 1978

International Nuclear Information System (INIS)

Winje, R.A.

1978-01-01

The HV Switch Tube Development Program encompassed development contracts to both Eimac (Division of Varian) and RCA. Both companies were required to develop a design for the tube and to build and test two tubes. The development program began in April, 1976. Currently, both companies have built tubes; Eimac has built one and RCA has built two. As initially built, both tubes exhibited unstable operation; however, RCA has implemented design changes which stabilized the tube operation. Eimac has a design modification which they believe will produce stable operation when the change is implemented. Both tubes have been tested to a limited degree and no other abnormal characteristics have been observed

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

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

Enhancement of weld failure and tube ejection model in PENTAP program

International Nuclear Information System (INIS)

Jung, Jaehoon; An, Sang Mo; Ha, Kwang Soon; Kim, Hwan Yeol

2014-01-01

The reactor vessel pressure, the debris mass, the debris temperature, and the component of material can have an effect on the penetration tube failure modes. Furthermore, these parameters are interrelated. There are some representative severe accident codes such as MELCOR, MAAP, and PENTAP program. MELCOR decides on a penetration tube failure by its failure temperature such as 1273K simply. MAAP considers all penetration failure modes and has the most advanced model for a penetration tube failure model. However, the validation work against the experimental data is very limited. PENTAP program which evaluates the possible penetration tube failure modes such as creep failure, weld failure, tube ejection, and a long term tube failure under given accident condition was developed by KAERI. The experiment for the tube ejection is being performed by KAERI. The temperature distribution and the ablation rate of both weld and lower vessel wall can be obtained through the experiment. This paper includes the updated calculation steps for the weld failure and the tube ejection modes of the PENTAP program to apply the experimental results. PENTAP program can evaluate the possible penetration tube failure modes. It still requires a large amount of efforts to increase the prediction of failure modes. Some calculation steps are necessary for applying the experimental and the numerical data in the PENTAP program. In this study, new calculation steps are added to PENTAP program to enhance the weld failure and tube ejection models using KAERI's experimental data which are the ablation rate and temperature distribution of weld and lower vessel wall

Structural integrity evaluations of CANDU pressure tubes

International Nuclear Information System (INIS)

Radu, Vasile

2003-01-01

The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (K r and L r ). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables K r and L r , the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

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)

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)

Characterization of Zircaloy-4 tubing procured for fuel cladding research programs

International Nuclear Information System (INIS)

Chapman, R.H.

1976-01-01

A quantity of Zircaloy-4 tubing [10.92 mm outside diameter by 0.635 mm wall thickness] was purchased specifically for use in a number of related fuel cladding research programs sponsored by the Division of Reactor Safety Research, Nuclear Regulatory Commission (NRC/RSR). Identical tubing (produced simultaneously and from the same ingot) was purchased concurrently by the Electric Power Research Institute (EPRI) for use in similar research programs sponsored by that organization. In this way, source variability and prior fabrication history were eliminated as parameters, thus permitting direct comparison (as far as as-received material properties are concerned) of experimental results from the different programs. The tubing is representative of current reactor technology. Consecutive serial numbers assigned to each tube identify the sequence of the individual tubes through the final tube wall reduction operation. The report presented documents the procurement activities, provides a convenient reference source of manufacturer's data and tubing distribution to the various users, and presents some preliminary characterization data. The latter have been obtained routinely in various research programs and are not complete. Although the number of analyses, tests, and/or examinations performed to date are insufficient to draw statistically valid conclusions with regard to material characterization, the data are expected to be representative of the as-received tubing. It is anticipated that additional characterizations will be performed and reported routinely by the various research programs that use the tubing

Integrated function nonimaging concentrating collector tubes for solar thermal energy

Science.gov (United States)

Winston, R.; Ogallagher, J. J.

1982-09-01

A substantial improvement in optical efficiency over contemporary external reflector evacuated tube collectors has been achieved by integrating the reflector surface into the outer glass envelope. Described are the design fabrication and test results for a prototype collector based on this concept. A comprehensive test program to measure performance and operational characteristics of a 2 sq m panel (45 tubes) has been completed. Efficiencies above 50% relative to beam at 200 C have been repeatedly demonstrated. Both the instantaneous and long term average performance of this totally stationary solar collector are comparable to those for tracking line focus parabolic troughs. The yield, reliability and stability of performance achieved have been excellent. Subcomponent assemblies and fabrication procedures have been used which are expected to be compatible with high volume production. The collector has a wide variety of applications in the 100 to 300 C range including industrial progress heat, air conditioning and Rankine engine operation.

Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

Energy Technology Data Exchange (ETDEWEB)

Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

2015-07-01

During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

Steam generator tubes integrity: In-service-inspection

International Nuclear Information System (INIS)

Comby, R.J.

1997-01-01

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

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

International Nuclear Information System (INIS)

Joo, Kyung Mun; Hong, Jun Hee

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-02-15

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

Experimental study of flow friction characteristics of integral pin-fin tubes

International Nuclear Information System (INIS)

Ding Ming; Yan Changqi; Sun Licheng

2007-01-01

Friction characteristics of integral pin-fin tubes, through which lubricating-oil flowed vertically, were studied experimentally. Effects of the pitch, the height of fins and the machining direction on friction coefficient were analyzed. The experimental results showed that the friction coefficient of the integral pin-fro tube was obviously lager than that of smooth tube. Compared with other influential factors, the effect of the height of fins was dominant. Because the three-dimensional pin fin could disturb and destroy the boundary layer, when the Reynolds Number reached 200-300, the friction coefficient curve began to bend, that was, a turning point was appeared in the friction coefficient curve. (authors)

Assembly and method for testing the integrity of stuffing tubes

Energy Technology Data Exchange (ETDEWEB)

Morrison, E.F.

1996-12-31

A stuffing tube integrity checking assembly includes first and second annular seals, with each seal adapted to be positioned about a stuffing tube penetration component. An annular inflation bladder is provided, the bladder having a slot extending longitudinally there along and including a separator for sealing the slot. A first valve is in fluid communication with the bladder for introducing pressurized fluid to the space defined by the bladder when mounted about the tube. First and second releasible clamps are provided. Each clamp assembly is positioned about the bladder for securing the bladder to one of the seals for thereby establishing a fluid-tight chamber about the tube.

Structural and leakage integrity assessment of WWER steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Structural and leakage integrity assessment of WWER steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

Structural and leakage integrity assessment of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Otruba, J.

1997-01-01

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

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

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

3D integrated HYDRA simulations of hohlraums including fill tubes

Science.gov (United States)

Marinak, M. M.; Milovich, J.; Hammel, B. A.; Macphee, A. G.; Smalyuk, V. A.; Kerbel, G. D.; Sepke, S.; Patel, M. V.

2017-10-01

Measurements of fill tube perturbations from hydro growth radiography (HGR) experiments on the National Ignition Facility show spoke perturbations in the ablator radiating from the base of the tube. These correspond to the shadow of the 10 μm diameter glass fill tube cast by hot spots at early time. We present 3D integrated HYDRA simulations of these experiments which include the fill tube. Meshing techniques are described which were employed to resolve the fill tube structure and associated perturbations in the simulations. We examine the extent to which the specific illumination geometry necessary to accommodate a backlighter in the HGR experiment contributes to the spoke pattern. Simulations presented include high resolution calculations run on the Trinity machine operated by the Alliance for Computing at Extreme Scale (ACES) partnership. This work was performed under the auspices of the Lawrence Livermore National Security, LLC, (LLNS) under Contract No. DE-AC52-07NA27344.

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-07-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

A charged-particle manipulator utilizing a co-axial tube electrodynamic trap with an integrated camera

International Nuclear Information System (INIS)

Jiang, L; Pau, S; Whitten, W B

2011-01-01

A charged-particle manipulator was designed and fabricated with an integrated imaging camera allowing real-time in-situ monitoring of trapped particle motion even when the trap device is under motion or rotation. The trap device was made of two co-axial electrically conductive tubes with diameters of 5.5 mm and 7 mm for the inner tube and outer tube, respectively; the imaging camera with its optical fiber bundle was integrated within the tubular trap device to realize a single instrument functioning as a manipulator. Motion of suspended microparticles of 3 μm to 50 μm in diameter can be monitored using the integrated camera regardless of the trap device orientations. This manipulator provides capability of controlled manipulation of trapped particles by tuning the operating conditions while monitoring the feedback of real-time particle motion. Imaging of suspended particles was not interrupted while the manipulator was translated and/or rotated. This integrated manipulator can be used for charged particle transport and repositioning.

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

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.

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin [Power Engineering Research Institute, KEPCO Engineering and Construction, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2014-08-15

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor.

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

International Nuclear Information System (INIS)

Oh, Young-Jin; Chang, Yoon-Suk

2014-01-01

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

A Novel Approach for an Integrated Straw Tube-Microstrip Detector

Science.gov (United States)

Basile, E.; Bellucci, F.; Benussi, L.; Bertani, M.; Bianco, S.; Caponero, M. A.; Colonna, D.; Di Falco, F.; Fabbri, F. L.; Felli, F.; Giardoni, M.; La Monaca, A.; Mensitieri, G.; Ortenzi, B.; Pallotta, M.; Paolozzi, A.; Passamonti, L.; Pierluigi, D.; Pucci, C.; Russo, A.; Saviano, G.; Casali, F.; Bettuzzi, M.; Bianconi, D.; Baruffaldi, F.; Perilli, E.; Massa, F.

2006-06-01

We report on a novel concept of silicon microstrips and straw tubes detector, where integration is accomplished by a straw module with straws not subjected to mechanical tension in a Rohacell/spl reg/ lattice and carbon fiber reinforced plastic shell. Results on mechanical and test beam performances are reported as well.

Structural integrity assessments of steam generator tubes using the FAD methodology

Energy Technology Data Exchange (ETDEWEB)

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

2015-12-15

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

Structural integrity assessments of steam generator tubes using the FAD methodology

International Nuclear Information System (INIS)

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

2015-01-01

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

A Novel Approach for an Integrated Straw tube-Microstrip Detector

OpenAIRE

Basile, E.; Bellucci, F.; Benussi, L.; Bertani, M.; Bianco, S.; Caponero, M. A.; Colonna, D.; Di Falco, F.; Fabbri, F. L.; Felli, F.; Giardoni, M.; La Monaca, A.; Mensitieri, G.; Ortenzi, B.; Pallotta, M.

2005-01-01

We report on a novel concept of silicon microstrips and straw tubes detector, where integration is accomplished by a straw module with straws not subjected to mechanical tension in a Rohacell $^{\\circledR}$ lattice and carbon fiber reinforced plastic shell. Results on mechanical and test beam performances are reported on as well.

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.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-07-01

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

Vibro-impact responses of a tube with tube--baffle interaction

International Nuclear Information System (INIS)

Shin, Y.S.; Sass, D.E.; Jendrzejczyk, J.A.

1978-01-01

The relatively small, inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the vibrational characteristics and the response of the tube. Numerical studies were made to predict the vibro-impact response of a tube with tube-baffle interaction. The finite element method has been employed with a non-linear elastic contact spring-dashpot to model the effect of the relative approach between the tube and the baffle plate. The coupled equations of motion are directly integrated with a proportional system damping represented by a linear combination of mass and stiffness. Lumped mass approach with explicit time integration scheme was found to be a suitable choice for tube-baffle impacting analysis. Fourier analyses indicate that the higher mode contributions to the tube response are significant for strong tube-baffle impacting. The contact damping forces are negligible compared with the contact spring forces. The numerical analysis results are in reasonably good agreement with those of the experiments

Steam generator tube rupture simulation using extended finite element method

Energy Technology Data Exchange (ETDEWEB)

Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurin; Natesan, Ken

2016-08-15

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

Steam generator tube rupture simulation using extended finite element method

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

2016-01-01

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

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

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

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.

Vibro-impact responses of a tube with tube--baffle interaction. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Shin, Y S; Sass, D E; Jendrzejczyk, J A

1978-01-01

The relatively small, inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the vibrational characteristics and the response of the tube. Numerical studies were made to predict the vibro-impact response of a tube with tube-baffle interaction. The finite element method has been employed with a non-linear elastic contact spring-dashpot to model the effect of the relative approach between the tube and the baffle plate. The coupled equations of motion are directly integrated with a proportional system damping represented by a linear combination of mass and stiffness. Lumped mass approach with explicit time integration scheme was found to be a suitable choice for tube-baffle impacting analysis. Fourier analyses indicate that the higher mode contributions to the tube response are significant for strong tube-baffle impacting. The contact damping forces are negligible compared with the contact spring forces. The numerical analysis results are in reasonably good agreement with those of the experiments.

Pressure tube replacement in Pickering NGS A units 1 and 2

International Nuclear Information System (INIS)

Irvine, H.S.; Bennett, E.J.; Talbot, K.H.

1986-10-01

Being able to technically and economically replace the most radioactive components (excluding the nuclear fuel) in operating reactors will help to ensure the ongoing acceptance of nuclear power as a viable energy source for the future. Ontario Hydro is well along the path to meeting the above objective for its CANDU-PHW reactors. Following the failure of a Zircaloy-II pressure tube in unit 2 of Pickering NGS A in August, 1983, Ontario Hydro has embarked on a program to replace all Zircaloy-II pressure tubes in units 1 and 2 at Pickering. This program integrates the in-house research, design, construction, and operating skills of a large utility (Ontario Hydro) with the skills of a national nuclear organization (Atomic Energy of Canada Limited) and the private engineering sector of the Canadian nuclear industry. The paper describes the background to the pressure tube failure in Pickering unit 2 and to the efforts incurred in understanding the failure mechanism and how similar failures are not expected for the zirconium-niobium pressure tube material used in all other large CANDU-PHW units after units 1 and 2 of Pickering NGS A. The tooling developed for the pressure tube replacement program is described as well as the organization to undertake the program in an operating nuclear station. The retubing of units 1 and 2 at Pickering NGS A is nearing a successful completion and shows the benefits of being able to integrate the various skills required for this success. Pressure tube replacement in a CANDU-PHW reactor is equivalent to replacement of the reactor vessel in a LWR. The fact that this replacement can be done economically and with acceptable radiation dose to workers augurs well for the continued viability of the use of nuclear energy for the benefit of mankind. (author)

Bruce and Darlington power pulse and pressure tube integrity programs -status 1995

Energy Technology Data Exchange (ETDEWEB)

Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Wylie, J [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1996-12-31

The optimum solution to pressure tube fretting at the inlet of the Bruce and Darlington channels, a concern which became very serious following inspections in early 1992, is to remove the inlet bundle and operate with a 12 fuel bundle channel. During analysis of this operating mode a `power pulse` was identified which could occur during an inlet header break where all the fuel in the channel moved rapidly to the inlet of the channel. The pulse was unacceptable and the units were derated until solutions could be implemented. A number of solutions were identified and each station has begun implementation of their specific solution. Implementation has not been without problems and this paper provides a status report on the progress to date of the long bundle implementation solution for Bruce B and Darlington and the fuelling with the flow solution being implemented at Bruce A. Both types of solution have a significant impact on the original concern, fretting of the pressure tube. (author). 1 ref., 6 figs.

Miniaturised Prandtl tube with integrated pressure sensors for micro-thruster plume characterisation

NARCIS (Netherlands)

Dijkstra, Marcel; Ma, Kechun; de Boer, Meint J.; Groenesteijn, Jarno; Lötters, Joost Conrad; Wiegerink, Remco J.

2014-01-01

A miniaturised Prandtl-tube sensor incorporating a 6 mm long 40 μm diameter microchannel with integrated pressure sensors has been realised. The sensor has been designed for the characterisation of rarefied plume flow from a MEMS-based monopropellant propulsion system for high-accuracy attitude

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

Directory of Open Access Journals (Sweden)

HEOK-SOON LIM

2014-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-02-15

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Integrating a Traveling Wave Tube into an AECR-U ion source

Energy Technology Data Exchange (ETDEWEB)

Covo, Michel Kireeff; Benitez, Janilee Y.; Ratti, Alessandro; Vujic, Jasmina L.

2011-07-01

An RF system of 500W - 10.75 to 12.75 GHz was designed and integrated into the Advanced Electron Cyclotron Resonance - Upgrade (AECR-U) ion source of the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory. The AECR-U produces ion beams for the Cyclotron giving large flexibility of ion species and charge states. The broadband frequency of a Traveling Wave Tube (TWT) allows modifying the volume that couples and heats the plasma. The TWT system design and integration with the AECR-U ion source and results from commissioning are presented.

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

SpekCalc: a program to calculate photon spectra from tungsten anode x-ray tubes

International Nuclear Information System (INIS)

Poludniowski, G; Evans, P M; Landry, G; DeBlois, F; Verhaegen, F

2009-01-01

A software program, SpekCalc, is presented for the calculation of x-ray spectra from tungsten anode x-ray tubes. SpekCalc was designed primarily for use in a medical physics context, for both research and education purposes, but may also be of interest to those working with x-ray tubes in industry. Noteworthy is the particularly wide range of tube potentials (40-300 kVp) and anode angles (recommended: 6-30 deg.) that can be modelled: the program is therefore potentially of use to those working in superficial/orthovoltage radiotherapy, as well as diagnostic radiology. The utility is free to download and is based on a deterministic model of x-ray spectrum generation (Poludniowski 2007 Med. Phys. 34 2175). Filtration can be applied for seven materials (air, water, Be, Al, Cu, Sn and W). In this note SpekCalc is described and illustrative examples are shown. Predictions are compared to those of a state-of-the-art Monte Carlo code (BEAMnrc) and, where possible, to an alternative, widely-used, spectrum calculation program (IPEM78). (note)

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

A summary of the assessment of fuel behaviour, fission product release and pressure tube integrity following a postulated large loss-of-coolant accident

International Nuclear Information System (INIS)

Langman, V.J.; Weaver, K.R.

1984-05-01

The Ontario Hydro analyses of fuel and pressure tube temperatures, fuel behaviour, fission product release and pressure tube integrity for large break loss-of-coolant accidents in Bruce A or Pickering A have been critically reviewed. The determinations of maximum fuel temperatures and fission product release are very uncertain, and pressure tube integrity cannot be assured where low steam flows are predicted to persist for times on the order of minutes

40° image intensifier tubes in an integrated helmet system

Science.gov (United States)

Schreyer, Herbert; Boehm, Hans-Dieter V.; Svedevall, B.

1993-12-01

EUROCOPTER has been under contract to the French and German ministries of defence for five years to develop the TIGER, a second generation anti-tank helicopter. A piloting thermal imager has been installed on a steerable platform in the helicopter nose in order to achieve the possibility of flying round the clock. In addition to this sensor, which is sensitive at a wavelength of 10 micrometers , the German side has proposed using an Integrated Helmet System in the PAH 2. This helmet, manufactured by GEC-Marconi Avionics, incorporates two cathode ray tubes (CRT) and two image intensifier tubes which allow the pilot to use an additional sensor in the visible and near infrared spectrum. The electronic part will be built by Teldix. EUROCOPTER DEUTSCHLAND has received the first demonstrator of this helmet for testing in the EUROCOPTER Visionics Laboratory. Later, the C-prototype will be integrated into a BK 117 helicopter (AVT Avionik Versuchstrager). This new helmet has a field of view of 40 degree(s), and exit pupil of 15 mm and improved possibilities of adjusting the optical part. Laboratory tests have been carried out to test important parameters like optical resolution under low light level conditions, field of view, eye relief or exit pupil. The CRT channels have been tested for resolution, distortion, vignetting and homogeneity. The requirements and the properties of the helmet, test procedures and the results of these tests are presented in the paper.

Operative behaviour of a condenser tube under ETA chemistry

International Nuclear Information System (INIS)

Chocron, Mauricio; Burkart, Arturo; Rodriguez, Ivanna; Raul, Manera; Diego, Quinteros

2012-09-01

Among the various recommendations for the surveillance of the integrity of the materials of the Secondary Cycle (Balance of Plant) it is the periodic removal of a steam generator tube and a condenser tube and their analysis. It considers assessment of the water chemistry, corrosion and the reciprocal effect on or from other components of the cycle. Embalse N.P.P. is a CANDU 6 type, Pressurized Heavy Water Reactor, located in Cordoba Province, Argentina. Previous papers have shown results on tubes removed from the steam generators (Bordoni et al., NPC'08, September 15-18, 2008, Berlin, Germany; 6 th Canadian Nuclear Society - Steam Generators Conference, November 8-11, 2009, Toronto, Canada). Considering that the Embalse BOP has mixed metallurgy, i.e., steam generator tubes made of A800, piping made of ferrous alloys and condenser tubes made of Admiralty Brass and also taking into account that the chemistry has been modified from Morpholine control to ETA control (Fernandez et. al, NPC'2010, October 3-7, Quebec City, Canada), it has been decided to remove and analyze a condenser tube that has been placed in operation coincidently with the establishment of the ETA chemical control. The extraction is dated along with the November 2011 Plant Programmed Outage. Objectives are assessing the operative behavior of the tube performing visual and optical microscope inspection, SEM analysis of the oxides and deposits in exposed surfaces and occluded locations like tube sheet and other tests as well. Results are compared to the same analysis performed on a new tube in storage and integrated with the chemical operative figures of the cycle during the period: chemical data and corrosion products transport. (authors)

OTSGI--a program analysing two-phase flow instabilities in helical tubes of once-through steam generator

International Nuclear Information System (INIS)

Shi Shaoping; Zhou Fangde; Wang Maohua

1998-01-01

The author has studied the two-phases flow instabilities of the helical tubes of once-through steam generator. Using liner-frequency-domain analytical method, the authors have derived out a mathematical model and designed the program. In this model, the authors also have considered the thermal dynamic characteristics of the tube's wall. The program is used to calculate the threshold of the stability and the influences of some factors, such as entrance throttling coefficient, system pressure, entrance supercooling degree, et al. The outcomes are compared with other studies

Thermal performance of capillary micro tubes integrated into the sandwich element made of concrete

DEFF Research Database (Denmark)

Mikeska, Tomas; Svendsen, Svend

2013-01-01

integrated into the thin plate of sandwich element made of HPC can supply the energy needed for heating and cooling. The investigations were conceived as a low temperature concept, where the difference between the temperature of circulating fluid and air in the room was kept in range of 1 to 4°C. © (2013......The thermal performance of radiant heating and cooling systems (RHCS) composed of capillary micro tubes (CMT) integrated into the inner plate of sandwich elements made of High Performance Concrete (HPC) was investigated in the article. Temperature distribution in HPC elements around integrated CMT...

Automation in tube finishing bay

International Nuclear Information System (INIS)

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

1997-01-01

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

Optimization of a Two Stage Pulse Tube Refrigerator for the Integrated Current Lead System

Science.gov (United States)

Maekawa, R.; Matsubara, Y.; Okada, A.; Takami, S.; Konno, M.; Tomioka, A.; Imayoshi, T.; Hayashi, H.; Mito, T.

2008-03-01

Implementation of a conventional current lead with a pulse tube refrigerator has been validated to be working as an Integrated Current Lead (ICL) system for the Superconducting Magnetic Energy Storage (SMES). Realization of the system is primarily accounted for the flexibility of a pulse tube refrigerator, which does not posses any mechanical piston and/or displacer. As for an ultimate version of the ICL system, a High Temperature Superconducting (HTS) lead links a superconducting coil with a conventional copper lead. To ensure the minimization of heat loads to the superconducting coil, a pulse tube refrigerator has been upgraded to have a second cooling stage. This arrangement reduces not only the heat loads to the superconducting coil but also the operating cost for a SMES system. A prototype two-stage pulse tube refrigerator, series connected arrangement, was designed and fabricated to satisfy the requirements for the ICL system. Operation of the first stage refrigerator is a four-valve mode, while the second stage utilizes a double inlet configuration to ensure its confined geometry. The paper discusses the optimization of second stage cooling to validate the conceptual design

Simulation of the space-time evolution of color-flux tubes (guidelines to the TERMITE program)

International Nuclear Information System (INIS)

Dyrek, A.

1990-08-01

We give the description of the computer program which simulates boost-invariant evolution of color-flux tubes in high-energy processes. The program provides a graphic demonstration of space-time trajectories of created particles and can also be used as Monte-Carlo generator of events. (author)

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

The Aphrodite boiling crisis program. Analysis of CHF tests performed on a vertical tube

International Nuclear Information System (INIS)

Souyri, A.; Conan, S.; Portesse, A.; Tremblay, D.

1992-09-01

In order to develop a comprehensive modelling of the boiling crisis phenomenon, the APHRODITE experimental program has been set up at ELECTRICITE DE FRANCE. Aiming at a better mechanistic understanding of this phenomenon, this program will investigate the influence of the experimental conditions (among which the mockup geometry and the boundary conditions) and the two-phase flow patterns via void fraction distributions. It has involved the construction of a R12 test loop, which can deliver a large thermal-hydraulic parameter ranges, and the development of a gamma-ray tomograph. The first experiments have been carried out on a vertical Inconel tube, 6 meters long with a bore diameter of 13 mm and a thickness of 0.5 mm. This electrically heated test section is heavily instrumented with 168 thermocouples welded along the tube, on its outer surface. After a refined calibration of the experimental procedure, a critical heat flux data bank has been collected within large pressure, mass velocity and critical steam quality ranges. These results are firstly compared with other CHF data obtained in similar conditions. Then several empirical correlations and a theoretical model for similar prediction in tubes are tested against these data

Helically coiled tube heat exchanger

International Nuclear Information System (INIS)

Harris, A.M.

1981-01-01

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

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

International Nuclear Information System (INIS)

Yoshihisa Shindo; Kazuo Haga

2005-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-02-15

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

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 2: Reactor pressure vessel embrittlement and thermal annealing; Reactor vessel lower head integrity; Evaluation and projection of steam generator tube condition and integrity

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: reactor pressure vessel embrittlement and thermal annealing; reactor vessel lower head integrity; and evaluation and projection of steam generator tube condition and integrity. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

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.

SRS Tank Structural Integrity Program

International Nuclear Information System (INIS)

Maryak, Matthew

2010-01-01

The mission of the Structural Integrity Program is to ensure continued safe management and operation of the waste tanks for whatever period of time these tanks are required. Matthew Maryak provides an overview of the Structural Integrity Program to open Session 5 (Waste Storage and Tank Inspection) of the 2010 EM Waste Processing Technical Exchange.

Streak tube development

International Nuclear Information System (INIS)

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

1979-01-01

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

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)

Study of thermal performance of capillary micro tubes integrated into the building sandwich element made of high performance concrete

DEFF Research Database (Denmark)

Mikeska, Tomas; Svendsen, Svend

2013-01-01

The thermal performance of radiant heating and cooling systems (RHCS) composed of capillary micro tubes (CMT) integrated into the inner plate of sandwich elements made of high performance concrete (HPC) was investigated in the article. Temperature distribution in HPC elements around integrated CM...... and cooling purposes of future low energy buildings. The investigations were conceived as a low temperature concept, where the difference between the temperature of circulating fluid and air in the room was kept in range of 1–4 °C.......The thermal performance of radiant heating and cooling systems (RHCS) composed of capillary micro tubes (CMT) integrated into the inner plate of sandwich elements made of high performance concrete (HPC) was investigated in the article. Temperature distribution in HPC elements around integrated CMT...... HPC layer covering the CMT. This paper shows that CMT integrated into the thin plate of sandwich element made of HPC can supply the energy needed for heating (cooling) and at the same time create the comfortable and healthy environment for the occupants. This solution is very suitable for heating...

Integration of a photocatalytic multi-tube reactor for indoor air purification in HVAC systems: a feasibility study.

Science.gov (United States)

van Walsem, Jeroen; Roegiers, Jelle; Modde, Bart; Lenaerts, Silvia; Denys, Siegfried

2018-04-24

This work is focused on an in-depth experimental characterization of multi-tube reactors for indoor air purification integrated in ventilation systems. Glass tubes were selected as an excellent photocatalyst substrate to meet the challenging requirements of the operating conditions in a ventilation system in which high flow rates are typical. Glass tubes show a low-pressure drop which reduces the energy demand of the ventilator, and additionally, they provide a large exposed surface area to allow interaction between indoor air contaminants and the photocatalyst. Furthermore, the performance of a range of P25-loaded sol-gel coatings was investigated, based on their adhesion properties and photocatalytic activities. Moreover, the UV light transmission and photocatalytic reactor performance under various operating conditions were studied. These results provide vital insights for the further development and scaling up of multi-tube reactors in ventilation systems which can provide a better comfort, improved air quality in indoor environments, and reduced human exposure to harmful pollutants.

A Laboratory Experimental Study: An FBG-PVC Tube Integrated Device for Monitoring the Slip Surface of Landslides

Science.gov (United States)

Zhang, Shaojie; Chen, Jiang; Teng, Pengxiao; Wei, Fangqiang; Chen, Qiao

2017-01-01

A new detection device was designed by integrating fiber Bragg grating (FBG) and polyvinyl chloride (PVC) tube in order to monitor the slip surface of a landslide. Using this new FBG-based device, a corresponding slope model with a pre-set slip surface was designed, and seven tests with different soil properties were carried out in laboratory conditions. The FBG sensing fibers were fixed on the PVC tube to measure strain distributions of PVC tube at different elevation. Test results indicated that the PVC tube could keep deformation compatible with soil mass. The new device was able to monitor slip surface location before sliding occurrence, and the location of monitored slip surface was about 1–2 cm above the pre-set slip surface, which basically agreed with presupposition results. The monitoring results are expected to be used to pre-estimate landslide volume and provide a beneficial option for evaluating the potential impact of landslides on shipping safety in the Three Gorges area. PMID:29084157

Association of a Proactive Swallowing Rehabilitation Program With Feeding Tube Placement in Patients Treated for Pharyngeal Cancer.

Science.gov (United States)

Ajmani, Gaurav S; Nocon, Cheryl C; Brockstein, Bruce E; Campbell, Nicholas P; Kelly, Amy B; Allison, Jamie; Bhayani, Mihir K

2018-04-19

A proactive speech and language pathology (SLP) program is an important component of the multidisciplinary care of patients with head and neck squamous cell carcinoma (HNSCC). Swallowing rehabilitation can reduce the rate of feeding tube placement, thereby significantly improving quality of life. To evaluate the initiation of a proactive SLP rehabilitation program at a single institution and its association with rates of feeding tube placement and dietary intake in patients with HNSCC. Cohort study at a tertiary care and referral center for patients with HNSCC serving the northern Chicago region. Patients were treated for squamous cell carcinomas of the hypopharynx, oropharynx, and nasopharynx from 2004 to 2015 with radiation or chemoradiation therapy in the definitive or adjuvant setting. Patients who received less than 5000 cGy radiation or underwent reirradiation were excluded. A proactive SLP program for patients with HNSCC was initiated in 2011. Study cohorts were divided into 2 groups: 2004 through 2010 and 2011 through 2015. Primary outcome variables were SLP referral placement and timing of the referral. Secondary outcomes were feeding tube placement and ability to tolerate any oral intake. A total of 254 patients met inclusion criteria (135 before and 119 after implementation of SLP program; median age, 60 years [range, 14-94 years]; 77% male). With the initiation of a proactive SLP program, pretreatment evaluations increased from 29 (21.5%) to 70 (58.8%; risk ratio [RR], 2.74; 95% CI, 1.92-3.91), and rate of referral overall at any time increased from 60.0% to 79.8% (RR, 1.33; 95% CI, 1.13-1.57). Feeding tube placement rates decreased from 45.9% (n = 62) to 29.4% (n = 35; RR, 0.64; 95% CI, 0.46-0.89). Among patients receiving a swallow evaluation, feeding tube requirements were less frequent for those receiving a pretreatment evaluation (31 of 99 [31%]) than for those referred during (11 of 18 [61%]) or after (38 of 59 [64%]) treatment. The rate

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)

Manufacturing of tailored tubes with a process integrated heat treatment

Science.gov (United States)

Hordych, Illia; Boiarkin, Viacheslav; Rodman, Dmytro; Nürnberger, Florian

2017-10-01

The usage of work-pieces with tailored properties allows for reducing costs and materials. One example are tailored tubes that can be used as end parts e.g. in the automotive industry or in domestic applications as well as semi-finished products for subsequent controlled deformation processes. An innovative technology to manufacture tubes is roll forming with a subsequent inductive heating and adapted quenching to obtain tailored properties in the longitudinal direction. This processing offers a great potential for the production of tubes with a wide range of properties, although this novel approach still requires a suited process design. Based on experimental data, a process simulation is being developed. The simulation shall be suitable for a virtual design of the tubes and allows for gaining a deeper understanding of the required processing. The model proposed shall predict microstructural and mechanical tube properties by considering process parameters, different geometries, batch-related influences etc. A validation is carried out using experimental data of tubes manufactured from various steel grades.

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

International Nuclear Information System (INIS)

Fontes, S.W.F.

1989-01-01

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

State Program Integrity Assessment (SPIA)

Data.gov (United States)

U.S. Department of Health & Human Services — The State Program Integrity Assessment (SPIA) is the Centers for Medicare and Medicaid Services (CMS) first national data collection on state Medicaid program...

Development of zirconium alloy tube manufacturing technology

International Nuclear Information System (INIS)

Kim, In Kyu; Park, Chan Hyun; Lee, Seung Hwan; Chung, Sun Kyo

2009-01-01

In late 2004, Korea Nuclear Fuel Company (KNF) launched a government funded joint development program with Westinghouse Electric Co. (WEC) to establish zirconium alloy tube manufacturing technology in Korea. Through this program, KNF and WEC have developed a state of the art facility to manufacture high quality nuclear tubes. KNF performed equipment qualification tests for each manufacturing machine with the support of WEC, and independently carried out product qualification tests for each tube product to be commercially produced. Apart from those tests, characterization test program consisting of specification test and characterization test was developed by KNF and WEC to demonstrate to customers of KNF the quality equivalency of products manufactured by KNF and WEC plants respectively. As part of establishment of performance evaluation technology for zirconium alloy tube in Korea, KNF carried out analyses of materials produced for the characterization test program using the most advanced techniques. Thanks to the accomplishment of the development of zirconium alloy tube manufacturing technology, KNF is expected to acquire positive spin off benefits in terms of technology and economy in the near future

An integrated automatic system for the eddy-current testing of the steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Woo, Hee Gon; Choi, Seong Su [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

1995-12-31

This research project was focused on automation of steam generator tubes inspection for nuclear power plants. ECT (Eddy Current Testing) inspection process in nuclear power plants is classified into 3 subprocesses such as signal acquisition process, signal evaluation process, and inspection planning and data management process. Having been automated individually, these processes were effectively integrated into an automatic inspection system, which was implemented in HP workstation with expert system developed (author). 25 refs., 80 figs.

An integrated automatic system for the eddy-current testing of the steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Woo, Hee Gon; Choi, Seong Su [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

1996-12-31

This research project was focused on automation of steam generator tubes inspection for nuclear power plants. ECT (Eddy Current Testing) inspection process in nuclear power plants is classified into 3 subprocesses such as signal acquisition process, signal evaluation process, and inspection planning and data management process. Having been automated individually, these processes were effectively integrated into an automatic inspection system, which was implemented in HP workstation with expert system developed (author). 25 refs., 80 figs.

Development of Program Evaluating the Effects on the Secondary Side of Steam Generator due to Foreign Objects

International Nuclear Information System (INIS)

Ju, Yoo Hyun; Nam, Choi Sung

2005-01-01

When materials such as metal are into the secondary side of steam generator, they, so called foreign objects, may have influences on the integrity of the steam generator tubes. They cause the tube wear due to the relative motion between the tubes and foreign objects and the tube impact due to flow. The best way to avoid the effects is to remove all the foreign objects. However, it is not easy to remove the foreign materials thoroughly due to their condition such as the location. If the locations of the foreign materials are in the middle of tube bundle and the tube arrangement of the steam generator is the triangle type, the equipment such as FOSAR(Foreign Object Search and Retrieval) can not reach their locations. If the foreign materials stick together with the tubes or tube sheet, they can not be removed. In the case of operating the steam generator with the foreign materials, the licensee must prove that the materials do not affect the tube integrity and do not threaten the pressure boundary with the analytical method. Considering the wear and impact by the foreign materials, KEPRI(Korea Electric Power Research Institute) developed the methodology to evaluate the foreign materials analytically. This methodology was described with a computer program in order to obtain the fast results. The program informs whether the tubes have the structural integrity when the foreign material strikes the tubes. Moreover, this gives us the remaining life of the steam generator tubes. In this paper, the program, which evaluates the effects of the foreign objects in the secondary side of steam generator, is introduced

Encapsulation of Fluidic Tubing and Microelectrodes in Microfluidic Devices: Integrating Off-Chip Process and Coupling Conventional Capillary Electrophoresis with Electrochemical Detection.

Science.gov (United States)

Becirovic, Vedada; Doonan, Steven R; Martin, R Scott

2013-08-21

In this paper, an approach to fabricate epoxy or polystyrene microdevices with encapsulated tubing and electrodes is described. Key features of this approach include a fixed alignment between the fluidic tubing and electrodes, the ability to polish the device when desired, and the low dead volume nature of the fluidic interconnects. It is shown that a variety of tubing can be encapsulated with this approach, including fused silica capillary, polyetheretherketone (PEEK), and perfluoroalkoxy (PFA), with the resulting tubing/microchip interface not leading to significant band broadening or plug dilution. The applicability of the devices with embedded tubing is demonstrated by integrating several off-chip analytical methods to the microchip. This includes droplet transfer, droplet desegmentation, and microchip-based flow injection analysis. Off-chip generated droplets can be transferred to the microchip with minimal coalescence, while flow injection studies showed improved peak shape and sensitivity when compared to the use of fluidic interconnects with an appreciable dead volume. Importantly, it is shown that this low dead volume approach can be extended to also enable the integration of conventional capillary electrophoresis (CE) with electrochemical detection. This is accomplished by embedding fused silica capillary along with palladium (for grounding the electrophoresis voltage) and platinum (for detection) electrodes. With this approach, up to 128,000 theoretical plates for dopamine was possible. In all cases, the tubing and electrodes are housed in a rigid base; this results in extremely robust devices that will be of interest to researchers wanting to develop microchips for use by non-experts.

30 CFR 250.517 - Tubing and wellhead equipment.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 2 2010-07-01 2010-07-01 false Tubing and wellhead equipment. 250.517 Section... Tubing and wellhead equipment. (a) No tubing string shall be placed in service or continue to be used unless such tubing string has the necessary strength and pressure integrity and is otherwise suitable for...

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Program collaboration and service integration activities among HIV programs in 59 U.S. health departments.

Science.gov (United States)

Fitz Harris, Lauren F; Toledo, Lauren; Dunbar, Erica; Aquino, Gustavo A; Nesheim, Steven R

2014-01-01

We identified the level and type of program collaboration and service integration (PCSI) among HIV prevention programs in 59 CDC-funded health department jurisdictions. Annual progress reports (APRs) completed by all 59 health departments funded by CDC for HIV prevention activities were reviewed for collaborative and integrated activities reported by HIV programs for calendar year 2009. We identified associations between PCSI activities and funding, AIDS diagnosis rate, and organizational integration. HIV programs collaborated with other health department programs through data-related activities, provider training, and providing funding for sexually transmitted disease (STD) activities in 24 (41%), 31 (53%), and 16 (27%) jurisdictions, respectively. Of the 59 jurisdictions, 57 (97%) reported integrated HIV and STD testing at the same venue, 39 (66%) reported integrated HIV and tuberculosis testing, and 26 (44%) reported integrated HIV and viral hepatitis testing. Forty-five (76%) jurisdictions reported providing integrated education/outreach activities for HIV and at least one other disease. Twenty-six (44%) jurisdictions reported integrated partner services among HIV and STD programs. Overall, the level of PCSI activities was not associated with HIV funding, AIDS diagnoses, or organizational integration. HIV programs in health departments collaborate primarily with STD programs. Key PCSI activities include integrated testing, integrated education/outreach, and training. Future assessments are needed to evaluate PCSI activities and to identify the level of collaboration and integration among prevention programs.

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

Experimental Study of Concrete-filled Carbon Fiber Reinforced Polymer Tube with Internal Reinforcement under Axially Loading

Directory of Open Access Journals (Sweden)

Wenbin SUN

2014-12-01

Full Text Available Comparing with the circular concrete columns confined with fiber reinforced polymer (FRP wrap or tube, the rectilinear confined columns were reported much less. Due to the non-uniform distribution of confining pressure in the rectilinear confined columns, the FRP confinement effectiveness was significant reduced. This paper presents findings of an experimental program where nine prefabricated rectangular cross-section CFRP tubes with CFRP integrated crossties filled concrete to form concrete-filled FRP tube (CFFT short columns and three plain concrete control specimens were tested. All specimens were axially loaded until failure. The rest results showed that the stress-strain curves of CFFTs consisted of two distinct branches, an ascending branch before the concrete peak stress was reaches and a second branch that terminated when the tube ruptured, and that the CFFTs with integrated crossties experienced most uniform confinement pressure distribution. Test research also found that the stress-strain curves of CFFTs indicated an increase in ductility. These demonstrate that this confinement system can produce higher lateral confinement stiffness. DOI: http://dx.doi.org/10.5755/j01.ms.20.4.6035

Clinical tube weaning supported by hunger provocation in fully-tube-fed children.

Science.gov (United States)

Hartdorff, Caroline M; Kneepkens, C M Frank; Stok-Akerboom, Anita M; van Dijk-Lokkart, Elisabeth M; Engels, Michelle A H; Kindermann, Angelika

2015-04-01

Children with congenital malformations, mental retardation, and complex early medical history frequently have feeding problems. Although tube feeding is effective in providing the necessary energy and nutrients, it decreases the child's motivation to eat and may lead to oral aversion. In this study, we sought to confirm our previous results, showing that a multidisciplinary clinical hunger provocation program may lead to quick resumption of oral feeding. In a crossover study, 22 children of 9 to 24 months of age who were fully dependent on tube feeding were randomly assigned to one of two groups: group A, intervention group (2-week multidisciplinary clinical hunger provocation program); and group B, control group (4-week outpatient treatment by the same multidisciplinary team). Patients failing one treatment were reassigned to the other treatment group. Primary outcome measures were at least 75% orally fed at the conclusion of the intervention and fully orally fed and gaining weight 6 months after the intervention. In group A, 9/11 patients were successfully weaned from tube feeding (2 failures: 1 developed ulcerative colitis, 1 drop-out). In group B, only 1 patient was weaned successfully; 10/11 were reassigned to the clinical hunger provocation program, all being weaned successfully. Six months after the intervention, 1 patient had to resume tube feeding. In total, in the control group, 1/11 (9%) was weaned successfully as compared with 18/21 (86%) in the hunger provocation group (P hunger provocation is an effective short-term intervention for weaning young children from tube feeding.

Tube collector with integrated tracking parabolic concentrator

Energy Technology Data Exchange (ETDEWEB)

Grass, C.; Benz, N.; Hacker, Z.; Timinger, A. [ZAE Bayern, Bavarian Centre for Applied Energy Research, Muenchen (Germany)

2000-07-01

Low concentrating CPC collectors usually do not track the sun and are mounted in east-west direction with a latitude dependent slope angle. They are most suitable for maximum working temperatures up to 200 250 deg. C. We present a novel evacuated tube-collector with a trough-like concentrating mirror. Single-axis tracking of the mirror is realized with a magnetic mechanism. The mirror is mounted inside the evacuated tube and hence protected from environmental influences. One axis tracking in combination with a small acceptance angle allows for higher concentration as compared to non-tracking concentrating collectors. Ray-tracing analysis shows a half acceptance angle of about 5 deg. at a geometrical concentration ratio of 3.2. The losses of evacuated tube collectors are dominated by the radiation losses of the absorber. Hence, reducing the absorber size can lead to higher efficiencies at high operating temperature levels. With the presented collector we aim for operating temperatures up to 400 deg. C. At temperatures of 300 deg. C we expect efficiencies of 65 %. This allows for application in industrial process heat generation, high efficient solar cooling and power generation. A first prototype was tested at the ZAE Bayern. The optical efficiency was measured to be 75 %. (au)

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)

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

International Nuclear Information System (INIS)

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

1997-01-01

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

NEI You Tube Videos: Amblyopia

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Full Text Available ... Program Vision and Aging Program African American Program Training and Jobs Fellowships NEI Summer Intern Program Diversity In Vision Research & Ophthalmology (DIVRO) Student Training Programs To search for current job openings visit HHS USAJobs Home >> NEI YouTube Videos >> ...

Integration of finite element analysis and design of experiments to analyse the geometrical factors in bi-layered tube hydroforming

International Nuclear Information System (INIS)

Alaswad, A.; Olabi, A.G.; Benyounis, K.Y.

2011-01-01

Tube hydroforming (THF) is a type of unconventional metal forming process in which high fluid pressure and axial feed are used to deform a tube blank in the desired shape. Bi-layered tube hydroforming is suitable to produce bi-layered joints to be used in special applications such as aerospace, oil production and nuclear power plants. In this work, a finite element study along with response surface methodology (RSM) for design of experiment (DOE) has been used to construct models for three responses namely: bulge height, thickness reduction, and wrinkle height as a function of geometrical factors for X shape bi-layered tube hydroforming. A finite element model was built and experimentally validated. The models developed using finite element analysis (FEA) and RSM was found to be educated. The factors effect and their interactions on the three responses were determined and discussed. Such integration was proved to be a successful technique that can be used to predict the geometry of the hydroformed part.

NEI You Tube Videos: Amblyopia

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Full Text Available ... and Aging Program African American Program Training and Jobs Fellowships NEI Summer Intern Program Diversity In Vision ... DIVRO) Student Training Programs To search for current job openings visit HHS USAJobs Home >> NEI YouTube Videos >> ...

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.

Fabrication and Characterization of All-Polystyrene Microfluidic Devices with Integrated Electrodes and Tubing.

Science.gov (United States)

Pentecost, Amber M; Martin, R Scott

2015-01-01

A new method of fabricating all-polystyrene devices with integrated electrodes and fluidic tubing is described. As opposed to expensive polystyrene (PS) fabrication techniques that use hot embossing and bonding with a heated lab press, this approach involves solvent-based etching of channels and lamination-based bonding of a PS cover, all of which do not need to occur in a clean room. PS has been studied as an alternative microchip substrate to PDMS, as it is more hydrophilic, biologically compatible in terms of cell adhesion, and less prone to absorption of hydrophobic molecules. The etching/lamination-based method described here results in a variety of all-PS devices, with or without electrodes and tubing. To characterize the devices, micrographs of etched channels (straight and intersected channels) were taken using confocal and scanning electron microscopy. Microchip-based electrophoresis with repetitive injections of fluorescein was conducted using a three-sided PS (etched pinched, twin-tee channel) and one-sided PDMS device. Microchip-based flow injection analysis, with dopamine and NO as analytes, was used to characterize the performance of all-PS devices with embedded tubing and electrodes. Limits of detection for dopamine and NO were 130 nM and 1.8 μM, respectively. Cell immobilization studies were also conducted to assess all-PS devices for cellular analysis. This paper demonstrates that these easy to fabricate devices can be attractive alternative to other PS fabrication methods for a wide variety of analytical and cell culture applications.

NEI You Tube Videos: Amblyopia

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Full Text Available ... Program Vision and Aging Program African American Program Training and Jobs Fellowships NEI Summer Intern Program Diversity In Vision Research & Ophthalmology (DIVRO) Student Training Programs To search for current job openings visit HHS USAJobs Home » NEI YouTube Videos » ...

Integrated Computer Controlled Glow Discharge Tube

Science.gov (United States)

Kaiser, Erik; Post-Zwicker, Andrew

2002-11-01

An "Interactive Plasma Display" was created for the Princeton Plasma Physics Laboratory to demonstrate the characteristics of plasma to various science education outreach programs. From high school students and teachers, to undergraduate students and visitors to the lab, the plasma device will be a key component in advancing the public's basic knowledge of plasma physics. The device is fully computer controlled using LabVIEW, a touchscreen Graphical User Interface [GUI], and a GPIB interface. Utilizing a feedback loop, the display is fully autonomous in controlling pressure, as well as in monitoring the safety aspects of the apparatus. With a digital convectron gauge continuously monitoring pressure, the computer interface analyzes the input signals, while making changes to a digital flow controller. This function works independently of the GUI, allowing the user to simply input and receive a desired pressure; quickly, easily, and intuitively. The discharge tube is a 36" x 4"id glass cylinder with 3" side port. A 3000 volt, 10mA power supply, is used to breakdown the plasma. A 300 turn solenoid was created to demonstrate the magnetic pinching of a plasma. All primary functions of the device are controlled through the GUI digital controllers. This configuration allows for operators to safely control the pressure (100mTorr-1Torr), magnetic field (0-90Gauss, 7amps, 10volts), and finally, the voltage applied across the electrodes (0-3000v, 10mA).

NEI You Tube Videos: Amblyopia

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Full Text Available ... and Aging Program African American Program Training and Jobs Fellowships NEI Summer Intern Program Diversity In Vision ... DIVRO) Student Training Programs To search for current job openings visit HHS USAJobs Home » NEI YouTube Videos » ...

Mixed wasted integrated program: Logic diagram

International Nuclear Information System (INIS)

Mayberry, J.; Stelle, S.; O'Brien, M.; Rudin, M.; Ferguson, J.; McFee, J.

1994-01-01

The Mixed Waste Integrated Program Logic Diagram was developed to provide technical alternative for mixed wastes projects for the Office of Technology Development's Mixed Waste Integrated Program (MWIP). Technical solutions in the areas of characterization, treatment, and disposal were matched to a select number of US Department of Energy (DOE) treatability groups represented by waste streams found in the Mixed Waste Inventory Report (MWIR)

Mixed wasted integrated program: Logic diagram

Energy Technology Data Exchange (ETDEWEB)

Mayberry, J.; Stelle, S. [Science Applications International Corp., Idaho Falls, ID (United States); O`Brien, M. [Univ. of Arizona, Tucson, AZ (United States); Rudin, M. [Univ. of Nevada, Las Vegas, NV (United States); Ferguson, J. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States); McFee, J. [I.T. Corp., Albuquerque, NM (United States)

1994-11-30

The Mixed Waste Integrated Program Logic Diagram was developed to provide technical alternative for mixed wastes projects for the Office of Technology Development`s Mixed Waste Integrated Program (MWIP). Technical solutions in the areas of characterization, treatment, and disposal were matched to a select number of US Department of Energy (DOE) treatability groups represented by waste streams found in the Mixed Waste Inventory Report (MWIR).

The effect of tube-support interaction on the dynamic response of heat exchanger tubes

International Nuclear Information System (INIS)

Shin, Y.S.; Jendrzejczyk, J.A.; Wambsganss, M.W.

1977-01-01

To avoid detrimental tube vibration in heat exchangers, resonant conditions and instabilitites must be avoided, and/or peak dynamic amplitudes must not exceed allowable limits. In attempting a theoretical analysis, questions arise as to the effects of tube/support interaction on tube vibrational characteristics (i.e. resonant frequencies, modes, damping) and response amplitude. As a part of ANL's Flow-Induced Vibration Program in support of the Clinch River Breeder Reactor Plant (CRBRP) steam generator design activity, tube/support interaction experiments are being performed not only to gain the insight into the dynamic behavior of CRBRP steam generator tubes, but also to provide the basis for developing design guidance. Test results were compared with anaytical results based on multispan tube with 'knife-edge' supports at the support locations. (Auth.)

State Program Integrity Reviews

Data.gov (United States)

U.S. Department of Health & Human Services — State program integrity reviews play a critical role in how CMS provides effective support and assistance to states in their efforts to combat provider fraud and...

Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

International Nuclear Information System (INIS)

Bucholz, J.A.

2000-01-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source

Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

Energy Technology Data Exchange (ETDEWEB)

Bucholz, J.A.

2000-07-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

Creating a YouTube-Like Collaborative Environment in Mathematics: Integrating Animated Geogebra Constructions and Student-Generated Screencast Videos

Science.gov (United States)

Lazarus, Jill; Roulet, Geoffrey

2013-01-01

This article discusses the integration of student-generated GeoGebra applets and Jing screencast videos to create a YouTube-like medium for sharing in mathematics. The value of combining dynamic mathematics software and screencast videos for facilitating communication and representations in a digital era is demonstrated herein. We share our…

Trial manufacture of simple integrated tube-type pyranometer by phycoerythrin and measurements of transmittance of solar radiation in crop canopies

International Nuclear Information System (INIS)

Yamamoto, H.; Honjo, H.; Kamota, F.; Suzuki, Y.; Hayakawa, S.

1998-01-01

We tried to construct a simple integrated tube-type pyranometer using phycoerythrin from seaweed pigment. The maximum sensitive wavehand of phycoerythrin was 550 nm - 560 nm, and this waveband was in the photosynthetically active radiation range. The acrylic tubes (outside diameter, 22 mm, length, 100 cm) were spread with white paints except for a strip 15 mm in width, and phycoerythrin was put into the acrylic tube. In the results from the outdoor measurements, the tube-type pyranometer showed a positive correlation between the transmittance of phycoerythrin (%) and the measured accumulated solar radiation (MJ n(-2)), but the slope of the linear equation was different in summer and winter. In an artificial climate room, the relationship between the transmissions of phycoerythrin and the accumulated solar radiation could be approximated by a quadratic equation at every temperature. In the measurements made outdoors, the accumulated solar radiation could be estimated using the transmittance of phycoerythrin and the mean air temperature during measurements

Self-shielding flex-circuit drift tube, drift tube assembly and method of making

Science.gov (United States)

Jones, David Alexander

2016-04-26

The present disclosure is directed to an ion mobility drift tube fabricated using flex-circuit technology in which every other drift electrode is on a different layer of the flex-circuit and each drift electrode partially overlaps the adjacent electrodes on the other layer. This results in a self-shielding effect where the drift electrodes themselves shield the interior of the drift tube from unwanted electro-magnetic noise. In addition, this drift tube can be manufactured with an integral flex-heater for temperature control. This design will significantly improve the noise immunity, size, weight, and power requirements of hand-held ion mobility systems such as those used for explosive detection.

Development, prevention, and treatment of feeding tube dependency.

Science.gov (United States)

Krom, Hilde; de Winter, J Peter; Kindermann, Angelika

2017-06-01

Enteral nutrition is effective in ensuring nutritional requirements and growth. However, when tube feeding lasts for a longer period, it can lead to tube dependency in the absence of medical reasons for continuation of tube feeding. Tube-dependent children are unable or refuse to start oral activities and they lack oral skills. Tube dependency has health-, psychosocial-, and economy-related consequences. Therefore, the transition to oral feeding is of great importance. However, this transition can be very difficult and needs a multidisciplinary approach. Most studies for treatment of tube dependency are based on behavioral interventions, such as family therapy, individual behavior therapy, neuro-linguistic programming, and parental anxiety reduction. Furthermore, oral motor therapy and nutritional adjustments can be helpful in tube weaning. The use of medication has been described in the literature. Although mostly chosen as the last resort, hunger-inducing methods, such as the Graz-model and the Dutch clinical hunger provocation program, are also successful in weaning children off tube feeding. The transition from tube to oral feeding is important in tube-dependent children but can be difficult. We present an overview for the prevention and treatment of tube dependency. What is known: • Longer periods of tube feeding can lead to tube dependency. • Tube weaning can be very difficult. What is new: • Weaning as soon as possible and therefore referral to a multidisciplinary team are recommended. • An overview of treatment options for tube dependency is presented in this article.

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Characterization, Monitoring and Sensor Technology Integrated Program

International Nuclear Information System (INIS)

1993-01-01

This booklet contains summary sheets that describe FY 1993 characterization, monitoring, and sensor technology (CMST) development projects. Currently, 32 projects are funded, 22 through the OTD Characterization, Monitoring, and Sensor Technology Integrated Program (CMST-IP), 8 through the OTD Program Research and Development Announcement (PRDA) activity managed by the Morgantown Energy Technology Center (METC), and 2 through Interagency Agreements (IAGs). This booklet is not inclusive of those CMST projects which are funded through Integrated Demonstrations (IDs) and other Integrated Programs (IPs). The projects are in six areas: Expedited Site Characterization; Contaminants in Soils and Groundwater; Geophysical and Hydrogeological Measurements; Mixed Wastes in Drums, Burial Grounds, and USTs; Remediation, D ampersand D, and Waste Process Monitoring; and Performance Specifications and Program Support. A task description, technology needs, accomplishments and technology transfer information is given for each project

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-07-01

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

Computation and measurement of calandria tube sag in PHWR

International Nuclear Information System (INIS)

Kim, Tae Ryong; Sohn, Seok Man

2003-01-01

Calandria tubes and liquid injection shutdown system (LISS) tubes in a pressurized heavy water reactor (PHWR) is known to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with LISS tube crossing beneath and calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted. (author)

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

Residual indicator bacteria in autosampler tubing: a field and laboratory assessment.

Science.gov (United States)

Hathaway, J M; Hunt, W F; Guest, R M; McCarthy, D T

2014-01-01

Microbial contamination in surface waters has become a worldwide cause for concern. As efforts are made to reduce this contamination, monitoring is integral to documenting and evaluating water quality improvements. Autosamplers are beneficial in such monitoring efforts, as large data sets can be generated with minimized effort. The extent to which autosamplers can be utilized for microbial monitoring is largely unknown due to concerns over contamination. Strict sterilization regimes for components contacting the water being sampled are difficult, and sometimes logistically implausible, when utilizing autosamplers. Field experimentation showed contamination of fecal coliform in autosamplers to be more of a concern than that of Escherichia coli. Further study in a controlled laboratory environment suggested that tubing configuration has a significant effect on residual E. coli concentrations in sampler tubing. The amount of time that passed since the last sample was collected from a given sampler (antecedent dry weather period - DWP) tubing was also a significant factor. At a DWP of 7 days, little to no contamination was found. Thus, simple protocols such as providing positive drainage of tubing between sample events and programming samplers to include rinses will reduce concerns of contamination in autosamplers.

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

ACCEPT: a three-dimensional finite element program for large deformation elastic-plastic-creep analysis of pressurized tubes (LWBR/AWBA Development Program)

International Nuclear Information System (INIS)

Hutula, D.N.; Wiancko, B.E.

1980-03-01

ACCEPT is a three-dimensional finite element computer program for analysis of large-deformation elastic-plastic-creep response of Zircaloy tubes subjected to temperature, surface pressures, and axial force. A twenty-mode, tri-quadratic, isoparametric element is used along with a Zircaloy materials model. A linear time-incremental procedure with residual force correction is used to solve for the time-dependent response. The program features an algorithm which automatically chooses the time step sizes to control the accuracy and numerical stability of the solution. A contact-separation capability allows modeling of interaction of reactor fuel rod cladding with fuel pellets or external supports

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

Development of technology on the material surveillance of CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author).

Development of technology on the material surveillance of CANDU pressure tubes

International Nuclear Information System (INIS)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author)

Disc-Donut-Tube wear test report, Phase I

International Nuclear Information System (INIS)

Kowal, K.; Knaus, S.E.

1976-06-01

The report describes a test program which simulated the wear-inducing conditions in the AI Prototype CRBR Steam Generator. This was accomplished by simulating the wear inducing loading and motion of a steam tube against ''disc-donut'' tube spacer plates. It was found that 2- 1 / 4 Cr-1 Mo tubes, wearing against 2- 1 / 4 Cr-l Mo tube spacer plates, seized and galled as deep as .017 inches. Inconel 718 tube spacer plates uniformly wore the tubes as deep as .012 in. Aluminum bronze inserts wore as deep as .003 inches into the tube

Centrifugal LabTube platform for fully automated DNA purification and LAMP amplification based on an integrated, low-cost heating system.

Science.gov (United States)

Hoehl, Melanie M; Weißert, Michael; Dannenberg, Arne; Nesch, Thomas; Paust, Nils; von Stetten, Felix; Zengerle, Roland; Slocum, Alexander H; Steigert, Juergen

2014-06-01

This paper introduces a disposable battery-driven heating system for loop-mediated isothermal DNA amplification (LAMP) inside a centrifugally-driven DNA purification platform (LabTube). We demonstrate LabTube-based fully automated DNA purification of as low as 100 cell-equivalents of verotoxin-producing Escherichia coli (VTEC) in water, milk and apple juice in a laboratory centrifuge, followed by integrated and automated LAMP amplification with a reduction of hands-on time from 45 to 1 min. The heating system consists of two parallel SMD thick film resistors and a NTC as heating and temperature sensing elements. They are driven by a 3 V battery and controlled by a microcontroller. The LAMP reagents are stored in the elution chamber and the amplification starts immediately after the eluate is purged into the chamber. The LabTube, including a microcontroller-based heating system, demonstrates contamination-free and automated sample-to-answer nucleic acid testing within a laboratory centrifuge. The heating system can be easily parallelized within one LabTube and it is deployable for a variety of heating and electrical applications.

Program Collaboration and Service Integration At-a-Glance

Centers for Disease Control (CDC) Podcasts

Dr. Kevin A. Fenton, Director of CDC's National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention, discusses program collaboration and service integration, a strategy that promotes better collaboration between public health programs and supports appropriate service integration at the point-of-care.

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

Development of a 3D electromagnetic model for eddy current tubing inspection application to steam generator tubing

Energy Technology Data Exchange (ETDEWEB)

Maillot, V. [Institut de Radioprotection et de Surete Nucleaire, IRSN, 92 - Fontenay aux Roses (France); Pichenot, G.; Premel, D.; Sollier, T. [CEA Saclay, DRT/DECS, 91 - Gif-sur-Yvette (France)

2003-10-01

In nuclear plants, the inspection of heat exchanger tubes is usually carried out by using eddy current nondestructive testing. A numerical model, based on a volume integral approach using the Green's dyadic formalism, has been developed, with support from the French Institute for Radiological Protection and Nuclear Safety, to predict the response of an eddy current bobbin coil to 3D flaws located in the tube's wall. With an aim of integrating this model into the NDE multi techniques platform CIVA, it has been validated with experimental data for 2D and 3D flaws. (authors)

Boosting program integrity and effectiveness of the cognitive behavioral program EQUIP for incarcerated youth in The Netherlands

NARCIS (Netherlands)

Helmond, P.; Overbeek, G.; Brugman, D.

2014-01-01

This study examined whether a "program integrity booster" could improve the low to moderate program integrity and effectiveness of the EQUIP program for incarcerated youth as practiced in The Netherlands. Program integrity was assessed in EQUIP groups before and after the booster. Youth residing in

Planning integration FY 1996 program plan. Revision 1

International Nuclear Information System (INIS)

1995-09-01

This Multi-Year Program Plan (MAP) Planning Integration Program, Work Breakdown Structure (WBS) Element 1.8.2, is the primary management tool to document the technical, schedule, and cost baseline for work directed by the US Department of Energy (DOE), Richland Operations Office (RL). As an approved document, it establishes an agreement between RL and the performing contractors for the work to be performed. It was prepared by Westinghouse Hanford Company (WHC) and Pacific Northwest Laboratory (PNL). The MYPPs for the Hanford Site programs are to provide a picture from fiscal year (FY) 1996 through FY 2002. At RL Planning and Integration Division (PID) direction, only the FY 1996 Planning Integration Program work scope has been planned and presented in this MAP. Only those known significant activities which occur after FY 1996 are portrayed in this MAP. This is due to the uncertainty of who will be accomplishing what work scope when, following the award of the Management and Integration (M ampersand I) contract

High temperature ceramic-tubed reformer

Science.gov (United States)

Williams, Joseph J.; Rosenberg, Robert A.; McDonough, Lane J.

1990-03-01

The overall objective of the HiPHES project is to develop an advanced high-pressure heat exchanger for a convective steam/methane reformer. The HiPHES steam/methane reformer is a convective, shell and tube type, catalytic reactor. The use of ceramic tubes will allow reaction temperature higher than the current state-of-the-art outlet temperatures of about 1600 F using metal tubes. Higher reaction temperatures increase feedstock conversion to synthesis gas and reduce energy requirements compared to currently available radiant-box type reformers using metal tubes. Reforming of natural gas is the principal method used to produce synthesis gas (primarily hydrogen and carbon monoxide, H2 and CO) which is used to produce hydrogen (for refinery upgrading), methanol, as well as several other important materials. The HiPHES reformer development is an extension of Stone and Webster's efforts to develop a metal-tubed convective reformer integrated with a gas turbine cycle.

Characterization of tube support alloys

International Nuclear Information System (INIS)

Vaia, A.R.

1985-01-01

The involvement and relationship of carbon steel corrosion products in the tube denting phenomenon promoted an intensive research effort to: 1) understand, reproduce, and arrest the denting process, and 2) evaluate alternative tube support materials to provide additional corrosion resistance. The paper summarizes a corrosion testing program for the verification of type 405 stainless steel under acid or all volatile treatment conditions

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...

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)

N Reactor pressure tube 2566 postirradiation examination

International Nuclear Information System (INIS)

Scott, K.V.

1978-01-01

Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

Preliminary Stress Analysis of an IHX Tube Support Plate in Prototype SFR

International Nuclear Information System (INIS)

Kim, Sung Kyun; Koo, Gyeong Hoi

2013-01-01

In this paper, the structural integrity about the conceptual design of IHX tube support plate was reviewed and the design should be changed because of its high stress concentration at the outer rim area. For reducing its maximum stress, two alternatives were proposed and reviewed for the structural integrity point of view. In both proposing support designs, the maximum stress decreases up to the stress design limit. Tube support plates (TSPs) of the intermediate heat exchanger (IHX) in Prototype GenIV Sodium Cooled Fast Reactor (PGSFR) act to horizontally support IHX tubes against hydraulic loadings and they have numerous flow holes where a primary sodium flows downward and secondary sodium flows upward. Due to its many penetrations, its geometric shape is quite complex and structurally its integrity is quite weaker than other parts. In this study, we investigated the structural integrity of the conceptually designed IHX tube support plate. In addition, TSP's supporting concepts were proposed to increase its structural integrity, and confirmed its integrity by using a finite element analysis

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

25 CFR 39.132 - Can a school integrate Language Development programs into its regular instructional program?

Science.gov (United States)

2010-04-01

... 25 Indians 1 2010-04-01 2010-04-01 false Can a school integrate Language Development programs into... Language Development Programs § 39.132 Can a school integrate Language Development programs into its regular instructional program? A school may offer Language Development programs to students as part of its...

Description of a program for steam generators; Descripcion de un programa de generadores de vapor

Energy Technology Data Exchange (ETDEWEB)

Campana, F. J.

2014-10-01

Steam Generators (SGs) are a key component of PWR nuclear power plants, maintaining their structural integrity throughout their life time is necessary to allow for long term operation (LTD) of PWR plants. NEI 97-06 provides the fundamental elements to be included in a SG Program. In addiction it describes performance criteria that SG tubes have to meet in order to provide reasonable assurance that the tubes are still able to maintain specific safety function. Hence, it is mandatory for plants with SGs to have defined a SG program consistent with NEI 97-06 and contains the elements which are described by it. This Program must contain some elements such as, Degradation Assessment, inspection and Integrity Assessment, among other. (Author)

Full length channel Pressure Tube sagging under completely voided full length pressure tube of an Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Negi, Sujay, E-mail: negi.sujay@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Kumar, Ravi, E-mail: ravikfme@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India); Mukopadhyay, D., E-mail: dmukho@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India)

2017-03-15

Highlights: • At 16 kW/m input, thermal stability was attained at 595 °C, without PT-CT contact. • At 20 kW/m step input, PT-CT contact occurred at 637 °C near bottom-center of the tube. • PT integrity was maintained throughout the experiment. - Abstract: An experimental investigation was conducted to simulate the sagging behavior of a full length Pressure Tube of a channel of 220 MWe Indian PHWR. The investigation aimed to recreate a condition resembling Loss of Coolant Accident (LOCA) with Emergency Core Cooling System (ECCS) failure in a nuclear power plant. A full length channel assembly immersed in moderator was subjected to electrical resistance heating of Pressure Tube (PT) to simulate the residual heat after shutting down of reactor. The temperature of PT started rising and the contact between PT and CT was established at the center of the tube where average bottom temperature was 637 °C. The integrity of PT was maintained throughout the experiment and the PT heat up was arrested on contact with the CT due to transfer of heat to the moderator.

Exploring Art and Science Integration in an Afterschool Program

Science.gov (United States)

Bolotta, Alanna

Science, technology, engineering, arts and math (STEAM) education integrates science with art, presenting a unique and interesting opportunity to increase accessibility in science for learners. This case study examines an afterschool program grounded in art and science integration. Specifically, I studied the goals of the program, it's implementation and the student experience (thinking, feeling and doing) as they participated in the program. My findings suggest that these programs can be powerful methods to nurture scientific literacy, creativity and emotional development in learners. To do so, this program made connections between disciplines and beyond, integrated holistic teaching and learning practices, and continually adapted programming while also responding to challenges. The program is therefore specially suited to engage the heads, hands and hearts of learners, and can make an important contribution to their learning and development. To conclude, I provide some recommendations for STEAM implementation in both formal and informal learning settings.

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Science.gov (United States)

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

2014-02-01

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

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

Development of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

International Nuclear Information System (INIS)

Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun

2004-02-01

The objective of this research is to develop on efficient integrity evaluation technology and to investigate the applicability of the newly-developed technology such as internet-based cyber platform etc. to Nuclear Power Plant(NPP) components. The development of an efficient structural integrity evaluation system is necessary for safe operation of NPP as the increase of operating periods. Moreover, material test data as well as emerging structural integrity assessment technology are also needed for the evaluation of aged components. The following five topics are covered in this project: development of the wall-thinning evaluation program for nuclear piping; development of structural integrity evaluation criteria for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for major components of NPP; ingegration of internet-based cyber platform and integrity evaluation program for primary components of NPP; effects of aging on strength of dissimilar welds

Development of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

2004-02-15

The objective of this research is to develop on efficient integrity evaluation technology and to investigate the applicability of the newly-developed technology such as internet-based cyber platform etc. to Nuclear Power Plant(NPP) components. The development of an efficient structural integrity evaluation system is necessary for safe operation of NPP as the increase of operating periods. Moreover, material test data as well as emerging structural integrity assessment technology are also needed for the evaluation of aged components. The following five topics are covered in this project: development of the wall-thinning evaluation program for nuclear piping; development of structural integrity evaluation criteria for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for major components of NPP; ingegration of internet-based cyber platform and integrity evaluation program for primary components of NPP; effects of aging on strength of dissimilar welds.

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

International Nuclear Information System (INIS)

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

2010-09-01

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

Gastrostomy Tube (G-Tube)

Science.gov (United States)

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

In Situ Remediation Integrated Program: FY 1994 program summary

International Nuclear Information System (INIS)

1995-04-01

The US Department of Energy (DOE) established the Office of Technology Development (EM-50) as an element of the Office of Environmental Management (EM) in November 1989. In an effort to focus resources and address priority needs, EM-50 introduced the concept of integrated programs (IPs) and integrated demonstrations (IDs). The In Situ Remediation Integrated Program (ISR IP) focuses research and development on the in-place treatment of contaminated environmental media, such as soil and groundwater, and the containment of contaminants to prevent the contaminants from spreading through the environment. Using in situ remediation technologies to clean up DOE sites minimizes adverse health effects on workers and the public by reducing contact exposure. The technologies also reduce cleanup costs by orders of magnitude. This report summarizes project work conducted in FY 1994 under the ISR IP in three major areas: treatment (bioremediation), treatment (physical/chemical), and containment technologies. Buried waste, contaminated soils and groundwater, and containerized waste are all candidates for in situ remediation. Contaminants include radioactive waste, volatile and nonvolatile organics, heavy metals, nitrates, and explosive materials

In Situ Remediation Integrated Program: FY 1994 program summary

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-04-01

The US Department of Energy (DOE) established the Office of Technology Development (EM-50) as an element of the Office of Environmental Management (EM) in November 1989. In an effort to focus resources and address priority needs, EM-50 introduced the concept of integrated programs (IPs) and integrated demonstrations (IDs). The In Situ Remediation Integrated Program (ISR IP) focuses research and development on the in-place treatment of contaminated environmental media, such as soil and groundwater, and the containment of contaminants to prevent the contaminants from spreading through the environment. Using in situ remediation technologies to clean up DOE sites minimizes adverse health effects on workers and the public by reducing contact exposure. The technologies also reduce cleanup costs by orders of magnitude. This report summarizes project work conducted in FY 1994 under the ISR IP in three major areas: treatment (bioremediation), treatment (physical/chemical), and containment technologies. Buried waste, contaminated soils and groundwater, and containerized waste are all candidates for in situ remediation. Contaminants include radioactive waste, volatile and nonvolatile organics, heavy metals, nitrates, and explosive materials.

Integrating Robot Task Planning into Off-Line Programming Systems

DEFF Research Database (Denmark)

Sun, Hongyan; Kroszynski, Uri

1988-01-01

a system architecture for integrated robot task planning. It identifies and describes the components considered necessary for implementation. The focus is on functionality of these elements as well as on the information flow. A pilot implementation of such an integrated system architecture for a robot......The addition of robot task planning in off-line programming systems aims at improving the capability of current state-of-the-art commercially available off-line programming systems, by integrating modeling, task planning, programming and simulation together under one platform. This article proposes...... assembly task is discussed....

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)

42 CFR 455.232 - Medicaid integrity audit program contractor functions.

Science.gov (United States)

2010-10-01

... 42 Public Health 4 2010-10-01 2010-10-01 false Medicaid integrity audit program contractor functions. 455.232 Section 455.232 Public Health CENTERS FOR MEDICARE & MEDICAID SERVICES, DEPARTMENT OF HEALTH AND HUMAN SERVICES (CONTINUED) MEDICAL ASSISTANCE PROGRAMS PROGRAM INTEGRITY: MEDICAID Medicaid...

YouTube and Facebook

DEFF Research Database (Denmark)

Robertson, Scott P.; Vatrapu, Ravi; Medina, Richard

This paper examines the links to YouTube from the Facebook “walls” of Barack Obama, Hillary Clinton, and John McCain over two years prior to the 2008 U.S. Presidential election. User-generated linkage patterns show how participants in these politically-related social networking dialogues used...... online video to make their points. We show a strong integration of the Web 2.0 and new media technologies of social networking and online video. We argue that political discussion in social networking environments can no longer be viewed as primarily textual, and that neither Facebook nor YouTube can...

Advanced evacuated tube collectors

Science.gov (United States)

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

1985-04-01

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

Experimental study for transient response of a double-tube thermosyphon (DTTH)

International Nuclear Information System (INIS)

Salem, M.A.M.

2010-01-01

Energy conservation is becoming increasingly important as the cost of fuel continuously rises. The heat pipe and the closed two-phase thermosyphon are particularly effective tools in the heat transfer process.A theoretical and experimental investigation was conducted to study the double-tube two-phase closed-thermosyphon (DTTH) behavior in transient regimes. Experiments were performed to investigate the effects of changing the heating and cooling rate as well as the evaporator length on the double tube thermosyphon in actual integrated operation (start-up, steady-state and shut-down). he necessity for a dynamic model of DTTH for some applications of discontinuous operation imposed the need to the current applied investigation. Therefore, the main objective of the current study is to develop a theoretical model that can predict the dynamic behavior of the double-tube evaporator by tracing various transient parameters during operation from start up to steady state until shut down condition. A model describing both thermal and phase flows of the closed two-phase double tube thermosyphon (DTTH) has been simulated. The theoretical model provides a general description of the behavior of our practical setup based on experimental observations which show a simple exponential behavior. It is based on a two thermal body description (evaporator wall and working fluid) there is good agreement between experiments data and numerical prediction.A computer simulation program based on the method was developed to estimate temperature and the other performance of double tube thermosyphon as well as the time needed to reach steady state condition. The governing equations of the simple 1-D model were solved by Engineering Equation Solver program (EES) using finite difference Euler method. A computer program is designed to solve these differential equations by an explicit finite difference method. The results from this model were found to be in general agreement with the experimental

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

Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

International Nuclear Information System (INIS)

Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

1982-01-01

Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

Developments of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

2003-03-15

The objective of this research is to develop an efficient evaluation technology and to investigate applicability of newly-developed technology, such as internet-based cyber platform, to operating power plants. Development of efficient evaluation systems for Nuclear Power Plant components, based on structural integrity assessment techniques, are increasingly demanded for safe operation with the increasing operating period of Nuclear Power Plants. The following five topics are covered in this project: development of assessment method for wall-thinned nuclear piping based on pipe test; development of structural integrity program for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for mam components of NPP; development of internet-based cyber platform and integrity program for primary components of NPP; effect of aging on strength of dissimilar welds.

Applications of ultrasonic phased array technique during fabrication of nuclear tubing and other components for the Indian nuclear power program

International Nuclear Information System (INIS)

Kapoor, K.

2015-01-01

Ultrasonic phased array technique has been applied in fabrication of nuclear fuel and structural at NFC. The integrity of the nuclear fuel and structural components is most crucial as they are exposed to severe environment during operation leading to rapid degradation of its properties during its lifecycle. Nuclear Fuel Complex has mandate for the fabrication of the nuclear fuel and core structurals for Indian PHWRs/BWR, sub-assemblies for the PFBR and steam generator tubing for PFBR and PHWRs which are the most critical materials for the Indian Nuclear Power program. NDE during fabrication of these materials is thus most crucial as it provides the confidence to the designer for safe operation during its lifetime. Many of these techniques have to be developed in-house to meet unique requirements of high sensitivity, resolution and shape of the components. Some of the advancements in the NDE during the fabrication include use of ultrasonic phased array which is detailed in this paper

Integral Fast Reactor Program

International Nuclear Information System (INIS)

Chang, Y.I.; Walters, L.C.; Laidler, J.J.; Pedersen, D.R.; Wade, D.C.; Lineberry, M.J.

1993-06-01

This report summarizes highlights of the technical progress made in the Integral Fast Reactor (IFR) Program in FY 1992. Technical accomplishments are presented in the following areas of the IFR technology development activities: (1) metal fuel performance, (2) pyroprocess development, (3) safety experiments and analyses, (4) core design development, (5) fuel cycle demonstration, and (6) LMR technology R ampersand D

Adaptation, testing and application of the two-dimensional FE computer program system for steam generator tube testing

International Nuclear Information System (INIS)

Betzold, K.

1987-01-01

The 2d-FE computing program system, taken over by EPRI, is used for the improvement of the eddy current test of steam generator heating tubes. The investigations focus on test tasks in the area of the tube plate and the scrap mark; among them: accumulation of mud in the cracking area and above the tube plate; circulating slots with and without accumulation of mud. The interaction of the factors of influence given by the test object and the parameters selectable by the tester as for example coil length and base space for absolute coils and differential coils as well as test frequencies are calculated and the form of the signal locus curves and the dynamic curves are listed in a sample catalogue. It is demonstrated with selected examples that the sample catalogue contributes to the test-specific design of the coil and to the choice of the test frequencies; interpretation of measured signals; deepening of the knowledge of the physical processe in eddy current tests. (orig./HP) [de

Detailed design of neutron guide tubes at the upgraded JRR-3, (1)

International Nuclear Information System (INIS)

Harami, Taikan; Umemura, Mutsumi; Ebisawa, Tohru.

1985-07-01

JRR-3, currently a heavy water moderated and cooled 10 MW reactor, is to be upgraded to a light water moderated and cooled, heavy water reflected 20 MW reactor. Two guide tubes for thermal neutron and three for cold will be installed in the reactor to transport thermal and cold neutrons from the reactor hall to the experiment hall. This describes the neutron guide tube transmission analysis program, NEUGT, which was developed to assess the design of the neutron guide tubes. The input data plotting program, PLOPINE and the output data plotting program, NEUPLOT are presented in the appendix. The NEUGT program not only calculates a neutron transmission and neutron spectra, assuming the Maxwellian spectra at the entrance of a guide tube, but also analyses the effect of abutment errors. This reports the description and the input data manual of the program in the text. Examples of analysis are given in the appendixes. The program is written in the FORTRAN 77 language for FACOM 380. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Preliminary design study of removable integral steam generator units of the multiple helically wound tube type for a 1250 MW(th) H.T.G.C. reactor

International Nuclear Information System (INIS)

Gilli, P.V.; Fritz, K.; Lippitsch, J.; Sandri, A.H.; Weiss, B.

1965-11-01

The possibilities of designing a multiple steam generator for a 1250 MW(th) High Temperature Gas-Cooled Reactor, consisting of 18 units which are able to pass through 5 ft diam. holes in the integral prestressed concrete pressure vessel are investigated. A lay-out and design with bundles of multi-start helical tubes is evolved, particular attention being paid to the questions of tube blanking and removal of the unit, and of selection of materials for superheater and reheater tubes. Thermal and stress calculations have been carried out, using the Waagner-Biro Computer Code ADURHELIX. (author)

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

Robust precision alignment algorithm for micro tube laser forming

NARCIS (Netherlands)

Folkersma, Ger; Brouwer, Dannis Michel; Römer, Gerardus Richardus, Bernardus, Engelina; Herder, Justus Laurens

2016-01-01

Tube laser forming on a small diameter tube can be used as a high precision actuator to permanently align small (optical)components. Applications, such as the alignment of optical fibers to photonic integrated circuits, often require sub-micron alignment accuracy. Although the process causes

[The development of an integrated suicide-violence prevention program for adolescents].

Science.gov (United States)

Park, Hyun Sook

2008-08-01

The purpose of this study was to develop an integrated suicide-violence prevention program for adolescents. Another purpose was to evaluate the effects of the integrated suicide-violence prevention program on self-esteem, parent-child communication, aggression, and suicidal ideation in adolescents. The study employed a quasi-experimental design. Participants for the study were high school students, 24 in the experimental group and 25 in the control group. Data was analyzed by using the SPSS/WIN. 11.5 program with chi2 test, t-test, and 2-way ANOVA. Participants in the integrated suicide-violence prevention program reported increased self-esteem scores, which was significantly different from those in the control group. Participants in the integrated suicide-violence prevention program reported decreased aggression and suicidal ideation scores, which was significantly different from those in the control group. The integrated suicide-violence prevention program was effective in improving self-esteem and decreasing aggression and suicidal ideation for adolescents. Therefore, this approach is recommended as the integrated suicide-violence prevention strategy for adolescents.

Program Collaboration and Service Integration At-a-Glance

Centers for Disease Control (CDC) Podcasts

2010-09-15

Dr. Kevin A. Fenton, Director of CDC's National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention, discusses program collaboration and service integration, a strategy that promotes better collaboration between public health programs and supports appropriate service integration at the point-of-care.  Created: 9/15/2010 by National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention.   Date Released: 9/15/2010.

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

Corrosion and Rupture of Steam Generator Tubings in PWRs

Energy Technology Data Exchange (ETDEWEB)

Hwang, Seong Sik; Kim, Hong Pyo

2007-08-15

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

Corrosion and Rupture of Steam Generator Tubings in PWRs

International Nuclear Information System (INIS)

Hwang, Seong Sik; Kim, Hong Pyo

2007-08-01

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

Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

International Nuclear Information System (INIS)

Kim, Y. S.; Jeong, Y. M.; Ahn, S. B.

2005-03-01

Major degradation of the feeder pipe is the thinning due to the flow accelerated corrosion and the cracking in the bent region due to the stress corrosion cracking. The feeder pipe in a PHWR is a pipe to supply the coolant to the pressure tube and the heated coolant to the steam generator for power generation. Approximately 380 pipes are installed on the inlet side and outlet side each with two bent regions in the 600 MW-class PHWR. After a leakage in the bent region of the feeder pipe, it is required to examine all the pipes in order to ensure the integrity of the pressure boundaries. It is not easy, however, to examine all the pipes with the conventional ultrasonic method, because of a high dose of radiation exposure and a limited accessibility to the pipe. In order to get rid of the limited accessibility, the ultrasonic guided wave method are developed for detection and evaluation of the cracks in the feeder pipe. The dispersion mode analysis was performed for the development of long-range guided wave inspection for the feeder pipe. An analytical approach for the straight pipe as well as numerical approach for the bent pipe with 2-D FFT were accomplished. A computer program for the calculation of the dispersion curves and wave structures was developed. Based on the dispersion curves and wave structure of the feeder pipe, candidates for the optimal parameters on the frequencies and vibration modes were selected. A time-frequency analysis methodology was developed for the mode identification of received ultrasonic signal. A high power tone-burst ultrasonic system has been setup for the generation of guided waves. Various artificial notches were fabricated on the bent feeder pipes for the experiment on the flaw detection. Considering the results of dispersion analysis and field condition, the torsional vibration mode, T(0,1) is selected for the first choice. An array of electromagnetic acoustic transducers (EMAT) was designed and fabricated for the generation of T

Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

Energy Technology Data Exchange (ETDEWEB)

Kim, Y S; Jeong, Y M; Ahn, S B [and others

2005-03-15

Major degradation of the feeder pipe is the thinning due to the flow accelerated corrosion and the cracking in the bent region due to the stress corrosion cracking. The feeder pipe in a PHWR is a pipe to supply the coolant to the pressure tube and the heated coolant to the steam generator for power generation. Approximately 380 pipes are installed on the inlet side and outlet side each with two bent regions in the 600 MW-class PHWR. After a leakage in the bent region of the feeder pipe, it is required to examine all the pipes in order to ensure the integrity of the pressure boundaries. It is not easy, however, to examine all the pipes with the conventional ultrasonic method, because of a high dose of radiation exposure and a limited accessibility to the pipe. In order to get rid of the limited accessibility, the ultrasonic guided wave method are developed for detection and evaluation of the cracks in the feeder pipe. The dispersion mode analysis was performed for the development of long-range guided wave inspection for the feeder pipe. An analytical approach for the straight pipe as well as numerical approach for the bent pipe with 2-D FFT were accomplished. A computer program for the calculation of the dispersion curves and wave structures was developed. Based on the dispersion curves and wave structure of the feeder pipe, candidates for the optimal parameters on the frequencies and vibration modes were selected. A time-frequency analysis methodology was developed for the mode identification of received ultrasonic signal. A high power tone-burst ultrasonic system has been setup for the generation of guided waves. Various artificial notches were fabricated on the bent feeder pipes for the experiment on the flaw detection. Considering the results of dispersion analysis and field condition, the torsional vibration mode, T(0,1) is selected for the first choice. An array of electromagnetic acoustic transducers (EMAT) was designed and fabricated for the generation of T

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)

State Program Integrity Review Reports List

Data.gov (United States)

U.S. Department of Health & Human Services — Comprehensive state program integrity (PI) review reports (and respective follow-up review reports) provide CMS assessment of the effectiveness of the states PI...

YouTube as an information source for pediatric adenotonsillectomy and ear tube surgery.

Science.gov (United States)

Sorensen, Jeffrey A; Pusz, Max D; Brietzke, Scott E

2014-01-01

Assess the overall quality of information on adenotonsillectomy and ear tube surgery presented on YouTube (www.youtube.com) from the perspective of a parent or patient searching for information on surgery. The YouTube website was systematically searched on select dates with a formal search strategy to identify videos pertaining to pediatric adenotonsillectomy and ear tube surgery. Only videos with at least 5 (ear tube surgery) or 10 (adenotonsillectomy) views per day were included. Each video was viewed and scored by two independent scorers. Videos were categorized by goal and scored for video/audio quality, accuracy, comprehensiveness, and procedure-specific content. Cross-sectional study. Public domain website. Fifty-five videos were scored for adenotonsillectomy and forty-seven for ear tube surgery. The most common category was educational (65.3%) followed by testimonial (28.4%), and news program (9.8%). Testimonials were more common for adenotonsillectomy than ear tube surgery (41.8% vs. 12.8%, p=0.001). Testimonials had a significantly lower mean accuracy (2.23 vs. 2.62, p=0.02), comprehensiveness (1.71 vs. 2.22, p=0.007), and TA specific content (0.64 vs. 1.69, p=0.001) score than educational type videos. Only six videos (5.9%) received high scores in both video/audio quality and accuracy/comprehensiveness of content. There was no significant association between the accuracy and comprehensive score and views, posted "likes", posted "dislikes", and likes/dislikes ratio. There was an association between "likes" and mean video quality (Spearman's rho=0.262, p=0.008). Parents/patients searching YouTube for information on pediatric adenotonsillectomy and ear tube surgery will generally encounter low quality information with testimonials being common but of significantly lower quality. Viewer perceived quality ("likes") did not correlate to formally scored content quality. Published by Elsevier Ireland Ltd.

Design and use of the ORNL HFIR [High Flux Isotope Reactor] pneumatic tube irradiation systems

International Nuclear Information System (INIS)

Dyer, F.F.; Emery, J.F.; Robinson, L.; Teasley, N.A.

1987-01-01

A second pneumatic tube that was recently installed in the High Flux Isotope Reactor for neutron activation analysis is described. Although not yet tested, the system is expected to have a thermal neutron flux of about 1.5 x 10 14 cm -2 s -1 . A delayed neutron counter is an integral part of the pneumatic tube, and all of the hardware is present to enable automated use of the counter. The system is operated with a Gould programmable controller that is programmed with an IBM personal computer. Automation of any mode of operation, including the delayed neutron counter, will only require a nominal amount of software development. Except for the lack of a hot cell, the irradiation facility has all of the advantageous features of an older pneumatic tube that has been in operation for 17 years. The design of the system and some applications and methods of operation are described

Integrating computer programs for engineering analysis and design

Science.gov (United States)

Wilhite, A. W.; Crisp, V. K.; Johnson, S. C.

1983-01-01

The design of a third-generation system for integrating computer programs for engineering and design has been developed for the Aerospace Vehicle Interactive Design (AVID) system. This system consists of an engineering data management system, program interface software, a user interface, and a geometry system. A relational information system (ARIS) was developed specifically for the computer-aided engineering system. It is used for a repository of design data that are communicated between analysis programs, for a dictionary that describes these design data, for a directory that describes the analysis programs, and for other system functions. A method is described for interfacing independent analysis programs into a loosely-coupled design system. This method emphasizes an interactive extension of analysis techniques and manipulation of design data. Also, integrity mechanisms exist to maintain database correctness for multidisciplinary design tasks by an individual or a team of specialists. Finally, a prototype user interface program has been developed to aid in system utilization.

Integrated program of using of Probabilistic Safety Analysis in Spain

International Nuclear Information System (INIS)

1998-01-01

Since 25 June 1986, when the CSN (Nuclear Safety Conseil) approve the Integrated Program of Probabilistic Safety Analysis, this program has articulated the main activities of CSN. This document summarize the activities developed during these years and reviews the Integrated programme

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

Two Inseparable Facets of Technology Integration Programs: Technology and Theoretical Framework

Science.gov (United States)

Demir, Servet

2011-01-01

This paper considers the process of program development aiming at technology integration for teachers. For this consideration, the paper focused on an integration program which was recently developed as part of a larger project. The participants of this program were 45 in-service teachers. The program continued four weeks and the conduct of the…

Achieving High Reliability Operations Through Multi-Program Integration

Energy Technology Data Exchange (ETDEWEB)

Holly M. Ashley; Ronald K. Farris; Robert E. Richards

2009-04-01

Over the last 20 years the Idaho National Laboratory (INL) has adopted a number of operations and safety-related programs which has each periodically taken its turn in the limelight. As new programs have come along there has been natural competition for resources, focus and commitment. In the last few years, the INL has made real progress in integrating all these programs and are starting to realize important synergies. Contributing to this integration are both collaborative individuals and an emerging shared vision and goal of the INL fully maturing in its high reliability operations. This goal is so powerful because the concept of high reliability operations (and the resulting organizations) is a masterful amalgam and orchestrator of the best of all the participating programs (i.e. conduct of operations, behavior based safety, human performance, voluntary protection, quality assurance, and integrated safety management). This paper is a brief recounting of the lessons learned, thus far, at the INL in bringing previously competing programs into harmony under the goal (umbrella) of seeking to perform regularly as a high reliability organization. In addition to a brief diagram-illustrated historical review, the authors will share the INL’s primary successes (things already effectively stopped or started) and the gaps yet to be bridged.

The U-tube: A new paradigm in borehole fluid sampling

Energy Technology Data Exchange (ETDEWEB)

Freifeld, B. M.

2009-10-01

Fluid samples from deep boreholes can provide insights into subsurface physical, chemical, and biological conditions. Recovery of intact, minimally altered aliquots of subsurface fluids is required for analysis of aqueous chemistry, isotopic composition, and dissolved gases, and for microbial community characterization. Unfortunately, for many reasons, collecting geofluids poses a number of challenges, from formation contamination by drilling to maintaining integrity during recovery from depths. Not only are there substantial engineering issues in retrieval of a representative sample, but there is often the practical reality that fluid sampling is just one of many activities planned for deep boreholes. The U-tube geochemical sampling system presents a new paradigm for deep borehole fluid sampling. Because the system is small, its ability to integrate with other measurement systems and technologies opens up numerous possibilities for multifunctional integrated wellbore completions. To date, the U-tube has been successfully deployed at four different field sites, each with a different deployment modality, at depths from 260 m to 2 km. While the U-tube has proven to be highly versatile, these installations have resulted in data that provide additional insights for improving future U-tube deployments.

Integrated maintenance program (IMP)

International Nuclear Information System (INIS)

Zemdegs, R.T.; Chout, Q.B.

1998-01-01

Approaches to the maintenance of nuclear power plants have undergone significant change in the past several decades. The traditional breakdown approach has been displaced by preventive (calendar-based) maintenance and more recently, by condition-based maintenance (CBM). This is largely driven by the fact that traditional maintenance programs, derived primarily from equipment vendor recommendations, are generally unsuccessful in controlling maintenance costs or equipment failures. Many advances in the maintenance field have taken place since the maintenance plans for Ontario Hydro's nuclear plants were initially established. Ontario Hydro nuclear plant operating costs can be substantially reduced and Incapability Factor improved with the application of modern maintenance processes and tools. Pickering is designated as the lead station for IMP. Of immediate concern is the fact that Pickering Nuclear Division has been experiencing a significant backlog of Operating Preventive Maintenance Callups. This backlog, over 2000, is unacceptable to both station management and the nuclear regulator, the Atomic Energy Control Board. In addition there are over 500 callups in various stages of revision (in hyperspace) without an adequate control nor reporting system to manage their completion. There is also considerable confusion about the classification of l icensing c allups, e.g. callups which are mandatory as a result of legal requirements. Furthermore the ineffectiveness of the Preventive Maintenance (PM) has been the subject of peer audits and Atomic Energy Control Board (AECB) findings over the past several years. The current preventive maintenance ratio PM2 /(PM+CM3) at Pickering ND is less than 20%, due to the current high load of equipment breakdown. This past summer, an Independent Integrated Performance Assessment (IIPA) review at Ontario Hydro confirmed these concerns. Over the past several years, Ontario Hydro nuclear staff have evaluated several programs to improve

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)

Ocean acidification impacts spine integrity but not regenerative capacity of spines and tube feet in adult sea urchins

Science.gov (United States)

Emerson, Chloe E.; Reinardy, Helena C.; Bates, Nicholas R.

2017-01-01

Increasing atmospheric carbon dioxide (CO2) has resulted in a change in seawater chemistry and lowering of pH, referred to as ocean acidification. Understanding how different organisms and processes respond to ocean acidification is vital to predict how marine ecosystems will be altered under future scenarios of continued environmental change. Regenerative processes involving biomineralization in marine calcifiers such as sea urchins are predicted to be especially vulnerable. In this study, the effect of ocean acidification on regeneration of external appendages (spines and tube feet) was investigated in the sea urchin Lytechinus variegatus exposed to ambient (546 µatm), intermediate (1027 µatm) and high (1841 µatm) partial pressure of CO2 (pCO2) for eight weeks. The rate of regeneration was maintained in spines and tube feet throughout two periods of amputation and regrowth under conditions of elevated pCO2. Increased expression of several biomineralization-related genes indicated molecular compensatory mechanisms; however, the structural integrity of both regenerating and homeostatic spines was compromised in high pCO2 conditions. Indicators of physiological fitness (righting response, growth rate, coelomocyte concentration and composition) were not affected by increasing pCO2, but compromised spine integrity is likely to have negative consequences for defence capabilities and therefore survival of these ecologically and economically important organisms. PMID:28573022

Ocean acidification impacts spine integrity but not regenerative capacity of spines and tube feet in adult sea urchins.

Science.gov (United States)

Emerson, Chloe E; Reinardy, Helena C; Bates, Nicholas R; Bodnar, Andrea G

2017-05-01

Increasing atmospheric carbon dioxide (CO 2 ) has resulted in a change in seawater chemistry and lowering of pH, referred to as ocean acidification. Understanding how different organisms and processes respond to ocean acidification is vital to predict how marine ecosystems will be altered under future scenarios of continued environmental change. Regenerative processes involving biomineralization in marine calcifiers such as sea urchins are predicted to be especially vulnerable. In this study, the effect of ocean acidification on regeneration of external appendages (spines and tube feet) was investigated in the sea urchin Lytechinus variegatus exposed to ambient (546 µatm), intermediate (1027 µatm) and high (1841 µatm) partial pressure of CO 2 ( p CO 2 ) for eight weeks. The rate of regeneration was maintained in spines and tube feet throughout two periods of amputation and regrowth under conditions of elevated p CO 2 . Increased expression of several biomineralization-related genes indicated molecular compensatory mechanisms; however, the structural integrity of both regenerating and homeostatic spines was compromised in high p CO 2 conditions. Indicators of physiological fitness (righting response, growth rate, coelomocyte concentration and composition) were not affected by increasing p CO 2 , but compromised spine integrity is likely to have negative consequences for defence capabilities and therefore survival of these ecologically and economically important organisms.

Device for removing a spent reactor core instrument tube

International Nuclear Information System (INIS)

Watanabe, Shigeru; Tsuji, Teruaki.

1980-01-01

Purpose: To easily and exactly execute works for removing a used reactor core instrument tube to be mounted in a reactor core from the lattice space of the core or for charging the tube into the lattice of the core. Constitution: When fuel assembly is pulled out of a reactor core and a spent reactor core instrument tube is then bent and removed from the core at periodical inspection time, a lower gripping unit integral with an upper gripping unit and a bending unit is provided at the lower end of a hanging rope of a winch, and lowered to the reactor core. Then, the spent reactor core instrument tube is gripped by the upper and lower gripping units, the bending unit is operated, the spent reactor core instrument tube is bent, and the tube is then pulled upwardly by the winch to remove the tube. (Aizawa, K.)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

Fluid mechanics and heat transfer spirally fluted tubing

Science.gov (United States)

Larue, J. C.; Libby, P. A.; Yampolsky, J. S.

1981-08-01

The objective of this program is to develop both a qualitative and a quantitative understanding of the fluid mechanics and heat transfer mechanisms that underlie the measured performance of the spirally fluted tubes under development at General Atomic. The reason for the interest in the spirally fluted tubes is that results to date have indicated three advantages to this tubing concept: The fabrication technique of rolling flutes on strip and subsequently spiralling and simultaneously welding the strip to form tubing results in low fabrication costs, approximately equal to those of commercially welded tubing. The heat transfer coefficient is increased without a concomitant increase of the friction coefficient on the inside of the tube. In single-phase axial flow of water, the helical flutes continuously induce rotation of the flow both within and without the tube as a result of the effect of curvature. An increase in condensation heat transfer on the outside of the tube is achieved. In a vertical orientation with fluid condensing on the outside of the helically fluted tube, the flutes provide a channel for draining the condensed fluid.

Report of the Integrated Program Planning Activity for the DOE Fusion Energy Sciences Program

International Nuclear Information System (INIS)

None

2000-01-01

This report of the Integrated Program Planning Activity (IPPA) has been prepared in response to a recommendation by the Secretary of Energy Advisory Board that, ''Given the complex nature of the fusion effort, an integrated program planning process is an absolute necessity.'' We, therefore, undertook this activity in order to integrate the various elements of the program, to improve communication and performance accountability across the program, and to show the inter-connectedness and inter-dependency of the diverse parts of the national fusion energy sciences program. This report is based on the September 1999 Fusion Energy Sciences Advisory Committee's (FESAC) report ''Priorities and Balance within the Fusion Energy Sciences Program''. In its December 5,2000, letter to the Director of the Office of Science, the FESAC has reaffirmed the validity of the September 1999 report and stated that the IPPA presents a framework and process to guide the achievement of the 5-year goals listed in the 1999 report. The National Research Council's (NRC) Fusion Assessment Committee draft final report ''An Assessment of the Department of Energy's Office of Fusion Energy Sciences Program'', reviewing the quality of the science in the program, was made available after the IPPA report had been completed. The IPPA report is, nevertheless, consistent with the recommendations in the NRC report. In addition to program goals and the related 5-year, 10-year, and 15-year objectives, this report elaborates on the scientific issues associated with each of these objectives. The report also makes clear the relationships among the various program elements, and cites these relationships as the reason why integrated program planning is essential. In particular, while focusing on the science conducted by the program, the report addresses the important balances between the science and energy goals of the program, between the MFE and IFE approaches, and between the domestic and international aspects

CACHE: an extended BASIC program which computes the performance of shell and tube heat exchangers

International Nuclear Information System (INIS)

Tallackson, J.R.

1976-03-01

An extended BASIC program, CACHE, has been written to calculate steady state heat exchange rates in the core auxiliary heat exchangers, (CAHE), designed to remove afterheat from High-Temperature Gas-Cooled Reactors (HTGR). Computationally, these are unbaffled counterflow shell and tube heat exchangers. The computational method is straightforward. The exchanger is subdivided into a user-selected number of lengthwise segments; heat exchange in each segment is calculated in sequence and summed. The program takes the temperature dependencies of all thermal conductivities, viscosities and heat capacities into account providing these are expressed algebraically. CACHE is easily adapted to compute steady state heat exchange rates in any unbaffled counterflow exchanger. As now used, CACHE calculates heat removal by liquid weight from high-temperature helium and helium mixed with nitrogen, oxygen and carbon monoxide. A second program, FULTN, is described. FULTN computes the geometrical parameters required as input to CACHE. As reported herein, FULTN computes the internal dimensions of the Fulton Station CAHE. The two programs are chained to operate as one. Complete user information is supplied. The basic equations, variable lists, annotated program lists, and sample outputs with explanatory notes are included

Integrated straight - through automatic non-destructive examination and data acquisition system for thin-wall tubes

International Nuclear Information System (INIS)

Stoessel, A.; Boulanger, G.; Furlan, J.; Mogavero, R.

1981-09-01

This non-destructive testing unit inspects the cladding tubes for the SUPER-PHENIX fast neutron reactor. The quality level demanded for these tubes, as well as their number, required designing an installation that combined high performance with a great testing rate and complete automation. The testing is effected under immersion by means of six transducers, focused in line, working at 30 MHz. The tubes are numbered on an automatic rig; marking is by dark rings obtained by superficial electrolysis of the tube and regularly distributed on the abscissa; the quality of the tube is not affected by this. The advantage of this numbering system is that it enables the tubes to be fed to the test set in any order. An acquisition unit, constituted of a microprocessor, a semi-graphical printer and a double floppy disk unit, makes it possible to enter, edit and store the information for each tube [fr

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

Directory of Open Access Journals (Sweden)

Feng Hong

2015-01-01

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

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

Science.gov (United States)

Hong, Feng; Wei, Bin; Chen, Lin

2015-01-01

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

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

Science.gov (United States)

Wei, Bin; Chen, Lin

2015-01-01

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

76 FR 34541 - Child and Adult Care Food Program Improving Management and Program Integrity

Science.gov (United States)

2011-06-13

... 7 CFR Parts 210, 215, 220 et al. Child and Adult Care Food Program Improving Management and Program..., 220, 225, and 226 RIN 0584-AC24 Child and Adult Care Food Program Improving Management and Program... management and integrity in the Child and Adult Care Food Program (CACFP), at 67 FR 43447 (June 27, 2002) and...

Instruments for non-destructive evaluation of advanced test reactor inpile tubes

International Nuclear Information System (INIS)

Livingston, R.A.; Beller, L.S.; Edgett, S.M.

1986-01-01

The Advanced Test Reactor is a 250 MW LWR used primarily for irradiation testing of materials contained in inpile tubes that pass through the reactor core. These tubes provided the high pressure and temperature water environment required for the test specimens. The reactor cooling water surrounding the inpile tubes is at much lower pressure and temperature. The structural integrity of the inpile tubes is monitored by routine surveillance to ensure against unplanned reactor shutdowns to replace defective inpile tubes. The improved instruments developed for inpile tube surveillance include a bore profilometer, ultrasonic flaw detetion system and bore diameter gauges. The design and function of these improved instruments is presented

YouTube and Academic Libraries: Building a Digital Collection

Science.gov (United States)

Cho, Allan

2013-01-01

Although still a relatively new technology with less than 10 years of history, YouTube's extensive reach and integration in mainstream society as well as lifelong learning habits of online users cannot be understated. This article examines how the YouTube collection at the University of British Columbia Library's Irving K. Barber Learning Centre…

Mixed Waste Integrated Program Quality Assurance requirements plan

International Nuclear Information System (INIS)

1994-01-01

Mixed Waste Integrated Program (MWIP) is sponsored by the US Department of Energy (DOE), Office of Technology Development, Waste Management Division. The strategic objectives of MWIP are defined in the Mixed Waste Integrated Program Strategic Plan, and expanded upon in the MWIP Program Management Plan. This MWIP Quality Assurance Requirement Plan (QARP) applies to mixed waste treatment technologies involving both hazardous and radioactive constituents. As a DOE organization, MWIP is required to develop, implement, and maintain a written Quality Assurance Program in accordance with DOE Order 4700.1 Project Management System, DOE Order 5700.6C, Quality Assurance, DOE Order 5820.2A Radioactive Waste Management, ASME NQA-1 Quality Assurance Program Requirements for Nuclear Facilities and ANSI/ASQC E4-19xx Specifications and Guidelines for Quality Systems for Environmental Data Collection and Environmental Technology Programs. The purpose of the MWIP QA program is to establish controls which address the requirements in 5700.6C, with the intent to minimize risks and potential environmental impacts; and to maximize environmental protection, health, safety, reliability, and performance in all program activities. QA program controls are established to assure that each participating organization conducts its activities in a manner consistent with risks posed by those activities

Mixed Waste Integrated Program Quality Assurance requirements plan

Energy Technology Data Exchange (ETDEWEB)

1994-04-15

Mixed Waste Integrated Program (MWIP) is sponsored by the US Department of Energy (DOE), Office of Technology Development, Waste Management Division. The strategic objectives of MWIP are defined in the Mixed Waste Integrated Program Strategic Plan, and expanded upon in the MWIP Program Management Plan. This MWIP Quality Assurance Requirement Plan (QARP) applies to mixed waste treatment technologies involving both hazardous and radioactive constituents. As a DOE organization, MWIP is required to develop, implement, and maintain a written Quality Assurance Program in accordance with DOE Order 4700.1 Project Management System, DOE Order 5700.6C, Quality Assurance, DOE Order 5820.2A Radioactive Waste Management, ASME NQA-1 Quality Assurance Program Requirements for Nuclear Facilities and ANSI/ASQC E4-19xx Specifications and Guidelines for Quality Systems for Environmental Data Collection and Environmental Technology Programs. The purpose of the MWIP QA program is to establish controls which address the requirements in 5700.6C, with the intent to minimize risks and potential environmental impacts; and to maximize environmental protection, health, safety, reliability, and performance in all program activities. QA program controls are established to assure that each participating organization conducts its activities in a manner consistent with risks posed by those activities.

Cotton transformation via pollen tube pathway.

Science.gov (United States)

Wang, Min; Zhang, Baohong; Wang, Qinglian

2013-01-01

Although many gene transfer methods have been employed for successfully obtaining transgenic cotton, the major constraint in cotton improvement is the limitation of genotype because the majority of transgenic methods require plant regeneration from a single transformed cell which is limited by cotton tissue culture. Comparing with other plant species, it is difficult to induce plant regeneration from cotton; currently, only a limited number of cotton cultivars can be cultured for obtaining regenerated plants. Thus, development of a simple and genotype-independent genetic transformation method is particularly important for cotton community. In this chapter, we present a simple, cost-efficient, and genotype-independent cotton transformation method-pollen tube pathway-mediated transformation. This method uses pollen tube pathway to deliver transgene into cotton embryo sacs and then insert foreign genes into cotton genome. There are three major steps for pollen tube pathway-mediated genetic transformation, which include injection of -foreign genes into pollen tube, integration of foreign genes into plant genome, and selection of transgenic plants.

Integrated Pest Management: A Curriculum for Early Care and Education Programs

Science.gov (United States)

California Childcare Health Program, 2011

2011-01-01

This "Integrated Pest Management Toolkit for Early Care and Education Programs" presents practical information about using integrated pest management (IPM) to prevent and manage pest problems in early care and education programs. This curriculum will help people in early care and education programs learn how to keep pests out of early…

Integrated Healthcare Delivery: A Qualitative Research Approach to Identifying and Harmonizing Perspectives of Integrated Neglected Tropical Disease Programs.

Directory of Open Access Journals (Sweden)

Arianna Rubin Means

2016-10-01

Full Text Available While some evidence supports the beneficial effects of integrating neglected tropical disease (NTD programs to optimize coverage and reduce costs, there is minimal information regarding when or how to effectively operationalize program integration. The lack of systematic analyses of integration experiences and of integration processes may act as an impediment to achieving more effective NTD programming. We aimed to learn about the experiences of NTD stakeholders and their perceptions of integration.We evaluated differences in the definitions, roles, perceived effectiveness, and implementation experiences of integrated NTD programs among a variety of NTD stakeholder groups, including multilateral organizations, funding partners, implementation partners, national Ministry of Health (MOH teams, district MOH teams, volunteer rural health workers, and community members participating in NTD campaigns. Semi-structured key informant interviews were conducted. Coding of themes involved a mix of applying in-vivo open coding and a priori thematic coding from a start list.In total, 41 interviews were conducted. Salient themes varied by stakeholder, however dominant themes on integration included: significant variations in definitions, differential effectiveness of specific integrated NTD activities, community member perceptions of NTD programs, the influence of funders, perceived facilitators, perceived barriers, and the effects of integration on health system strength. In general, stakeholder groups provided unique perspectives, rather than contrarian points of view, on the same topics. The stakeholders identified more advantages to integration than disadvantages, however there are a number of both unique facilitators and challenges to integration from the perspective of each stakeholder group.Qualitative data suggest several structural, process, and technical opportunities that could be addressed to promote more effective and efficient integrated NTD

Advances by the Integral Fast Reactor Program

International Nuclear Information System (INIS)

Lineberry, M.J.; Pedersen, D.R.; Walters, L.C.; Cahalan, J.E.

1991-01-01

The advances by the Integral Fast Reactor Program at Argonne National Laboratory are the subject of this paper. The Integral Fast Reactor (IFR) is an advanced liquid-metal-cooled reactor concept being developed at Argonne National Laboratory. The advances stressed in the paper include fuel irradiation performance, improved passive safety, and the development of a prototype fuel cycle facility. 14 refs

Exergetic optimization of shell and tube heat exchangers using a genetic based algorithm

Energy Technology Data Exchange (ETDEWEB)

Oezcelik, Yavuz [Ege University, Bornova, Izmir (Turkey). Engineering Faculty, Chemical Engineering Department

2007-08-15

In the computer-based optimization, many thousands of alternative shell and tube heat exchangers may be examined by varying the high number of exchanger parameters such as tube length, tube outer diameter, pitch size, layout angle, baffle space ratio, number of tube side passes. In the present study, a genetic based algorithm was developed, programmed, and applied to estimate the optimum values of discrete and continuous variables of the MINLP (mixed integer nonlinear programming) test problems. The results of the test problems show that the genetic based algorithm programmed can estimate the acceptable values of continuous variables and optimum values of integer variables. Finally the genetic based algorithm was extended to make parametric studies and to find optimum configuration of heat exchangers by minimizing the sum of the annual capital cost and exergetic cost of the shell and tube heat exchangers. The results of the example problems show that the proposed algorithm is applicable to find optimum and near optimum alternatives of the shell and tube heat exchanger configurations. (author)

Digital Radiography Qualification of Tube Welding

Science.gov (United States)

Carl, Chad

2012-01-01

The Orion Project will be directing Lockheed Martin to perform orbital arc welding on commodities metallic tubing as part of the Multi Purpose Crew Vehicle assembly and integration process in the Operations and Checkout High bay at Kennedy Space Center. The current method of nondestructive evaluation is utilizing traditional film based x-rays. Due to the high number of welds that are necessary to join the commodities tubing (approx 470), a more efficient and expeditious method of nondestructive evaluation is desired. Digital radiography will be qualified as part of a broader NNWG project scope.

Point-of-care detection and real-time monitoring of intravenously delivered drugs via tubing with an integrated SERS sensor.

Science.gov (United States)

Wu, Hsin-Yu; Cunningham, Brian T

2014-05-21

We demonstrate an approach for detection, identification, and kinetic monitoring of drugs flowing within tubing, through the use of a plasmonic nanodome array (PNA) surface. The PNA structures are fabricated using a low-cost nanoreplica molding process upon a flexible plastic substrate that is subsequently integrated with a flow cell that connects in series with ordinary intravenous (IV) drug delivery tubing. To investigate the potential clinical applications for point-of-care detection and real-time monitoring, we perform SERS detection of ten pharmaceutical compounds (hydrocodone, levorphanol, morphine, oxycodone, methadone, phenobarbital, dopamine, diltiazem, promethazine, and mitoxantrone). We demonstrate dose-dependent SERS signal magnitude, resulting in detection limits (ng ml(-1)) well below typical administered dosages (mg ml(-1)). Further, we show that the detected drugs are not permanently attached to the PNA surface, and thus our approach is capable of performing continuous monitoring of drug delivery as materials flow through IV tubing that is connected in series with the sensor. Finally, we demonstrate the potential co-detection of multiple drugs when they are mixed together, and show excellent reproducibility and stability of SERS measurements for periods extending at least five days. The capabilities reported here demonstrate the potential to use PNA SERS surfaces for enhancing the safety of IV drug delivery.

Development of Zirconium alloys (for pressure tubes)

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

1997-09-01

The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

75 FR 34805 - Program Integrity Issues

Science.gov (United States)

2010-06-18

... Mathematics Access to Retain Talent Grant (National Smart Grant) Programs. DATES: We must receive your... Association of College and University Business Officers, representing business officers. Val Meyers, Michigan... identifying and handling test score abnormalities, ensuring the integrity of the testing environment, and...

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

Clinical capabilities of graduates of an outcomes-based integrated medical program

Directory of Open Access Journals (Sweden)

Scicluna Helen A

2012-06-01

Full Text Available Abstract Background The University of New South Wales (UNSW Faculty of Medicine replaced its old content-based curriculum with an innovative new 6-year undergraduate entry outcomes-based integrated program in 2004. This paper is an initial evaluation of the perceived and assessed clinical capabilities of recent graduates of the new outcomes-based integrated medical program compared to benchmarks from traditional content-based or process-based programs. Method Self-perceived capability in a range of clinical tasks and assessment of medical education as preparation for hospital practice were evaluated in recent graduates after 3 months working as junior doctors. Responses of the 2009 graduates of the UNSW’s new outcomes-based integrated medical education program were compared to those of the 2007 graduates of UNSW’s previous content-based program, to published data from other Australian medical schools, and to hospital-based supervisor evaluations of their clinical competence. Results Three months into internship, graduates from UNSW’s new outcomes-based integrated program rated themselves to have good clinical and procedural skills, with ratings that indicated significantly greater capability than graduates of the previous UNSW content-based program. New program graduates rated themselves significantly more prepared for hospital practice in the confidence (reflective practice, prevention (social aspects of health, interpersonal skills (communication, and collaboration (teamwork subscales than old program students, and significantly better or equivalent to published benchmarks of graduates from other Australian medical schools. Clinical supervisors rated new program graduates highly capable for teamwork, reflective practice and communication. Conclusions Medical students from an outcomes-based integrated program graduate with excellent self-rated and supervisor-evaluated capabilities in a range of clinically-relevant outcomes. The program

Temperature and thermal stress analysis of a switching tube anode

International Nuclear Information System (INIS)

Sutton, S.B.

1979-01-01

In the design of high power density switching tubes which are subjected to cyclic thermal loads, the temperature induced stresses must be minimized in order to maximize the life expectancy of the tube. Following are details of an analysis performed for the Magnetic Fusion Program at the Lawrence Livermore Laboratory on a proposed tube. The tube configuration is given. The problem was simplified to one-dimensional approximations for both the thermal and stress analyses. The underlying assumptions and their implications are discussed

Česká YouTube scéna

OpenAIRE

NEMRAVOVÁ, Barbora

2017-01-01

The aim of this bachelor thesis is to introduce Czech YouTubers, the internet celebrities, to the readers. It is focused on marketing strategies by which it is possible to earn money on YouTube. The thesis studies methods like collaborations with companies, sponsoring and YouTube Partner Program in more details. There is a brief history of YouTube mentioned in this thesis and how the webpage works for the users. Also there are described types of videos and types of YouTubers divided by orient...

Plugging criteria for WWER SG tubes

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Plugging criteria for WWER SG tubes

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

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

Measurement of unsteady flow forces in inline and staggered tube bundles with fixed and vibrating tubes

International Nuclear Information System (INIS)

Michel, A.; Heinecke, E.; Decken, C.B. von der.

1986-01-01

Unsteady flow forces arising in heat exchangers with cross-flow may lead to serious vibrations of the tubes. These vibrations can destroy the tubes in the end supports or in the baffles, which would require expensive repairs. The flow forces reach unexpectedly by high values if the vibration of the tube intensifies these forces. To clear up this coupling mechanism the flow forces and the vibration amplitude were measured simultaneously in a staggered and in an inline tube bundle. Considering the tube as a one-mass oscillator excited by the flow force, the main parameters can be derived, i.e. dynamic pressure, reduced mass, eigenfrequency and damping. These parameters form a dimensionless model number describing the coherence of the vibration amplitude and the force coefficient. The validity of this number has been confirmed by varying the test conditions. With the aid of this model number, the expected force coefficient can be calculated and then using a finite-element program information can be obtained about mechanical tensions and the lifetime of the heat exchanger tubes. With this model number the results of other authors, who measured the vibration amplitude only, could be confirmed in good agreement. The experiments were carried out in air with Reynolds numbers 10 4 5 . (orig.) [de

Pressure distribution over tube surfaces of tube bundle subjected to two phase cross flow

International Nuclear Information System (INIS)

Sim, Woo Gun

2013-01-01

Two phase vapor liquid flows exist in many shell and tube heat exchangers such as condensers, evaporators and nuclear steam generators. To understand the fluid dynamic forces acting on a structure subjected to a two phase flow, it is essential to obtain detailed information about the characteristics of a two phase flow. The characteristics of a two phase flow and the flow parameters were introduced, and then, an experiment was performed to evaluate the pressure loss in the tube bundles and the fluid dynamic force acting on the cylinder owing to the pressure distribution. A two phase flow was pre mixed at the entrance of the test section, and the experiments were undertaken using a normal triangular array of cylinders subjected to a two phase cross flow. The pressure loss along the flow direction in the tube bundles was measured to calculate the two phase friction multiplier, and the multiplier was compared with the analytical value. Furthermore, the circular distributions of the pressure on the cylinders were measured. Based on the distribution and the fundamental theory of two phase flow, the effects of the void fraction and mass flux per unit area on the pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure on the tube by a numerical method. It was found that for low mass fluxes, the measured two phase friction multipliers agree well with the analytical results, and good agreement for the effect of the void fraction on the drag coefficients, as calculated by the measured pressure distributions, is shown qualitatively, as compared to the existing experimental results

Determining Interest in YouTube Topics for Extension-Authored Video Development

Science.gov (United States)

Parish, Jane A.; Karisch, Brandi B.

2013-01-01

With an audience of over 1 billion users per month, YouTube is an attractive medium for delivering Extension programming. Amidst growing competition for viewership, determining content that is in demand by Extension clientele on YouTube is a daunting challenge that Extension educators face. The YouTube Search function of Google Trends and…

High-performance vacuum tubes for more energy efficiency. Building-integrated CPC vacuum tube collectors unite several functions.; Hochleistungs-Vakuumroehren fuer mehr Energieeffizienz. Gebaeudeintegrierte CPC-Vakuumroehren-Kollektoren vereinen mehrere Funktionen

Energy Technology Data Exchange (ETDEWEB)

Theiss, Eric

2013-10-15

The performance of solar collectors primarily contributes to increased efficiency and reduced operating costs of solar thermal systems. With the use of building-integrated CPC vacuum tube collectors an extremely high energy yield is achieved on a smaller collector gross area. As a building-integrated system solution the CPC facade provide panels in addition to its use as spandrel panels within the glazed buildings not only an architectural design element, but unite as a multifunctional component for several functions. [German] Die Leistungsfaehigkeit der Solarkollektoren traegt primaer zur Effizienzsteigerung und Reduzierung der Betriebskosten einer Solarthermieanlagen bei. Mit dem Einsatz gebaeudeintegrierter CPC-Vakuumroehrenkollektoren wird auf einer kleineren Kollektorbruttoflaeche ein extrem hoher Energieertrag erreicht. Als gebaeudeintegrierte Systemloesung bieten die CPC-Fassadenkollektoren neben dem Einsatz als Bruestungselemente auch innerhalb der verglasten Gebaeuden nicht nur ein architektonisches Gestaltungselement, sondern vereinen als multifunktionaler Bestandteil noch mehrere Funktionen.

Falling film flow, heat transfer and breakdown on horizontal tubes

International Nuclear Information System (INIS)

Rogers, J.T.

1980-11-01

Knowledge of falling film flow and heat transfer characteristics on horizontal tubes is required in the assessment of certain CANDU reactor accident sequences for those CANDU reactors which use moderator dump as one of the shut-down mechanisms. In these reactors, subsequent cooling of the calandria tubes is provided by falling films produced by sprays. This report describes studies of falling film flow and heat transfer characteristics on horizontal tubes. Analyses using integral methods are given for laminar and turbulent flow, ignoring and accounting for momentum effects in the film. Preliminary experiments on film flow stability on horizontal tubes are described and various mechanisms of film breakdown are examined. The work described in this report shows that in LOCA with indefinitely delayed ECI in the NPD or Douglas Point (at 70 percent power) reactors, the falling films on the calandria tubes will not be disrupted by any of the mechanisms considered, provided that the pressure tubes do not sag onto the calandria tubes. However, should the pressure tubes sag onto the calandria tubes, film disruption will probably occur

Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

Energy Technology Data Exchange (ETDEWEB)

Jeong, Y. M.; Kim, Y. S.; Im, K. S.; Kim, K. S.; Ahn, S. B

2007-06-15

Zr-2.5Nb pressure tubes are one of the most critical structural components governing the lifetime of the heavy water reactors to carry fuel bundles and heavy coolant water inside. Since they are being degraded during their operation in reactors due to dimensional changes caused by creep and irradiation growth, neutron irradiation and delayed hydride cracking, it is required to evaluate their degradation by conducting material testing and examinations on the highly irradiated pressure tubes in hot cells and to keep tracking of their degradation behavior with operation time, which are the aim of this project.

Improving the calandria tubes for CANDU reactors

International Nuclear Information System (INIS)

Coleman, C.E.; Fong, R.W.L.; Doubt, G.L.

1997-01-01

CANDU calandria tubes are made from annealed Zircaloy-2 sheet formed into a cylinder and welded along its length to make the tube. The current calandria tubes have given exemplary service for many years. With more stringent regulations and the need to accommodate warm cooling water in tropical countries, we started a development program to increase the margins for failure during postulated accidents. These improvements involve increasing the tube strength and optimising the heat-transfer from an excessively hot fuel channel to the cool moderator. If the postulated accident involves a pressure tube break, it would be desirable if the calandria tube withstood the full pressure of the heat-transport system. The weakest link in current calandria tubes is the weld. Thickening the weld can increase the strength by 20% while seamless tubes can be 45% stronger than current tubes. The latter tubes can hold full system pressure for many hours without failure. If during the postulated accident the fuel and pressure tube become excessively hot but do not touch the calandria tube, the radiant heat loss must be maximised. Current calandria tubes have an absorptivity (emissivity) of about 0.2. To protect the fuel and the fuel channel we have devised a finish to the inside surface of the calandria tube that increases the emissivity to 0.7. If during the postulated accident the hot pressure tube touches the cool calandria tube, the contact conductance and the critical heat flux must be optimised to ensure nucleate boiling of the moderator at the outside surface of the calandria tube and therefore efficient exploitation of the moderator as a heat sink. In laboratory tests small ridges on the inside surface and roughening of the outside surface have been shown to increase the margins against failure and increase the possible moderator temperatures thus providing the opportunity to decrease the cost of the moderator heat-exchange system and remove restrictions on reactor operation in

Thermal optimization of primary side in double-tube OTSG

International Nuclear Information System (INIS)

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

2011-01-01

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

Measurement and computation for sag of calandria tube due to irradiation creep in PHWR

International Nuclear Information System (INIS)

Son, S. M.; Lee, W. R.; Lee, S. K.; Lee, J. S.; Kim, T. R.; Na, B. K.; Namgung I.

2003-01-01

Calandria tubes and Liquid Injection Shutdown System(LISS) tubes in a Pressurized Heavy Water Reactor(PHWR) are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with LISS tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measuring probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sags of both tubes in the PHWR. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

Thermodynamic analysis of a pulse tube engine

International Nuclear Information System (INIS)

Moldenhauer, Stefan; Thess, André; Holtmann, Christoph; Fernández-Aballí, Carlos

2013-01-01

Highlights: ► Numerical model of the pulse tube engine process. ► Proof that the heat transfer in the pulse tube is out of phase with the gas velocity. ► Proof that a free piston operation is possible. ► Clarifying the thermodynamic working principle of the pulse tube engine. ► Studying the influence of design parameters on the engine performance. - Abstract: The pulse tube engine is an innovative simple heat engine based on the pulse tube process used in cryogenic cooling applications. The working principle involves the conversion of applied heat energy into mechanical power, thereby enabling it to be used for electrical power generation. Furthermore, this device offers an opportunity for its wide use in energy harvesting and waste heat recovery. A numerical model has been developed to study the thermodynamic cycle and thereby help to design an experimental engine. Using the object-oriented modeling language Modelica, the engine was divided into components on which the conservation equations for mass, momentum and energy were applied. These components were linked via exchanged mass and enthalpy. The resulting differential equations for the thermodynamic properties were integrated numerically. The model was validated using the measured performance of a pulse tube engine. The transient behavior of the pulse tube engine’s underlying thermodynamic properties could be evaluated and studied under different operating conditions. The model was used to explore the pulse tube engine process and investigate the influence of design parameters.

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

Intermediate heat exchanger tube vibration induced by cross and parallel mixed flow

International Nuclear Information System (INIS)

Kawamura, Koji

1986-01-01

The characteristics of pool type LMFBR intermediate heat exchanger (IHX) tube vibrations induced by cross and parallel mixed flow were basically investigated. Secondary coolant in IHX tube bundle is mixed flow of parallel jit flow along the tube axis through flow holes in baffle plates and cross flow. By changing these two flow rate, flow distributions vary in the tube bundle. Mixed flow also induces vibrations which cause fretting wear and fatigue of tube. It is therefore very important to evaluate the tube vibration characteristics for estimating the tube integrity. The results show that the relationships between tube vibrations and flow distributions in the tube bundle were cleared, and mixed flow induced tube vibration could be evaluated on the base of the characteristics of both parallel and cross flow induced vibration. From these investigations it could be concluded that the characteristics of tube vibration for various flow distributions can be systematically evaluated. (author)

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

US/DOE Man-Machine Integration program for liquid metal reactors

International Nuclear Information System (INIS)

D'Zmura, A.P.; Seeman, S.E.

1985-03-01

The United States Department of Energy (DOE) Man-Machine Integration program was started in 1980 as an addition to the existing Liquid Metal Fast Breeder Reactor safety base technology program. The overall goal of the DOE program is to enhance the operational safety of liquid metal reactors by optimum integration of humans and machines in the overall reactor plant system and by application of the principles of human-factors engineering to the design of equipment, subsystems, facilities, operational aids, procedures and environments. In the four years since its inception the program has concentrated on understanding the control process for Liquid Metal Reactors (LMRs) and on applying advanced computer concepts to this process. This paper describes the products that have been developed in this program, present computer-related programs, and plans for the future

Advanced Ultrasupercritical (AUSC) Tube Membrane Panel Development

Energy Technology Data Exchange (ETDEWEB)

Pschirer, James [Alstom Power Inc., Windsor, CT (United States); Burgess, Joshua [Alstom Power Inc., Windsor, CT (United States); Schrecengost, Robert [Alstom Power Inc., Windsor, CT (United States)

2017-08-16

Alstom Power Inc., a wholly owned subsidiary of the General Electric Company (GE), has completed the project “Advanced Ultrasupercritical (AUSC) Tube Membrane Panel Development” under U.S. Department of Energy (DOE) Award Number DE-FE0024076. This project was part of DOE’s Novel Crosscutting Research and Development to Support Advanced Energy Systems program. AUSC Tube Membrane Panel Development was a two and one half year project to develop and verify the manufacturability and serviceability of welded tube membrane panels made from high performance materials suitable for the AUSC steam cycles, defined as high pressure steam turbine inlet conditions of 700-760°C (1292-1400°F) and 24.5-35MPa (3500-5000psi). The difficulty of this challenge lies in the fact that the membrane-welded construction imposes demands on the materials that are unlike any that exist in other parts of the boiler. Tube membrane panels have been designed, fabricated, and installed in boilers for over 50 years with relatively favorable experience when fabricated from carbon and Cr-Mo low alloy steels. The AUSC steam cycle requires membrane tube panels fabricated from materials that have not been used in a weldment with metal temperatures in the range of 582-610°C (1080-1130°F). Fabrication materials chosen for the tubing were Grade 92 and HR6W. Grade 92 is a creep strength enhanced ferritic Cr-Mo alloy and HR6W is a high nickel alloy. Once the materials were chosen, GE performed the engineering design of the panels, prepared shop manufacturing drawings, and developed manufacturing and inspection plans. After the materials were purchased, GE manufactured and inspected the tube membrane panels, determined if post fabrication heat treatment of the tube membrane panels was needed, performed pre- and post-weld heat treatment on the Grade 92 panels, conducted final nondestructive inspection of any heat treated tube membrane panels, conducted destructive inspection of the completed tube

Ageing management of AG3NET beam tubes in ORPHEE Research

International Nuclear Information System (INIS)

Florence, Gupta; Maud, Corbel

2013-01-01

The materials used in research reactors come from the best compromise between research needs and safety issues such as integrity of equipment during their whole life. For example, aluminium alloys such as AG3NET are interesting for research reactors dedicated to the production of neutron flux since they are transparent to neutrons but they become fragile under irradiation. Therefore the evolution of material's mechanical properties under irradiation is a topic of interest for research reactors safety and operators must implement an ageing management program of equipment subject to irradiation. This kind of aluminium alloys compound is used in many French research reactors like the Jules Horowitz reactor (JHR) and ORPHEE reactor operated by the Atomic Energy and Alternative Energies Commission (CEA) or the high flux reactor (HFR) operated by the Laue-Langevin Institute (ILL). Particularly, in the ORPHEE reactor, AG3NET is used for beam tubes, located in the heavy water tank surrounding the core, which guide neutrons towards experimental stations. The failure of a beam tube in ORPHEE reactor can lead to a reactivity insertion in the core, whose effects can be managed by the control rods system. Nevertheless, to control the effects of ageing on such equipment, the operator plans to replace the beam tubes on the basis of a criterion he defined. For the ORPHEE's second periodic safety review, the operator has re-evaluated the situation of the beam tubes with regard of this criterion and has established a beam tube replacement schedule. The 'Institut de Radioprotection et de Surete Nucleaire' (IRSN), as a technical support of the French nuclear safety authority, assessed the elements presented by the CEA for this periodic safety review and concluded that the replacement criterion used for these equipment lead to reach a fragile behaviour of the materials. Thus, the breaking of several beam tubes can't be excluded but this situation can leads to severe consequences on the

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)

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

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

Developing an integrated dam safety program

International Nuclear Information System (INIS)

Nielsen, N. M.; Lampa, J.

1996-01-01

An effort has been made to demonstrate that dam safety is an integral part of asset management which, when properly done, ensures that all objectives relating to safety and compliance, profitability, stakeholders' expectations and customer satisfaction, are achieved. The means to achieving this integration of the dam safety program and the level of effort required for each core function have been identified using the risk management approach to pinpoint vulnerabilities, and subsequently to focus priorities. The process is considered appropriate for any combination of numbers, sizes and uses of dams, and is designed to prevent exposure to unacceptable risks. 5 refs., 1 tab

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

Eddy current technology for heat exchanger and steam generator tube inspection

Energy Technology Data Exchange (ETDEWEB)

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

2004-07-01

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

Eddy current technology for heat exchanger and steam generator tube inspection

International Nuclear Information System (INIS)

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

2004-01-01

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

Surveillance test of OWL-2 inpile tube

International Nuclear Information System (INIS)

Shimizu, Masatsugu; Itoh, Noboru

1976-08-01

A series of irradiation surveillance tests performed in integrity evaluation of an inpile tube for the test loop OWL-2 are described. Specimens were exposed to the neutron fluences from 1 x 10 20 to 3.4 x 10 21 n/cm 2 (>1 MeV), and subjected to post-irradiation tensile test at room temperature and service temperature 285 0 C. The strength increased and the ductility decreased with increasing neutron fluence. The reduction in fracture ductility due to neutron irradiation in the fluence range was insignificant, and the elongation of 33% was retained even for the maximum neutron fluence at 285 0 C. Little decrease of the ductility with fluence indicates that the tube would be in service for long time, ie to the integral fluence of 3.4 x 10 21 n/cm 2 . (auth.)

Integrated data base program

International Nuclear Information System (INIS)

Notz, K.J.

1981-01-01

The IDB Program provides direct support to the DOE Nuclear Waste Management and Fuel Cycle Programs and their lead sites and support contractors by providing and maintaining a current, integrated data base of spent fuel and radioactive waste inventories and projections. All major waste types (HLW, TRU, and LLW) and sources (government, commerical fuel cycle, and I/I) are included. A major data compilation was issued in September, 1981: Spent Fuel and Radioactive Waste Inventories and Projections as of December 31, 1980, DOE/NE-0017. This report includes chapters on Spent Fuel, HLW, TRU Waste, LLW, Remedial Action Waste, Active Uranium Mill Tailings, and Airborne Waste, plus Appendices with more detailed data in selected areas such as isotopics, radioactivity, thermal power, projections, and land usage. The LLW sections include volumes, radioactivity, thermal power, current inventories, projected inventories and characteristics, source terms, land requirements, and a breakdown in terms of government/commercial and defense/fuel cycle/I and I

Integrating the GalileoScope into Successful Outreach Programming

Science.gov (United States)

Michaud, Peter D.; Slater, S.; Goldstein, J.; Harvey, J.; Garcia, A.

2010-01-01

Since 2004, the Gemini Observatory’s week-long Journey Through the Universe (JTtU) program has successfully shared the excitement of scientific research with teachers, students and the public on Hawaii’s Big Island. Based on the national JTtU program started in 1999, the Hawai‘i version reaches an average of 7,000 students annually and each year features a different theme shared with a diverse set of learners. In 2010, the theme includes the integration of the GalileoScope-produced as a keystone project for the International Year of Astronomy. In preparation, a pilot teacher workshop (held in October 2009) introduced local island teachers to the GalileoScope and a 128-page educator’s activity resource book coordinated by the University of Wyoming. Response from this initial teacher’s workshop has been strong and evaluations plus follow-up actions by participating teachers illustrate that the integration of the GalileoScope has been successful based upon this diverse sample. Integrating GalileoScopes into Chilean schools in 2010 is also underway at Gemini South. This program will solicit informal proposals from educators who wish to use the telescopes in classrooms and a Spanish version of the teacher resource book is planned. The authors conclude that integration of the GalileoScope into an existing outreach program is an effective way to keep content fresh, relevant and engaging for both educators and students. This initiative is funded by Gemini Observatory outreach program. The Gemini Observatory is operated by the Association of Universities for Research in Astronomy, Inc., under a cooperative agreement with the NSF on behalf of the Gemini partnership: the National Science Foundation (US), the Science and Technology Facilities Council (UK), the National Research Council (Canada), CONICYT (Chile), the Australian Research Council (Australia), Ministério da Ciência e Tecnologia (Brazil), and Ministerio de Ciencia, Tecnología e Innovación Productiva

21 CFR 868.5800 - Tracheostomy tube and tube cuff.

Science.gov (United States)

2010-04-01

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

Attitudes Toward Integration as Perceived by Preservice Teachers Enrolled in an Integrated Mathematics, Science, and Technology Teacher Education Program.

Science.gov (United States)

Berlin, Donna F.; White, Arthur L.

2002-01-01

Describes the purpose of the Master of Education (M. Ed.) Program in Integrated Mathematics, Science, and Technology Education (MSAT Program) at The Ohio State University and discusses preservice teachers' attitudes and perceptions toward integrated curriculum. (Contains 35 references.) (YDS)

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

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

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

Ultrasonic inspection of tube to tube plate welds

International Nuclear Information System (INIS)

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

1985-01-01

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

Hazardous Waste Remedial Actions Program: integrating waste management

International Nuclear Information System (INIS)

Petty, J.L.; Sharples, F.E.

1986-01-01

The Hazardous Waste Remedial Actions Program was established to integrate Defense Programs' activities in hazardous and mixed waste management. The Program currently provides centralized planning and technical support to the Office of the Assistant Secretary for Defense Programs. More direct project management responsibilities may be assumed in the future. The Program, under the direction of the ASDP's Office of Defense Waste and Transportation Management, interacts with numerous organizational entities of the Department. The Oak Ridge Operations Office has been designated as the Lead Field Office. The Program's four current components cover remedial action project identification and prioritization; technology adaptation; an informative system; and a strategy study for long-term, ''corporate'' project and facility planning

The Environment for Application Software Integration and Execution (EASIE), version 1.0. Volume 2: Program integration guide

Science.gov (United States)

Jones, Kennie H.; Randall, Donald P.; Stallcup, Scott S.; Rowell, Lawrence F.

1988-01-01

The Environment for Application Software Integration and Execution, EASIE, provides a methodology and a set of software utility programs to ease the task of coordinating engineering design and analysis codes. EASIE was designed to meet the needs of conceptual design engineers that face the task of integrating many stand-alone engineering analysis programs. Using EASIE, programs are integrated through a relational data base management system. In volume 2, the use of a SYSTEM LIBRARY PROCESSOR is used to construct a DATA DICTIONARY describing all relations defined in the data base, and a TEMPLATE LIBRARY. A TEMPLATE is a description of all subsets of relations (including conditional selection criteria and sorting specifications) to be accessed as input or output for a given application. Together, these form the SYSTEM LIBRARY which is used to automatically produce the data base schema, FORTRAN subroutines to retrieve/store data from/to the data base, and instructions to a generic REVIEWER program providing review/modification of data for a given template. Automation of these functions eliminates much of the tedious, error prone work required by the usual approach to data base integration.

Integral quality programs for radiodiagnostics Services

International Nuclear Information System (INIS)

Alastuey, F.; Barranco, C.; Marco, R.; Perez, C.; Sanchez, J.; Pardo, J.; Madrid, G.

1993-01-01

The aim of the work entitled ''Integral Quality Programs for Radiodiagnostics Services'' is to present the experience accumulated over the past 10 years by the Radiodiagnostics Service of C.M.E. Ramon y Cajal in Zaragoza. The term ''integral quality'' will be defined conceptually in order to differentiate it from the classical quality control which refers exclusively to the control of radiology equipment. The problem will be reviewed from the historical point of view and a basic, homologated model, contrasted on the basis of the work of these 10 years, is proposed mainly to serve as the backbone for the working system in a Radiodiagnostics Service. (Author) 46 ref

A stochastic-programming approach to integrated asset and liability ...

African Journals Online (AJOL)

This increase in complexity has provided an impetus for the investigation into integrated asset- and liability-management frameworks that could realistically address dynamic portfolio allocation in a risk-controlled way. In this paper the authors propose a multi-stage dynamic stochastic-programming model for the integrated ...

Integrated Data Base Program: a status report

International Nuclear Information System (INIS)

Notz, K.J.; Klein, J.A.

1984-06-01

The Integrated Data Base (IDB) Program provides official Department of Energy (DOE) data on spent fuel and radioactive waste inventories, projections, and characteristics. The accomplishments of FY 1983 are summarized for three broad areas: (1) upgrading and issuing of the annual report on spent fuel and radioactive waste inventories, projections, and characteristics, including ORIGEN2 applications and a quality assurance plan; (2) creation of a summary data file in user-friendly format for use on a personal computer and enhancing user access to program data; and (3) optimizing and documentation of the data handling methodology used by the IDB Program and providing direct support to other DOE programs and sites in data handling. Plans for future work in these three areas are outlined. 23 references, 11 figures

Heat Exchanger Tube to Tube Sheet Joints Corrosion Behavior

Directory of Open Access Journals (Sweden)

M. Iancu

2013-03-01

Full Text Available Paper presents the studies made by the authors above the tube to tube sheet fittings of heat exchanger with fixed covers from hydrofining oil reforming unit. Tube fittings are critical zones for heat exchangers failures. On a device made from material tube and tube sheet at real joints dimensions were establish axial compression force and traction force at which tube is extracted from expanded joint. Were used two shapes joints with two types of fittings surfaces, one with smooth hole of tube sheet and other in which on boring surface we made a groove. From extracted expanded tube zones were made samples for corrosion tests in order to establish the corrosion rate, corrosion potential and corrosion current in working mediums such as hydrofining oil and industrial water at different temperatures. The corrosion rate values and the temperature influence are important to evaluate joints durability and also the results obtained shows that the boring tube sheet shape with a groove on hole tube shape presents a better corrosion behavior then the shape with smooth hole tube sheet.

Tube spacer grid for a heat-exchanger tube bundle

International Nuclear Information System (INIS)

Scheidl, H.

1976-01-01

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

Eddy-current inspection of ferromagnetic tubing using pulsed magnetic saturation

Energy Technology Data Exchange (ETDEWEB)

Dodd, C V; Deeds, W E

1986-07-01

A pulsed eddy-current system has been designed and developed for nondestructive evaluation of 2.25Cr-1Mo steam generator tubing from the bore side. Since the tubing is ferromagnetic, a large current pulse is sent through a driver coil to produce magnetic saturation all the way through the tube wall. A pickup coil produces an output pulse that is dependent upon the tube properties as well as the driving pulse. The output pulse heights at selected times are used as data that are computer-correlated with calibration data taken from machined standards. Performance data, circuit diagrams, and computer programs are given for the system, which has been demonstrated to detect small flaws located near the outside of a thick ferromagnetic tube.

Proposals for investigating instrument tube line breaks in pressurized water reactors

International Nuclear Information System (INIS)

Fletcher, C.D.; Charlton, T.R.; Loomis, G.G.; Hall, D.G.; Cozzuol, J.M.

1985-11-01

Questions posed by the NRC pertaining to instrument tube critical flow and applicability of the Semiscale experimental facility are evaluated. A program is recommended to investigate the issue of generic PWR safety following hypothetical rupture of instrument tubes due to consequences of seismic events

Enhancement of heat transfer. The performance of micro-fin tubes

International Nuclear Information System (INIS)

Muzzio, A.

2001-01-01

Micro-fin tubes are characterised by numerous, very small integral fins that spiral down the inner surface. A very interesting feature of their performance in flow boiling and condensation is a large heat transfer enhancement accompanied by a low pressure drop penalty. This paper presents a general overview of micro-fin tubes and of their performance in evaporation, condensation and single-phase flow [it

GTSP, automatic ultrasonic inspection of Guide Tube Support Pin in nuclear power plants

International Nuclear Information System (INIS)

2008-01-01

1 - Description of program or function: GTSP Visitor is a program for automatic detection of known object's position in video frames. It is especially designed for automatic ultrasonic inspection of guide tube support pin (GTSP) in nuclear power plant. 2 - Methods: A GTSP and its position are detected by two-step matched filter algorithm. In first step, a video frame including GTSPs are transformed by DFT. DFTed image is multiplied by matched filter, made from a guide tube image, in frequency domain for estimate Guide Tube center position. Guide Tube areas around estimated center position are erased (pixel values of image are filled with zeros). In next step, image whose guide tube areas were erased is processed as described above but using a different matched filter made from a support pin?s image. Then the positions of two GTSPs are estimated and their orientation is estimated too. Finally its position and orientations are used for control the robot toward the desired position. 3 - Restrictions on the complexity of the problem: Robot control is out of the scope of this program. OpenCV and compatible camera are necessary

Development of austenitic stainless steel tubes for nuclear reactor cladding

International Nuclear Information System (INIS)

Padilha, A.F.; Ferreira, P.I.; Andrade, P.I.; Andrade, A.H.P. de; Meyerhof, S.; Mauricio, J.

1984-01-01

In the development of thin wall tubes for nuclear reactor fuel cladding applications, a great number of activities, related to the fabrication process as the qualification are involved. A test program was envisaged to verify the quality of seam welded stainless steel tubes (AISI 304), obtained as a result of an effort by the IPEN-CNEN/SP and the brazilian industry. The relevant aspects involved in the preparation of the tubes and some preliminary test results are presented. (Author) [pt

Application of the Lion's integral to calculate heat transfer with the N2O4 turbulent flow in a tube

International Nuclear Information System (INIS)

Petrovich, V.Yu.; Tverkovkin, B.E.; Nesterenko, V.B.

1976-01-01

When carrying out engineering calculation of heat transfer in the case of turbulent flow of non-equilibrium reacting gas in a tube, it is necessary to dispose of criterion dependence to calculate Nusselt number. As a rule, dependences obtained by empirical methods are not widely adopted. It is proposed that the integral of Lion type be used for the heat transfer calculation in the form of which an expression for Nusselt number has been written under the conditions of turbulent flow with a non-equilibrium chemical reaction. On calculating turbulent fluctuations Millionshchikov two-layer model is used. A simple approximation of source-discharge of the mass of mixture components is suggested for the sake of simplification of Lion integral. The proposed theoretical dependences for the heat transfer calculation when chemical reactions are available substantially extend the field of application of Lion integral and may be used designing equipment with a chemically reacting coolant

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

Energy Technology Data Exchange (ETDEWEB)

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

1996-12-31

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

OSMOSE: An experimental program for the qualification of integral cross sections of actinides

International Nuclear Information System (INIS)

Hudelot, J. P.; Klann, R.; Fougeras, P.; Jorion, F.; Drin, N.; Donnet, L.

2004-01-01

The accurate integral cross sectional reaction rates in representative spectra for the actinides are discussed at OSMOSE program. The first step in obtaining better nuclear data consists of measuring accurate integral data and comparing it to integrated energy dependent data: this comparison provides a direct assessment of the effect of deficiencies in the differential data. The OSMOSE program includes a complete analytical program associated with experimental measurement program and aims at understanding and resolving discrepancies between calculated and measured values. The measurement covers a wide range of neutron spectra, from over-moderate thermal spectra to fast spectra. (authors)

Mechanistic modeling of heat transfer process governing pressure tube-to-calandria tube contact and fuel channel failure

International Nuclear Information System (INIS)

Luxat, J.C.

2002-01-01

Heat transfer behaviour and phenomena associated with ballooning deformation of a pressure tube into contact with a calandria tube have been analyzed and mechanistic models have been developed to describe the heat transfer and thermal-mechanical processes. These mechanistic models are applied to analyze experiments performed in various COG funded Contact Boiling Test series. Particular attention is given in the modeling to characterization of the conditions for which fuel channel failure may occur. Mechanistic models describing the governing heat transfer and thermal-mechanical processes are presented. The technical basis for characterizing parameters of the models from the general heat transfer literature is described. The validity of the models is demonstrated by comparison with experimental data. Fuel channel integrity criteria are proposed which are based upon three necessary and sequential mechanisms: Onset of CHF and local drypatch formation at contact; sustained film boiling in the post-contact period; and creep strain to failure of the calandria tube while in sustained film boiling. (author)

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

The National Shipbuilding Research Program. REAPS 5th Annual Technical Symposium Proceedings. Paper No. 19: Interactive Lines Generation (HULGEN) With a Storage Tube (User Guide)

National Research Council Canada - National Science Library

Fuller, Arthur L

1978-01-01

.... It was originally written for refresh. graphics scopes with light pens. Those earlier versions of the program, although done for light pen picks, operated in a way that made conversion to storage tube graphics very practical...

Highlights of the metallurgical behaviour of CANDU pressure tubes

International Nuclear Information System (INIS)

Price, E.G.

1984-10-01

This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

Fretting-wear damage of heat exchanger tubes: a proposed damage criterion based on tube vibration response

International Nuclear Information System (INIS)

Yetisir, M.; McKerrow, E.; Pettigrew, M.J.

1997-01-01

A simple criterion is proposed to estimate fretting-wear damage in heat exchanger tubes with clearance supports. The criterion is based on parameters such as vibration frequency, mid-span vibration amplitude, span length, tube mass and an empirical wear coefficient. It is generally accepted that fretting-wear damage is proportional to a parameter called work-rate. Work-rate is a measure of the dynamic interaction between a vibrating tube and its supports. Due to the complexity of the impact-sliding behavior at the clearance-supports, work-rate calculations for heat exchanger tubes require specialized non-linear finite element codes. These codes include contact models for various clearance-support geometries. Such non-linear finite element analyses are complex, expensive and time consuming. The proposed criterion uses the results of linear vibration analysis (i.e., vibration frequency and mid-span vibration amplitude due to turbulence) and does not require a non-linear analysis. It can be used by non-specialists for a quick evaluation of the expected work-rate, and hence, the fretting-wear damage of heat exchanger tubes. The proposed criterion was obtained from an extensive parametric study that was conducted using a non-linear finite element program. It is shown that, by using the proposed work-rate criteria, work-rate can be estimated within a factor of two. This result, however, requires further testing with more complicated flow patterns. (author)

Moderator mixing after a pressure tube failure

International Nuclear Information System (INIS)

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

1997-01-01

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

YouTube: Educational Potentials and Pitfalls

Science.gov (United States)

Jones, Troy; Cuthrell, Kristen

2011-01-01

The instructional potential of video technology in the classroom is promising, especially in light of the 21st Century Learning Framework (Siegle, 2009). Studies show positive gains in student outcomes as a result of the integration of video technology in instruction. This article explores potential uses of YouTube as an instructional aid in…

Integrating human resources and program-planning strategies.

Science.gov (United States)

Smith, J E

1989-06-01

The integration of human resources management (HRM) strategies with long-term program-planning strategies in hospital pharmacy departments is described. HRM is a behaviorally based, comprehensive strategy for the effective management and use of people that seeks to achieve coordination and integration with overall planning strategies and other managerial functions. It encompasses forecasting of staffing requirements; determining work-related factors that are strong "motivators" and thus contribute to employee productivity and job satisfaction; conducting a departmental personnel and skills inventory; employee career planning and development, including training and education programs; strategies for promotion and succession, including routes of advancement that provide alternatives to the managerial route; and recruitment and selection of new personnel to meet changing departmental needs. Increased competitiveness among hospitals and a shortage of pharmacists make it imperative that hospital pharmacy managers create strategies to attract, develop, and retain the right individuals to enable the department--and the hospital as a whole--to grow and change in response to the changing health-care environment in the United States. Pharmacy managers would be greatly aided in this mission by the establishment of a well-defined, national strategic plan for pharmacy programs and services that includes an analysis of what education and training are necessary for their successful accomplishment. Creation of links between overall program objectives and people-planning strategies will aid hospital pharmacy departments in maximizing the long-term effectiveness of their practice.

Transfer of hydrogen and helium through corrugated, flexible tubes

International Nuclear Information System (INIS)

Schippl, K.

2001-01-01

The transfer of liquid gas or cold gas through corrugated tubes is an alternative to rigid systems for the use in reactor technique. Advantages: flexibility for easy installation; these tubes together with their associated terminations and hardware are assembled, leak-tested and evacuated at the factory. This permits simple and cost saving installation on site. All tubes are helium leak-tested with a sensitivity of 10E -9 mbar 1/sec. Following the leak test, the vacuum space is pumped down to the operation vacuum level and properly sealed. The vacuum integrity is guaranteed as a result of the high degree of cleanliness observed during production and from the use of a specially selected better material inside the vacuum space. Disadvantage: pressure is limited to 20 bar. To fulfil all rules of the reactor safety, different tests have to be done. Because of the longitudinal weld of the corrugated tube, a bursting test of different sizes gives the best information of the liability of this kind of tube. It can be shown that the bursting pressure of such a tube is more than 5 times higher than the max. working pressure

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.

IMP - INTEGRATED MISSION PROGRAM

Science.gov (United States)

Dauro, V. A.

1994-01-01

IMP is a simulation language that is used to model missions around the Earth, Moon, Mars, or other planets. It has been used to model missions for the Saturn Program, Apollo Program, Space Transportation System, Space Exploration Initiative, and Space Station Freedom. IMP allows a user to control the mission being simulated through a large event/maneuver menu. Up to three spacecraft may be used: a main, a target and an observer. The simulation may begin at liftoff, suborbital, or orbital. IMP incorporates a Fehlberg seventh order, thirteen evaluation Runge-Kutta integrator with error and step-size control to numerically integrate the equations of motion. The user may choose oblate or spherical gravity for the central body (Earth, Mars, Moon or other) while a spherical model is used for the gravity of an additional perturbing body. Sun gravity and pressure and Moon gravity effects are user-selectable. Earth/Mars atmospheric effects can be included. The optimum thrust guidance parameters are calculated automatically. Events/maneuvers may involve many velocity changes, and these velocity changes may be impulsive or of finite duration. Aerobraking to orbit is also an option. Other simulation options include line-of-sight communication guidelines, a choice of propulsion systems, a soft landing on the Earth or Mars, and rendezvous with a target vehicle. The input/output is in metric units, with the exception of thrust and weight which are in English units. Input is read from the user's input file to minimize real-time keyboard input. Output includes vehicle state, orbital and guide parameters, event and total velocity changes, and propellant usage. The main output is to the user defined print file, but during execution, part of the input/output is also displayed on the screen. An included FORTRAN program, TEKPLOT, will display plots on the VDT as well as generating a graphic file suitable for output on most laser printers. The code is double precision. IMP is written in

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)

Stochastic programming problems with generalized integrated chance constraints

Czech Academy of Sciences Publication Activity Database

Branda, Martin

2012-01-01

Roč. 61, č. 8 (2012), s. 949-968 ISSN 0233-1934 R&D Projects: GA ČR GAP402/10/1610 Grant - others:SVV(CZ) 261315/2010 Institutional support: RVO:67985556 Keywords : chance constraints * integrated chance constraints * penalty functions * sample approximations * blending problem Subject RIV: BB - Applied Statistics, Operational Research Impact factor: 0.707, year: 2012 http://library.utia.cas.cz/separaty/2012/E/branda-stochastic programming problems with generalized integrated.pdf

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)

International Piping Integrity Research Group (IPIRG) Program. Final report

International Nuclear Information System (INIS)

Wilkowski, G.; Schmidt, R.; Scott, P.

1997-06-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Program. The IPIRG Program was an international group program managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United States. The program objective was to develop data needed to verify engineering methods for assessing the integrity of circumferentially-cracked nuclear power plant piping. The primary focus was an experimental task that investigated the behavior of circumferentially flawed piping systems subjected to high-rate loadings typical of seismic events. To accomplish these objectives a pipe system fabricated as an expansion loop with over 30 meters of 16-inch diameter pipe and five long radius elbows was constructed. Five dynamic, cyclic, flawed piping experiments were conducted using this facility. This report: (1) provides background information on leak-before-break and flaw evaluation procedures for piping, (2) summarizes technical results of the program, (3) gives a relatively detailed assessment of the results from the pipe fracture experiments and complementary analyses, and (4) summarizes advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG program

International Piping Integrity Research Group (IPIRG) Program. Final report

Energy Technology Data Exchange (ETDEWEB)

Wilkowski, G.; Schmidt, R.; Scott, P. [and others

1997-06-01

This is the final report of the International Piping Integrity Research Group (IPIRG) Program. The IPIRG Program was an international group program managed by the U.S. Nuclear Regulatory Commission and funded by a consortium of organizations from nine nations: Canada, France, Italy, Japan, Sweden, Switzerland, Taiwan, the United Kingdom, and the United States. The program objective was to develop data needed to verify engineering methods for assessing the integrity of circumferentially-cracked nuclear power plant piping. The primary focus was an experimental task that investigated the behavior of circumferentially flawed piping systems subjected to high-rate loadings typical of seismic events. To accomplish these objectives a pipe system fabricated as an expansion loop with over 30 meters of 16-inch diameter pipe and five long radius elbows was constructed. Five dynamic, cyclic, flawed piping experiments were conducted using this facility. This report: (1) provides background information on leak-before-break and flaw evaluation procedures for piping, (2) summarizes technical results of the program, (3) gives a relatively detailed assessment of the results from the pipe fracture experiments and complementary analyses, and (4) summarizes advances in the state-of-the-art of pipe fracture technology resulting from the IPIRG program.

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

Automated Determination of Oxygen-Dependent Enzyme Kinetics in a Tube-in-Tube Flow Reactor.

Science.gov (United States)

Ringborg, Rolf H; Toftgaard Pedersen, Asbjørn; Woodley, John M

2017-09-08

Enzyme-mediated oxidation is of particular interest to synthetic organic chemists. However, the implementation of such systems demands knowledge of enzyme kinetics. Conventionally collecting kinetic data for biocatalytic oxidations is fraught with difficulties such as low oxygen solubility in water and limited oxygen supply. Here, we present a novel method for the collection of such kinetic data using a pressurized tube-in-tube reactor, operated in the low-dispersed flow regime to generate time-series data, with minimal material consumption. Experimental development and validation of the instrument revealed not only the high degree of accuracy of the kinetic data obtained, but also the necessity of making measurements in this way to enable the accurate evaluation of high K MO enzyme systems. For the first time, this paves the way to integrate kinetic data into the protein engineering cycle.

Using Research to Design Integrated Education and Training Programs

Science.gov (United States)

Pappalardo, Michele; Schaffer, William R.

2016-01-01

With the passage of the Workforce Innovation and Opportunity Act (WIOA) of 2014, Northampton Community College began the creation of Integrated Education and Training (IE&T) programs in October 2015. After a needs assessment was conducted with the partners, programs were created to address the needs in the hospitality and healthcare sectors.…

Modeling, Prediction, and Control of Heating Temperature for Tube Billet

Directory of Open Access Journals (Sweden)

Yachun Mao

2015-01-01

Full Text Available Annular furnaces have multivariate, nonlinear, large time lag, and cross coupling characteristics. The prediction and control of the exit temperature of a tube billet are important but difficult. We establish a prediction model for the final temperature of a tube billet through OS-ELM-DRPLS method. We address the complex production characteristics, integrate the advantages of PLS and ELM algorithms in establishing linear and nonlinear models, and consider model update and data lag. Based on the proposed model, we design a prediction control algorithm for tube billet temperature. The algorithm is validated using the practical production data of Baosteel Co., Ltd. Results show that the model achieves the precision required in industrial applications. The temperature of the tube billet can be controlled within the required temperature range through compensation control method.

Spanish approach to research and development applied to steam generator tubes structural integrity and life management

International Nuclear Information System (INIS)

Lozano, J.; Bollini, G.J.

1997-01-01

The operating experience acquired from certain Spanish Nuclear Power Plant steam generators shows that the tubes, which constitute the second barrier to release of fission products, are susceptible to mechanical damage and corrosion as a result of a variety of mechanisms, among them wastage, pitting, intergranular attack (IGA), stress-corrosion cracking (SCC), fatigue-induced cracking, fretting, erosion/corrosion, support plate denting, etc. These problems, which are common in many plants throughout the world, have required numerous investments by the plants (water treatment plants, replacement of secondary side materials such as condensers and heaters, etc.), have meant costs (operation, inspection and maintenance) and have led to the unavailability of the affected units. In identifying and implementing all these preventive and corrective measures, the Spanish utilities have moved through three successive stages: in the initial stage, the main source of information and of proposals for solutions was the Plant Vendor, whose participation in this respect was based on his own Research and Development programs; subsequently, the Spanish utilities participated jointly in the EPRI Steam Generator Owners Group, collaborating in financing; finally, the Spanish utilities set up their own Steam Generator Research and Development program, while maintaining relations with EPRI programs and those of other countries through information interchange

Flux tubes in U(1) - Do they attract or repel each other?

International Nuclear Information System (INIS)

Zach, M.; Faber, M.; Skala, P.

1998-01-01

The dually transformed path integral of four-dimensional U(1) lattice gauge theory is used for a numerical investigation of multiply charged systems and the interaction between flux tubes. For this aim, it is convenient to implement periodically closed flux tubes (torelons) in the dual formulation. We calculate the free energy as well as the total electro-magnetic energy of doubly charged flux tubes as a function of the coupling β. The main results are that the string tension scales proportionally to the charge (contrary to the Coulomb potential) and in the range 0.9<β<1.0 we find a clear signal for attraction between flux tubes. (orig.)

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

Tube holding system

International Nuclear Information System (INIS)

Cunningham, R.C.

1978-01-01

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

Layout of PWR in-core instrumentation system tubing and support structure with Bechtel 3D-CADD

International Nuclear Information System (INIS)

Ichikawa, T.; Pfeifer, B.W.; Mulay, J.N.

1987-01-01

The optimization study of the PWR In-Core Instrumentation System (ICIS) tubing layout and support structure presented an opportunity to utilize the Bechtel 3D-CADD program to perform this task. This paper provides a brief summary of the Bechtel 3D-CADD program development and capabilities and outlines the process of developing and optimizing the ICIS tube layout. Specific aspects relating to the ICIS tube layout criteria, support, alignment, electronic interference check and erection sequence are provided. (orig.)

Foreign energy conservation integrated programs

International Nuclear Information System (INIS)

Lisboa, Maria Luiza Viana; Bajay, Sergio Valdir

1999-01-01

The promotion of energy economy and efficiency is recognized as the single most cost-effective and least controversial component of any strategy of matching energy demand and supply with resource and environmental constraints. Historically such efficiency gains are not out of reach for the industrialized market economy countries, but are unlikely to be reached under present conditions by developing countries and economics in transition. The aim of the work was to analyze the main characteristics of United Kingdom, France, Japan, Canada, Australia and Denmark energy conservation integrated programs

Integrating declarative knowledge programming styles and tools for building expert systems

Energy Technology Data Exchange (ETDEWEB)

Barbuceanu, M; Trausan-Matu, S; Molnar, B

1987-01-01

The XRL system reported in this paper is an integrated knowledge programming environment whose major research theme is the investigation of declarative knowledge programming styles and features and of the way they can be effectively integrated and used to support AI programming. This investigation is carried out in the context of the structured-object representation paradigm which provides the glue keeping XRL components together. The paper describes several declarative programming styles and associated support tools available in XRL. These include an instantiation system supporting a generalized view of the ubiquous frame installation process, a description based programming system providing a novel declarative programming style which embeds a mathematical oriented description language in the structured object environment and a transformational interpreter for using it, a semantics oriented programming framework which offers a specific semantic construct based approach supporting maintenance and evolution and a self description and self generation tool which applies the latter approach to XRL itself. 29 refs., 16 figs.

IAEA integrated safeguards instrumentation program (I2SIP)

International Nuclear Information System (INIS)

Arlt, R.; Fortakov, V.; Gaertner, K.J.

1995-01-01

This article is a review of the IAEA integrated safeguards instrumentation program. The historical development of the program is outlined, and current activities are also noted. Brief technical descriptions of certain features are given. It is concluded that the results of this year's efforts in this area will provide significant input and be used to assess the viability of the proposed concepts and to decide on the directions to pursue in the future

US Department of Energy Integrated Resource Planning Program: Accomplishments and opportunities

Energy Technology Data Exchange (ETDEWEB)

White, D.L. [Oak Ridge National Lab., TN (United States); Mihlmester, P.E. [Aspen Systems Corp., Oak Ridge, TN (United States)

1993-12-17

The US Department of Energy Integrated Resource Planning Program supports many activities and projects that enhance the process by which utilities assess demand and supply options and, subsequently, evaluate and select resources. The US Department of Energy program coordinates integrated resource planning in risk and regulatory analysis; utility and regional planning; evaluation and verification; information transfer/technological assistance; and demand-side management. Professional staff from the National Renewable Energy Laboratory, Oak Ridge National Laboratory, Lawrence Berkeley Laboratory, and Pacific Northwest Laboratories collaborate with peers and stakeholders, in particular, the National Association of Regulatory Utility Commissioners, and conduct research and activities for the US Department of Energy. Twelve integrated resource planning activities and projects are summarized in this report. The summaries reflect the diversity of planning and research activities supported by the Department. The summaries also reflect the high levels of collaboration and teaming that are required by the Program and practiced by the researchers. It is concluded that the Program is achieving its objectives by encouraging innovation and improving planning and decision making. Furthermore, as the Department continues to implement planned improvements in the Program, the Department is effectively positioned to attain its ambitious goals.

HF electronic tubes. Technologies, grid tubes and klystrons

International Nuclear Information System (INIS)

Lemoine, Th.

2009-01-01

This article gives an overview of the basic technologies of electronic tubes: cathodes, electronic optics, vacuum and high voltage. Then the grid tubes, klystrons and inductive output tubes (IOT) are introduced. Content: 1 - context and classification; 2 - electronic tube technologies: cathodes, electronic optics, magnetic confinement (linear tubes), periodic permanent magnet (PPM) focussing, collectors, depressed collectors; 3 - vacuum technologies: vacuum quality, surface effects and interaction with electrostatic and RF fields, secondary emission, multipactor effect, thermo-electronic emission; 4 - grid tubes: operation of a triode, tetrodes, dynamic operation and classes of use, 'common grid' and 'common cathode' operation, ranges of utilisation and limitations, operation of a tetrode on unadjusted load, lifetime of a tetrode, uses of grid tubes; 5 - klystrons: operation, impact of space charge, multi-cavity klystrons, interaction efficiency, extended interaction klystrons, relation between interaction efficiency, perveance and efficiency, ranges of utilization and power limitations, multi-beam klystrons and sheet beam klystrons, operation on unadjusted load, klystron band pass and lifetime, uses; 6 - IOT: principle of operation, ranges of utilisation and limitations, interaction efficiency and depressed collector IOT, IOT lifetime and uses. (J.S.)

What is Program Collaboration and Service Integration (PCSI)?

Centers for Disease Control (CDC) Podcasts

2009-12-07

This podcast provides a description of Program Collaboration and Service Integration (PCSI).  Created: 12/7/2009 by National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention (NCHHSTP).   Date Released: 12/7/2009.

A program for performing angular integrations for transition operators

International Nuclear Information System (INIS)

Froese Fischer, C.; Godefroid, M.R.; Hibbert, A.

1991-01-01

The MCHF-MLTPOL program performs the angular integrations necessary for expressing the matrix elements of transition operators, E1, E2, ..., or M1, M2, ..., as linear combinations of radial integrals. All matrix elements for transitions between two lists of configuration states will be evaluated. A limited amount of non-orthogonality is allowed between orbitals of the initial and final state. (orig.)

Enhancement of leak rate estimation model for corroded cracked thin tubes

International Nuclear Information System (INIS)

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

2010-01-01

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

Plastic collapse of tubes submitted to a ring load by optimization

International Nuclear Information System (INIS)

Zouain, N.

1982-05-01

The limit analysis of a tube with finite lenght, made of a rigid - plastic material, is considered for the case of an internal load uniformely distributed in a cross section of the tube. The exact creep law is calculated for several qualitatively differents cases, namely different tube lenghts. The corresponding stress and collapse mechanisms are given so that they can be compared to the approximations developed here. The static and kinematic theorems on plastic collapse are used for establishing two numerical methods of resolution, specifically mathematical programming and finite element method. These mathematical methods are applied to collapse load for the considered tube. (E.G.) [pt

Treatment of hazardous and toxic liquids using Rochem Disc Tube technology

International Nuclear Information System (INIS)

LaMonica, D.

1992-01-01

Rochem Separation Systems, established in 1990 as a subsidiary of the international Rochem Group, has advanced the treatment of hazardous and toxic liquids with its unique, patented Disc Tube technology. Developed in 1987 at Rochem's design and production facilities in Hamburg, Germany, the Disc Tube technology is a series of membrane modules that greatly reduce the problems that hamper the effectiveness of other treatment technologies (i.e. fouling, scaling, cost, etc.). Applications of the Disc Tube technology include reverse osmosis and ultrafiltration. Rochem was recently accepted into the EPA Superfund Site program as a result of its Disc Tube technology. 1 fig., 1 tab

Human Research Program Integrated Research Plan. Revision A January 2009

Science.gov (United States)

2009-01-01

The Integrated Research Plan (IRP) describes the portfolio of Human Research Program (HRP) research and technology tasks. The IRP is the HRP strategic and tactical plan for research necessary to meet HRP requirements. The need to produce an IRP is established in HRP-47052, Human Research Program - Program Plan, and is under configuration management control of the Human Research Program Control Board (HRPCB). Crew health and performance is critical to successful human exploration beyond low Earth orbit. The Human Research Program (HRP) is essential to enabling extended periods of space exploration because it provides knowledge and tools to mitigate risks to human health and performance. Risks include physiological and behavioral effects from radiation and hypogravity environments, as well as unique challenges in medical support, human factors, and behavioral or psychological factors. The Human Research Program (HRP) delivers human health and performance countermeasures, knowledge, technologies and tools to enable safe, reliable, and productive human space exploration. Without HRP results, NASA will face unknown and unacceptable risks for mission success and post-mission crew health. This Integrated Research Plan (IRP) describes HRP s approach and research activities that are intended to address the needs of human space exploration and serve HRP customers and how they are integrated to provide a risk mitigation tool. The scope of the IRP is limited to the activities that can be conducted with the resources available to the HRP; it does not contain activities that would be performed if additional resources were available. The timescale of human space exploration is envisioned to take many decades. The IRP illustrates the program s research plan through the timescale of early lunar missions of extended duration.

Integration Of Innovative Technologies And Affective Teaching amp Learning In Programming Courses

Directory of Open Access Journals (Sweden)

Alvin Prasad

2015-08-01

Full Text Available Abstract Technology has been integral component in the teaching and learning process in this millennium. In this review paper we evaluate the different technologies which are used to currently facilitate the teaching and learning of computer programming courses. The aim is to identify problems or gaps in technology usage in the learning environment and suggest affective solutions for technology integration into programming courses at the University levels in the future. We believe that with the inclusion of suggested innovative technologies and affective solutions in programming courses teaching and learning will be attractive and best for the programming industry.

A superconducting supercollider calorimeter photomultiplier tube preamplifier circuit

Energy Technology Data Exchange (ETDEWEB)

Panescu, D; Lackey, J; Robl, P; Smith, W H [Wisconsin Univ., Madison, WI (United States). Physics Dept.

1992-07-15

This study presents the design of the front end amplifier for a scintillator calorimeter with photomultiplier tube (PMT) readout. The design is based on analytical computations and SPICE simulations, and is checked against tests performed on a prototyped circuit. We were looking to achieve (1) a very low droop within the 4 ns after the integration of the photomultiplier tube (PMT) signal was completed, (2) a very low noise figure for the whole amplifier in a 100 MHz bandwidth, (3) an input impedance optimized for the PMT which is actually used, (4) baseline restoration as quick as possible at the output of the clip amps, (5) no loss of information due to the saturation at intermediary stages (e.g. integrator), and (6) an output driving 100 {Omega} twisted pair cables, or 50 {Omega} coaxial cables, in order to transmit the signal to switched capacitor arrays for analog storage. (orig.).

47 CFR 76.504 - Limits on carriage of vertically integrated programming.

Science.gov (United States)

2010-10-01

... programming. 76.504 Section 76.504 Telecommunication FEDERAL COMMUNICATIONS COMMISSION (CONTINUED) BROADCAST... Limits on carriage of vertically integrated programming. (a) Except as otherwise provided in this section... national video programming services owned by the cable operator or in which the cable operator has an...

Tube to tube excursive instability - sensitivities and transients

International Nuclear Information System (INIS)

Brown, M.; Layland, M.W.

1980-01-01

A simple basic analysis of excursive instability in a boiler tube shows how it depends upon operating conditions and physical properties. A detailed mathematical model of an AGR boiler is used to conduct a steady state parameter sensitivity survey. It is possible from this basis to anticipate the effects of changes in operating conditions and changes in design parameters upon tube to tube stability. Dynamic responses of tubes operating near the stability threshold are examined using a mathematical model. Simulated excursions are triggered by imparting small abrupt pressure changes on the boiler inlet pressure. The influences of the magnitude of the pressure change, waterside friction factor and gas side coupling between tubes are examined. (author)

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

Condenser tube buckling within tube-tubesheet joints

International Nuclear Information System (INIS)

Willertz, L.E.; Kalnins, A.; Updike, D.P.

1991-01-01

The problem of the appearance of protrusions, or bumps, in the interior of roller-expanded tubes within a tubesheet is addressed. Such bumps have been observed in condensers of power plants. A brief history of the reported occurrences of the bumps is given. The hypothesis is advanced that the mechanics of the formation of the bumps is similar to a buckling problem that has 'bifurcation at infinity'. Following this hypothesis, a two-dimensional physical model is developed, and the application of this model to study a three-dimensional bump is proposed. It is proposed in this paper that an initial deviation from the circular shape of the tube required to produce a bump. It is shown that without such a deviation the tubes cannot buckle. An experiment with short tube segments has been performed that verifies some of the features of the observed condenser tube bumps. Exactly what force produced the initial deviation for the observed bumps is still unknown. Available evidence implicates the hydro-laser jet that is used in the cleaning of tubes and tubesheets. A scenario of how a bump could have been produced by the hydro-laser jet is proposed. (author)

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

Energy transformation and flow topology in an elbow draft tube

Directory of Open Access Journals (Sweden)

Štefan D.

2012-06-01

Full Text Available Paper presents a computational study of energy transformation in two geometrical configurations of Kaplan turbine elbow draft tube. Pressure recovery, hydraulic efficiency and loss coefficient are evaluated for a series of flow rates and swirl numbers corresponding to operating regimes of the turbine. These integral characteristics are then correlated with local flow field properties identified by extraction of topological features. Main focus is to find the reasons for hydraulic efficiency drop of the elbow draft tube.

Diagnosis of Heat Exchanger Tube Failure in Fossil Fuel Boilers Through Estimation of Steady State Operating Conditions

International Nuclear Information System (INIS)

Herszage, A.; Toren, M.

1998-01-01

Estimation of operating conditions for fossil fuel boiler heat exchangers is often required due to changes in working conditions, design modifications and especially for monitoring performance and failure diagnosis. Regular heat exchangers in fossil fuel boilers are composed of tube banks through which water or steam flow, while hot combustion (flue) gases flow outside the tubes. This work presents a top-down approach to operating conditions estimation based on field measurements. An example for a 350 MW unit superheater is thoroughly discussed. Integral calculations based on measurements for all unit heat exchangers (reheaters, superheaters) were performed first. Based on these calculations a scheme of integral conservation equations (lumped parameter) was then formulated at the single tube level. Steady state temperatures of superheater tube walls were obtained as a main output, and were compared to the maximum allowable operating temperatures of the tubes material. A combined lumped parameter - CFD (Computational Fluid Dynamics, FLUENT code) approach constitutes an efficient tool in certain cases. A brief report of such a case is given for another unit superheater. We conclude that steady state evaluations based on both integral and detailed simulations are a valuable monitoring and diagnosis tool for the power generation industry

Integrated Financial Management Program

Science.gov (United States)

Pho, Susan

2004-01-01

Having worked in the Employees and Commercial Payments Branch of the Financial Management Division for the past 3 summers, I have seen the many changes that have occurred within the NASA organization. As I return each summer, I find that new programs and systems have been adapted to better serve the needs of the Center and of the Agency. The NASA Agency has transformed itself the past couple years with the implementation of the Integrated Financial Management Program (IFMP). IFMP is designed to allow the Agency to improve its management of its Financial, Physical, and Human Resources through the use of multiple enterprise module applications. With my mentor, Joseph Kan, being the branch chief of the Employees and Commercial Payments Branch, I have been exposed to several modules, such as Travel Manager, WebTads, and Core Financial/SAP, which were implemented in the last couple of years under the IFMP. The implementation of these agency-wide systems has sometimes proven to be troublesome. Prior to IFMP, each NASA Center utilizes their own systems for Payroll, Travel, Accounts Payable, etc. But with the implementation of the Integrated Financial Management Program, all the "legacy" systems had to be eliminated. As a result, a great deal of enhancement and preparation work is necessary to ease the transformation from the old systems to the new. All this work occurs simultaneously; for example, e-Payroll will "go live" in several months, but a system like Travel Manager will need to have information upgraded within the system to meet the requirements set by Headquarters. My assignments this summer have given me the opportunity to become involved with such work. So far, I have been given the opportunity to participate in projects resulting from a congressional request, several bankcard reconciliations, updating routing lists for Travel Manager, updating the majordomo list for Travel Manager approvers and point of contacts, and a NASA Headquarters project involving

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

Nondestructive inspection of the tubes of TRIGA IPR-R1 reactor heat exchanger by eddy current testing

International Nuclear Information System (INIS)

Silva Junior, Silverio F.; Silva, Roger F.; Oliveira, Paulo F.; Barreto, Erika S.; Ribeiro, Isabela G.; Fraiz, Felipe C.

2013-01-01

The IPR-R1 TRIGA MARK 1 reactor is an open pool type reactor, cooled light water. It is used for research activities, personnel training and radioisotopes production, in operation since 1960 at the Nuclear Technology Development Center - CDTN/CNEN. It operates at a maximum thermal power of 100 kW and usually, the fuel cooling is done by natural circulation. If necessary, an external auxiliary cooling system, with a shell-and-tube type heat exchanger, can be used to improve the water heat removal. As part of the ageing management program of the reactor, a nondestructive evaluation of their heat exchanger stainless steel tubes will be performed, in order to verify its integrity. The examinations will be performed using the eddy current test method, which allows the detection and characterization of structural discontinuities in the wall of the tubes, if existing. For this purpose, probes and reference standards were designed and manufactured at CDTN facilities and test procedures were established and validated. In this paper, a description of the proposed infrastructure as well as the test methodology to be used in the examinations are presented and discussed. (author)

MINIMIZING THE PREPARATION TIME OF A TUBES MACHINE: EXACT SOLUTION AND HEURISTICS

Directory of Open Access Journals (Sweden)

Robinson S.V. Hoto

Full Text Available ABSTRACT In this paper we optimize the preparation time of a tubes machine. Tubes are hard tubes made by gluing strips of paper that are packed in paper reels, and some of them may be reused between the production of one and another tube. We present a mathematical model for the minimization of changing reels and movements and also implementations for the heuristics Nearest Neighbor, an improvement of a nearest neighbor (Best Nearest Neighbor, refinements of the Best Nearest Neighbor heuristic and a heuristic of permutation called Best Configuration using the IDE (integrated development environment WxDev C++. The results obtained by simulations improve the one used by the company.

Design of experiments and equipment to test the ballooning characteristics of CANDU pressure tubes

International Nuclear Information System (INIS)

Forrest, C.F.; Stern, F.; Hart, R.G.

1992-01-01

Experiments have been planned and an apparatus has been designed to enable creep testing of end-of-life pressure tube specimens in a LOCA environment. Effects that could be studied include: annealing of irradiation damage during transient heating; effects of hydride blisters on pressure tube ballooning strains; and, effects of uniformly-distributed hydrogen content on pressure tube ballooning strains. The proposed experimental program will consist of separate effects creep tests on pressure tube sections under transient heating conditions

The PISC programme on defective steam generator tubes inspection. A status report

International Nuclear Information System (INIS)

Birac, C.; Comby, R.; Maciga, G.; Von Estorff, U.; Zanella, G.L.

1994-06-01

The general objective of the PISC Program (Programme for the Inspection of Steel Components) is to assess experimentally procedures and techniques in use for the in-service inspection of pressure components. The program is mainly a round robin test, the results of which are compared with real characteristics of the flaws obtained by destructive analysis. Materials tested are INCONEL 600 tubes, diameter 22.22 mm, wall thickness 1.27 mm. The technique applied is eddy current testing. The program of capability tests on loose tubes was started in 1990, the round robin tests ended in 1993. The preliminary results are presented. (R.P.). 8 refs., 9 figs., 4 tabs

Development and quality assessments of commercial heat production of ATF FeCrAl tubes

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-09-01

Development and quality assessment of the 2^nd generation ATF FeCrAl tube production with commercial manufacturers were conducted. The manufacturing partners include Sophisticated Alloys, Inc. (SAI), Butler, PA for FeCrAl alloy casting via vacuum induction melting, Oak Ridge National Laboratory (ORNL) for extrusion process to prepare the master bars/tubes to be tube-drawn, and Rhenium Alloys, Inc. (RAI), North Ridgeville, OH, for tube-drawing process. The masters bars have also been provided to Los Alamos National Laboratory (LANL) who works with Century Tubes, Inc., (CTI), San Diego, CA, as parallel tube production effort under the current program.

Integration of safety engineering into a cost optimized development program.

Science.gov (United States)

Ball, L. W.

1972-01-01

A six-segment management model is presented, each segment of which represents a major area in a new product development program. The first segment of the model covers integration of specialist engineers into 'systems requirement definition' or the system engineering documentation process. The second covers preparation of five basic types of 'development program plans.' The third segment covers integration of system requirements, scheduling, and funding of specialist engineering activities into 'work breakdown structures,' 'cost accounts,' and 'work packages.' The fourth covers 'requirement communication' by line organizations. The fifth covers 'performance measurement' based on work package data. The sixth covers 'baseline requirements achievement tracking.'

Integrated inspection programs at Bruce Heavy Water Plant

Energy Technology Data Exchange (ETDEWEB)

Brown, K C [Ontario Hydro, Tiverton, ON (Canada)

1993-12-31

Quality pressure boundary maintenance and an excellent loss prevention record at Bruce Heavy Water Plant are the results of the Material and Inspection Unit`s five inspection programs. Experienced inspectors are responsible for the integrity of the pressure boundary in their own operating area. Inspectors are part of the Technical Section, and along with unit engineering staff, they provide technical input before, during, and after the job. How these programs are completed, and the results achieved, are discussed. 5 figs., 1 appendix.

Integrated inspection programs at Bruce Heavy Water Plant

International Nuclear Information System (INIS)

Brown, K.C.

1992-01-01

Quality pressure boundary maintenance and an excellent loss prevention record at Bruce Heavy Water Plant are the results of the Material and Inspection Unit's five inspection programs. Experienced inspectors are responsible for the integrity of the pressure boundary in their own operating area. Inspectors are part of the Technical Section, and along with unit engineering staff, they provide technical input before, during, and after the job. How these programs are completed, and the results achieved, are discussed. 5 figs., 1 appendix

A Study of Nonlinear Variable Viscosity in Finite-Length Tube with Peristalsis

Directory of Open Access Journals (Sweden)

Y. Abd Elmaboud

2014-01-01

Full Text Available Peristaltic motion of an incompressible Newtonian fluid with variable viscosity induced by periodic sinusoidal traveling wave propagating along the walls of a finite-length tube has been investigated. A perturbation method of solution is sought. The viscosity parameter α (α << 1 is chosen as a perturbation parameter and the governing equations are developed up to the first-order in the viscosity parameter (α. The analytical solution has been derived for the radial velocity at the tube wall, the axial pressure gradient across the length of the tube, and the wall shear stress under the assumption of low Reynolds number and long wavelength approximation. The impacts of physical parameters such as the viscosity and the parameter determining the shape of the constriction on the pressure distribution and on the wall shear stress for integral and non-integral number of waves are illustrated. The main conclusion that can be drawn out of this study is that the peaks of pressure fluctuate with time and attain different values with non-integral numbers of peristaltic waves. The considered problem is very applicable in study of biological flow and industrial flow.

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)

Development of engineering program for integrity evaluation of pipes with local wall thinned defects

International Nuclear Information System (INIS)

Park, Chi Yong; Lee, Sung Ho; Kim, Tae Ryong; Park, Sang Kyu

2008-01-01

Integrity evaluation of pipes with local wall thinning by erosion and corrosion is increasingly important in maintenance of wall thinned carbon steel pipes in nuclear power plants. Though a few program for integrity assessment of wall thinned pipes have been developed in domestic nuclear field, however those are limited to straight pipes and methodology proposed in ASME Sec.XI Code Case N-597. Recently, the engineering program for integrity evaluation of pipes with all kinds of local wall defects such as straight, elbow, reducer and branch pipes was developed successfully. The program was designated as PiTEP (Pipe Thinning Evaluation Program), which name was registered as a trademark in the Korea Intellectual Property Office. A developed program is carried out by sequential step of four integrity evaluation methodologies, which are composed of construction code, code case N-597, its engineering method and two developed owner evaluation method. As PiTEP program will be performed through GUI (Graphic User Interface) with user's familiarity, it would be conveniently used by plant engineers with only measured thickness data, basic operation conditions and pipe data

Process improvement program evolves into compliance program at an integrated delivery system.

Science.gov (United States)

Tyk, R C; Hylton, P G

1998-09-01

An integrated delivery system discovered questionable practices when it undertook a process-improvement initiative for its revenue-to-cash cycle. These discoveries served as a wake-up call to the organization that it needed to develop a comprehensive corporate compliance program. The organization engaged legal counsel to help it establish such a program. A corporate compliance officer was hired, and a compliance committee was set up. They worked with counsel to develop the structure and substance of the program and establish a corporate code of conduct that became a part of the organization's policies and procedures. Teams were formed in various areas of the organization to review compliance-related activities and suggest improvements. Clinical and nonclinical staff attended mandatory educational sessions about the program. By approaching compliance systematically, the organization has put itself in an excellent position to avoid fraudulent and abusive activities- and the government scrutiny they invite.

Concrete containment integrity program at EPRI

International Nuclear Information System (INIS)

Winkleblack, R.K.; Tang, Y.K.

1984-01-01

Many in the nuclear power plant business believe that the catastrophic failure mode for reactor containment structures is unrealistic. One of the goals of the EPRI containment integrity program is to demonstrate that this is true. The objective of the program is to provide the utility industry with an experimental data base and a test-validated analytical method for realistically evaluating the actual over-pressure capability of concrete containment buildings and to predict leakage behavior if higher pressures were to occur. The ultimate goal of this research effort is to characterize the containment leakage mode and rate as a function of internal pressure and time so that the risk can be realistically assessed for hypothetical degraded core accidents. Progress in the first and second phases of the three-phase analytical and testing efforts is discussed

Chest tube insertion

Science.gov (United States)

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

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.

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

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)

Quality assurance and quality control in fabrication of heat exchanger tubes

International Nuclear Information System (INIS)

Duennewald, A.

1980-01-01

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

Control of Cell Wall Extensibility during Pollen Tube Growth

OpenAIRE

Hepler, Peter K.; Rounds, Caleb M.; Winship, Lawrence J.

2013-01-01

Tip-growing pollen tubes achieve rapid elongation while maintaining cell wall integrity by balancing local expansion, controlled by local changes in wall viscosity, against exocytosis, influenced by the activity of the actin cytoskeleton, cellular energetics, and calcium and proton physiology.

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)

DOE`s integrated low-level waste management program and strategic planning

Energy Technology Data Exchange (ETDEWEB)

Duggan, G. [Dept. of Energy, Washington, DC (United States). Office of Environmental Restoration and Waste Management; Hwang, J. [Science Applications International Corp., Germantown, MD (United States)

1993-03-01

To meet the DOE`s commitment to operate its facilities in a safe, economic, and environmentally sound manner, and to comply with all applicable federal, state, and local rules, regulations, and agreements, DOE created the Office of Environmental Restoration and Waste Management (EM) in 1989 to focus efforts on controlling waste management and cleaning up contaminated sites. In the first few years of its existence, the Office of Waste Management (EM-30) has concentrated on operational and corrective activities at the sites. In 1992, the Office of Waste Management began to apply an integrated approach to managing its various waste types. Consequently, DOE established the Low-Level Waste Management Program (LLWMP) to properly manage its complex-wide LLW in a consistent manner. The objective of the LLWMP is to build and operate an integrated, safe, and cost-effective program to meet the needs of waste generators. The program will be based on acceptable risk and sound planning, resulting in public confidence and support. Strategic planning of the program is under way and is expected to take two to three years before implementation of the integrated waste management approach.

Study of creep collapse of tubes subject to external pressure at elevated temperature

International Nuclear Information System (INIS)

Takikawa, N.

1982-01-01

Intermediate heat exchanger (IHX) tubes of VHTR form the boundary between the primary and secondary coolants of the reactor. The tubes are subject to external pressures at a postulated secondary coolant depressurization accident, which might lead to creep collapse. Therefore, it is necessary to ensure the integrity against creep collapse by analysis. The objective of this work is to study a simplified analytical method for predicting collapse time of a curved tube subjected to an external pressure. The study is made based on the comparison of experimental collapse time of curved and straight tubes. Creep collapse tests were conducted under an elevated temperature and an external pressure. Test results showed that curved tubes had longer collapse time than straight tubes with the same cross sectional ovality. The simplified analytical method for a curved tube is proposed in this report, which is to compute collapse time of a straight tube with the same ovality. And in this method the computed time is considered as collapse time of the curved tube. The above test results show that this simplified method gives the conservative collapse time. And it is confirmed by additional IHX tube tests that the method is applicable to creep collapse analysis of IHX tubes

Unpacking vertical and horizontal integration: childhood overweight/obesity programs and planning, a Canadian perspective

Directory of Open Access Journals (Sweden)

Ashley Lisa

2010-05-01

Full Text Available Abstract Background Increasingly, multiple intervention programming is being understood and implemented as a key approach to developing public health initiatives and strategies. Using socio-ecological and population health perspectives, multiple intervention programming approaches are aimed at providing coordinated and strategic comprehensive programs operating over system levels and across sectors, allowing practitioners and decision makers to take advantage of synergistic effects. These approaches also require vertical and horizontal (v/h integration of policy and practice in order to be maximally effective. Discussion This paper examines v/h integration of interventions for childhood overweight/obesity prevention and reduction from a Canadian perspective. It describes the implications of v/h integration for childhood overweight and obesity prevention, with examples of interventions where v/h integration has been implemented. An application of a conceptual framework for structuring v/h integration of an overweight/obesity prevention initiative is presented. The paper concludes with a discussion of the implications of vertical/horizontal integration for policy, research, and practice related to childhood overweight and obesity prevention multiple intervention programs. Summary Both v/h integration across sectors and over system levels are needed to fully support multiple intervention programs of the complexity and scope required by obesity issues. V/h integration requires attention to system structures and processes. A conceptual framework is needed to support policy alignment, multi-level evaluation, and ongoing coordination of people at the front lines of practice. Using such tools to achieve integration may enhance sustainability, increase effectiveness of prevention and reduction efforts, decrease stigmatization, and lead to new ways to relate the environment to people and people to the environment for better health for children.

Unpacking vertical and horizontal integration: childhood overweight/obesity programs and planning, a Canadian perspective.

Science.gov (United States)

Maclean, Lynne M; Clinton, Kathryn; Edwards, Nancy; Garrard, Michael; Ashley, Lisa; Hansen-Ketchum, Patti; Walsh, Audrey

2010-05-17

Increasingly, multiple intervention programming is being understood and implemented as a key approach to developing public health initiatives and strategies. Using socio-ecological and population health perspectives, multiple intervention programming approaches are aimed at providing coordinated and strategic comprehensive programs operating over system levels and across sectors, allowing practitioners and decision makers to take advantage of synergistic effects. These approaches also require vertical and horizontal (v/h) integration of policy and practice in order to be maximally effective. This paper examines v/h integration of interventions for childhood overweight/obesity prevention and reduction from a Canadian perspective. It describes the implications of v/h integration for childhood overweight and obesity prevention, with examples of interventions where v/h integration has been implemented. An application of a conceptual framework for structuring v/h integration of an overweight/obesity prevention initiative is presented. The paper concludes with a discussion of the implications of vertical/horizontal integration for policy, research, and practice related to childhood overweight and obesity prevention multiple intervention programs. Both v/h integration across sectors and over system levels are needed to fully support multiple intervention programs of the complexity and scope required by obesity issues. V/h integration requires attention to system structures and processes. A conceptual framework is needed to support policy alignment, multi-level evaluation, and ongoing coordination of people at the front lines of practice. Using such tools to achieve integration may enhance sustainability, increase effectiveness of prevention and reduction efforts, decrease stigmatization, and lead to new ways to relate the environment to people and people to the environment for better health for children.

A miniature pulse tube cryocooler used in a superspectral imager

Science.gov (United States)

Jiang, Zhenhua; Wu, Yinong

2017-05-01

In this paper, we describe a hihg0 frequency pulse tube cryocooler used in a superspectral imager to be launched in 2020. The superspectral imager is a field-dividing optical imaging system and uses 14 sets of integrated IR detector cryocooler dewar assembly. For the requirements of less heat loss an smaller size, each set is highly integrated by directly mounting the IR dectector's sapphire substrate on the pulse tube's cold tip, and welding the dewar's housing to the flange of the cold finger. Driven by a pair of moving magnet linear motors, the dual-opposed piston compressor of the croycooler is running at 120Hz. Filled with customized stainless screens in the regenerator, the cryolooler reaches 8.1% carnot efficiency at the cooling power of 1W@80K with 34Wac input power.

Failure maps for internally pressurized Zr-2.5% Nb pressure tubes with circumferential temperature variations

International Nuclear Information System (INIS)

Shewfelt, R.S.W.

1986-01-01

During some postulated loss-of-coolant accidents, the pressure tube temperature may rise before the internal pressure drops, causing the pressure tube to balloon. The temperature around the pressure tube circumference would likely be nonuniform, producing localized deformation that could possibly cause failure. The computer program, GRAD, was used to determine the circumferential temperature distribution required to cause an internally pressurized Zr-2.5% Nb pressure tube to fail before coming into full contact with its calandria tube. These results were used to construct failure maps. 7 refs

Development of a Mathematics, Science, and Technology Education Integrated Program for a Maglev

Science.gov (United States)

Park, Hyoung Seo

2006-01-01

The purpose of the study was to develop an MST Integrated Program for making a Maglev hands-on activity for higher elementary school students in Korea. In this MST Integrated Program, students will apply Mathematics, Science, and Technology principles and concepts to the design, construction, and evaluation of a magnetically levitated vehicle. The…

Determination of dislocation density in Zr-2.5Nb pressure tubes by x-ray

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Isaenkova, Perlovich; Cheong, Y. M.; Kim, S. S.; Yim, K. S.; Kwon, Sang Chul

2000-11-01

For X-ray determination of the dislocation density in CANDU Zr-2.5%Nb pressure tubes, a program was developed, using the Fourier analysis of X-ray line profiles and calculation of dislocation density by values of the coherent block size and the lattice distortion. The coincidence of obtained values of c- and a-dislocations with those, determined by the X-ray method for the same tube in AECL, was assumed to be the main criterion of validity of the developed program. The final variant of the program allowed to attain a rather close coincidence of calculated dislocation densities with results of AECL. The dislocation density was determined in all the zirconium grains with different orientations based on the texture of the stree-relieved CANDU tube. The complete distribution of c-dislocation density in -Zr grains depecding on their crystallographic orientations was constructed. The distribution of a-dislocation density within the texture maximum at L-direction, containing prismatic axes of all grains, was constructed as well. The analysis of obtained distributions testifies that -Zr grains of the stree-relieved CANDU tube significantly differ in their dislocation densities. Plotted diagrams of correlation between the dislocation density and the pole density allow to estimate the actual connection between texture and dislocation distribution in the studied tube. The distributions of volume fractions of all the zirconium grains depending on their dislocation density were calculated both for c- and a-dislocations. The distributions characterizes quantitatively the inhomogeneity of substructure conditions in the stress-relieved CANDU tube. the optimal procedure for determination of Nb content in {beta}-phases of CANDU Zr-2.5%Nb pressure tubes was also established.

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)

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)

Unsteady Model for Transverse Fluid Elastic Instability of Heat Exchange Tube Bundle

Directory of Open Access Journals (Sweden)

Jun Liu

2014-01-01

Full Text Available From the viewpoint of practical application, based on the unsteady analytical model for transverse fluid elastic instability of tube array proposed by Yetisir and the linear attenuation function introduced by Li Ming, a new explicit model based on nonsteady state “streamtube” hypothesis is proposed and solved using complex number method. In the model, numerical integral is avoided and inappropriate aspects in Li Ming model are modified. Using the model, the fluid elastic instability analysis of a single flexible tube is made. The stability graphs for four typical types of tube array are plotted and contrasted with experimental results. It is found that the current explicit model is effective in the analysis of transverse fluid elastic instability of tube bundle.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

In Situ Remediation Integrated Program, Evaluation and assessment of containment technology

International Nuclear Information System (INIS)

Gerber, M.A.; Fayer, M.J.

1994-04-01

The In Situ Remediation Integrated Program (ISRIP) was established by the US Department of Energy (DOE) to advance the state-of-the art of innovative in situ remediation technologies to the point of demonstration and to broaden the applicability of these technologies to the widely varying site remediation requirements throughout the DOE complex. This program complements similar ongoing integrated demonstration programs being conducted at several DOE sites. The ISRIP has been conducting baseline assessments on in situ technologies to support program planning. Pacific Northwest Laboratory conducted an assessment and evaluation of subsurface containment barrier technology in support of ISRIP's Containment Technology Subprogram. This report summarizes the results of that activity and provides a recommendation for priortizing areas in which additional research and development is needed to advance the technology to the point of demonstration in support of DOE's site restoration activities

AN INTEGRATIVE GROUP PSYCHOTHERAPY PROGRAM FOR CHILDREN. THE WIZARDING SCHOOL

Directory of Open Access Journals (Sweden)

Oana Maria Popescu

2012-02-01

Full Text Available One of the most important tendencies in child psychotherapy is the integration of various psychotherapeutic approaches and technical interventions belonging to different orientations. Based on the Harry Potter stories, the „Wizarding School” structured group therapy program is a 12-step integratively oriented program applicable in personal development, individual and group therapy for children aged 6 to 13 (at present being adapted for adult psychotherapy. The program takes place within a fairy tale, being therefore a type of informal hypnotic trance. The interventions are drawn from the lessons described in Harry Potter’s story at Hogwarts, based on the fundamental principles of child psychotherapy and including elements of play therapy, art therapy, hypnotherapy, cognitive- behavioural therapy, transactional analysis, supportive therapy, family therapy and person centred therapy. From a theoretical point of view the program is based on elements from a number of psychotherapeutic approaches, the main concept being that we need to create a therapeutic myth that is acceptable to a child. The program is not suitable for children with structural deficits, who have difficulties in making the difference between fantasy and reality.

Effect of boric acid on intergranular corrosion in tube support plate crevices

International Nuclear Information System (INIS)

Brunet, J.P.; Campan, J.L.

1993-10-01

Intergranular attack on steam generator tubing is one important phenomenon involved in availability of Pressurized Water Reactors. Boric acid appears to be a possible candidate for inhibiting the corrosion process. The program performed in Cadarache was supposed to give statistical informations on the boric acid effect. It was based on a large number of samples initially attacked during a program performed by BABCOCK ampersand WILCOX. These samples were sleeved onto Alloy 690 tubes, in order to prevent premature cracking. Unfortunately it was not possible to find chemical conditions able to produce significant additional corrosion; we postulated mainly due to a drastic reduction of the thermal flux resulting from the increase of the tube wall thickness under the tube support plates (TSP). The tests demonstrate that such sleeve could be a possible remedy of the corrosion when introduced under the TSP. The tests show indications of a possible beneficial effect of the boric acid, a large variability of the heats sensitivity to the IGA and a predominant effect of Na 2 CO 3 on IGA production

Computational fluid dynamic analysis of a guide tube in a PWR

International Nuclear Information System (INIS)

Hofmann, F.; Archambeau, F.; Chaize, C.

2000-01-01

Security in nuclear power plants demands severe limitations of the maximal drop time of rod cluster control assemblies. In February 1995, several assemblies of the Chinese plant in Daya Bay failed to comply with these requirements. Electricite De France undertook a research program to get a better insight of this problem since the plant has been built by French and also because the French new four-loops N4 reactor was equipped with the same guide tubes. This paper is limited to a numerical study of the influence of the pressure forces applied to control rods and due to flow circulation through the guide tubes. After a validation test case, a first calculation has been carried out on a simplified N4 guide tube. The sensitivity of the pressure forces to transverse flow and to modifications of the geometry has been determined. The program has been extended to guide tubes used in 1300-MW reactors and similar computations have been done. To make simulations more representative, a global computation of the flow in the whole upper internals plenum (UIP) will be achieved to provide accurate boundary conditions for local calculations with better resolution.

Computational fluid dynamic analysis of a guide tube in a PWR

International Nuclear Information System (INIS)

Hofmann, F.; Archambeau, F.

1997-01-01

Security in nuclear power plants demands severe limitations of the maximal drop time of rod cluster control assemblies. In February 1995, several assemblies of the Chinese plant in DAYA BAY failed to comply with these requirements. Electricite de France undertook a research program to get a better insight of this problem since the plant has been built by French and also because the French new 4 loops N4 reactor was equipped with the same guide tubes. This paper is limited to a numerical study of the influence of the pressure forces applied to control rods and due to flow circulation through the guide tubes. After a validation test-case, a first calculation has been carried out on a simplified N4 guide tube. The sensitivity of the pressure forces to transverse flow and to modifications of the geometry has been determined. The program has been extended to guide tubes used in 1300 MW reactors and similar computations have been done. To make simulations more representative, a global computation of the flow in the whole upper internals plenum (UIP) will be achieved to provide accurate boundary conditions for local calculations with better resolution. (author)

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

Evaporation of new refrigerants on tubes with improved surfaces; Evaporation de nouveaux refrigerants sur des tubes a surface amelioree

Energy Technology Data Exchange (ETDEWEB)

Kattan, N.; Favrat, D.; Thome, J. R.; Nidegger, E.; Zuercher, O. [Ecole Polytechnique Federale, Lab. d` Energetique Industrielle (LENI), Lausanne (Switzerland)

1995-07-15

The substitution of old refrigerants in refrigeration systems, heat pumps and organic Rankine cycles for heat recovery, request a good knowledge of heat transfer properties of substitute fluids. The test measurements in LENI test facility (concentric tubes with water flowing in a counter-current flow) with new refrigerants like HFC134a, HCFC123, R-404A, R-402A, have established a new data bank with new refrigerants, a comparison with old refrigerants like CFC11, CFC12 CFC/HCFC502 and with existent correlations. Correlations were programmed to calculate and compare heat transfer coefficient during the tests. To develop a new correlation based on flow regimes, a high speed Sony video tape camera is used to observe and identify flow patterns. Important images are captured, digitalized, stored for later analysis and sent to a color plotter. Several flow pattern maps were programmed and compared to flow regimes observed on the test rig. Local flow boiling heat transfer coefficients were measured for HFC134a and HCFC123 evaporating inside a microfin tube. In addition, microfin heat transfer augmentation relative to plain tube test data was investigated. The presence of oil in the evaporator has an effect on heat transfer coefficient. Local flow boiling heat transfer coefficients were measured for refrigerant HFC134a-oil ester (Mobil EAL Arctic 68). A new thermodynamic approach for modeling mixtures of refrigerants and lubricating oils is developed. A very high accuracy, straight vibrating tube type of density flowmeter is used to measure oil concentrations of flowing HFC134a-oil mixtures. (author) 28 figs., 25 refs.

Pressure tube reactor

International Nuclear Information System (INIS)

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

1993-01-01

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

Implementation of an integrity management program in a crude oil pipeline system

Energy Technology Data Exchange (ETDEWEB)

Martinez, Maria; Tomasella, Marcelo [Oleoductos del Valle, General Roca (Argentina); Rossi, Juan; Pellicano, Adolfo [SINTEC S.A. , Mar del Plata, Buenos Aires (Argentina)

2005-07-01

The implementation of an Integrity Management Program (IMP) in a crude oil pipeline system is focused on the accomplishment of two primary corporative objectives: to increase safety operation margins and to optimize available resources. A proactive work philosophy ensures the safe and reliable operation of the pipeline in accordance with current legislation. The Integrity Management Program is accomplished by means of an interdisciplinary team that defines the strategic objectives that complement and are compatible with the corporative strategic business plan. The implementation of the program is based on the analysis of the risks due to external corrosion, third party damage, design and operations, and the definition of appropriate mitigation, inspection and monitoring actions, which will ensure long-term integrity of the assets. By means of a statistical propagation model of the external defects, reported by high-resolution magnetic inspection tool (MFL), together with the information provided by corrosion sensors, field repair interventions, close internal surveys and operation data, projected defect depth; remaining strength and failure probability distributions were obtained. From the analysis, feasible courses of action were established, including the inspection and repair plan, the internal inspection program and both corrosion monitoring and mitigation programs. (author)

Verification tests for GRAD, a computer program to predict nonuniform deformation and failure of Zr-2.5 wt percent Nb pressure tubes during a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Shewfelt, R.S.W.; Godin, D.P.

1985-03-01

During a postulated loss-of-coolant accident in a CANDU reactor, the temperature of the pressure tubes could rise sufficiently so that ballooning could occur. It is also likely that there would be a variation in temperature around the tube circumference, causing the deformation to be nonuniform. Since the deformation of the pressure tube controls how the core heat is transferred to the surrounding moderator, which is a large heat sink, a computer program, GRAD, has been developed to predict this nonuniform deformation. Numerous biaxial creep tests were done, where the temperature of internally pressurized sections of Zr-2.5 wt percent Nb pressure tubes were ramped to check the ability of GRAD to predict the resulting nonuniform deformation and possible tube failure. GRAD was successful in predicting the average transverse creep strain observed during the tests and the local transverse creep strain at the end of the tests. GRAD was also able to predict the failure time and average transverse creep strain at failure for all the specimens that failed

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

A typical pressurized water reactor (PWR) steam generator has approximately 10,000 tubes. These tubes have different geometries, supporting conditions, and different material properties due to the non-uniform temperature distribution throughout the steam generator. Even though some tubes may have the same geometry and boundary conditions, the non-uniform distribution of coolant densities adjacent to the tubes causes them to have different added mass effects and dynamic characteristics. Therefore, for a reliable design of the steam generator, a separate modal analysis for each tube is necessary to perform the FIV (flow-induced vibration) analysis. However, the modal analysis of a tube including the finite element modeling is cumbersome and takes lots of time. And when a commercial finite element code is used, interfacing the modal analysis result, such as natural frequencies and mode shapes, with the FIV analysis procedure requires an additional significant amount of time and can possibly incur inadvertent error due to the complexity of data processing. It is therefore impossible to perform the complete FIV analysis for ten thousands of tubes when designing or maintaining a steam generator although it is necessary. Rather, to verify the safe design against the FIV, only a couple of tubes are chosen based on engineering judgment or past experience. In this paper, a computer program, PIAT-MODE, was developed which is able to perform modal analysis of all tubes of a PWR steam generator in a very efficient way. The geometries and boundary conditions of every tube were incorporated into PIAT-MODE using appropriate mathematical formulae. Material property data including the added mass effect was also included in the program. Once a specific tube is selected, the program automatically constructs the finite element model and generates the modal data very quickly. Therefore, modal analysis can be performed for every single tube in a straight way. When PIAT-MODE is coupled

Analysis to develop a program for energy-integrated farm systems

Energy Technology Data Exchange (ETDEWEB)

Eakin, D.E.; Clark, M.A.; Inaba, L.K.; Johnson, K.I.

1981-09-01

A program to use renewable energy resources and possibly develop decentralization of energy systems for agriculture is discussed. The purpose of the research presented is to establish the objective of the program and identify guidelines for program development. The program's objective is determined by: (1) an analysis of the technologies that could be utilized to transform renewable farm resources to energy by the year 2000, (2) the quantity of renewable farm resources that are available, and (3) current energy-use patterns. Individual research, development, and demonstration projects are fit into a national program of energy-integrated farm systems on the basis of: (1) market need, (2) conversion potential, (3) technological opportunities, and (4) acceptability. Quantification of these factors for the purpose of establishing program guidelines is conducted using the following four precepts: (1) market need is identified by current use of energy for agricultural production; (2) conversion potential is determined by the availability of renewable resources; and (3) technological opportunities are determined by the state-of-the-art methods, techniques, and processes that can convert renewable resources into farm energy. Each of these factors is analyzed in Chapters 2 to 4. Chapter 5 draws on the analysis of these factors to establish the objective of the program and identify guidelines for the distribution of program funds. Chapter 6 then discusses the acceptability of integrated farm systems, which can not be quantified like the other factors.

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.)

Feasibility study on the guided wave technique for condenser tube in NPP

International Nuclear Information System (INIS)

Choi, Sung Nam; Kim, Young Ho; Kim, Hyung Nam; Yoo, Hyun Joo; Hwang, W. G.

2004-01-01

The condenser tube is examined by the eddy current test (ECT) method to identify the integrity of the nuclear power plant. Because ECT probe is moved through the tube inside to identify flaws, the ECT probe should be exchanged periodically due to the wear of probe surface in order to remove the noise form the ECT signal. Moreover, it is impossible to examine the tube by ECT method because the ECT probe can not move through the inside due to the deformation such as dent. Recently, the theory of guided wave was established and the equipment applying the theory has been actively developed so as to overcome the limitation of ECT method for the tube inspection of heater exchanger in nuclear power plant. The object of this study is to know the feasibility of applying the guided wave technique to condenser tube in NPP

Integrating Cybersecurity into the Program Management Organization

Science.gov (United States)

2015-05-13

penalty for failing to comply with a collection of information if it does not display a currently valid OMB control number. 1. REPORT DATE 13 MAY 2015 2...Threat to our National Economy DOD Cybersecurity Gaps Could Be Canary in Federal Acquisition Coal Mine Intangible Assets Create Vulnerabilities...operational approach integrates with current or planned CONOPS, BCP, information architecture, programs or initiatives Development  Approach to

Thermal-hydraulic study of integrated steam generator in PWR

International Nuclear Information System (INIS)

Osakabe, Masahiro

1989-01-01

One of the safety aspects of innovative reactor concepts is the integration of steam generators (SGs) into the reactor vessel in the case of the pressurized water reactor (PWR). All of the reactor system components including the pressurizer are within the reactor vessel in the SG integrated PWR. The simple heat transfer code was developed for the parametric study of the integrated SG. The code was compared to the once-through 19-tube SG experiment and the good agreement between the experimental results and the code predictions was obtained. The assessed code was used for the parametric study of the integrated once-through 16 m-straight-tube SG installed in the annular downcomer. The proposed integrated SG as a first attempt has approximately the same tube size and pitch as the present PWR and the SG primary and secondary sides in the present PWR is inverted in the integrated PWR. Based on the study, the reactor vessel size of the SG integrated PWR was calculated. (author)

Integration of the TNXYZ computer program inside the platform Salome

International Nuclear Information System (INIS)

Chaparro V, F. J.

2014-01-01

The present work shows the procedure carried out to integrate the code TNXYZ as a calculation tool at the graphical simulation platform Salome. The TNXYZ code propose a numerical solution of the neutron transport equation, in several groups of energy, steady-state and three-dimensional geometry. In order to discretized the variables of the transport equation, the code uses the method of discrete ordinates for the angular variable, and a nodal method for the spatial dependence. The Salome platform is a graphical environment designed for building, editing and simulating mechanical models mainly focused on the industry and unlike other software, in order to form a complete scheme of pre and post processing of information, to integrate and control an external source code. Before the integration the in the Salome platform TNXYZ code was upgraded. TNXYZ was programmed in the 90s using Fortran 77 compiler; for this reason the code was adapted to the characteristics of the current Fortran compilers; in addition, with the intention of extracting partial results over the process sequence, the original structure of the program underwent a modularization process, i.e. the main program was divided into sections where the code performs major operations. This procedure is controlled by the information module (YACS) on Salome platform, and it could be useful for a subsequent coupling with thermal-hydraulics codes. Finally, with the help of the Monte Carlo code Serpent several study cases were defined in order to check the process of integration; the verification process consisted in performing a comparison of the results obtained with the code executed as stand-alone and after modernized, integrated and controlled by the Salome platform. (Author)

Use of CATHENA to model calandria-tube/moderator heat transfer after pressure-tube/calandria-tube ballooning contact

International Nuclear Information System (INIS)

Fan, H.Z.; Bilanovic, Z.; Nitheanandan, T.

2004-01-01

A study was performed to assess the effect of the calandria-tube/moderator heat transfer after pressure-tube/calandria tube ballooning contact using CATHENA. Results of this study indicated that the analytical tool, CATHENA, can be applied for pool boiling heat transfer on the external surface of a large diameter tube, such as the calandria tube used in CANDU reactors. The methodology in such CANDU-generic study can be used to simulate the tube surface with multiple boiling regimes and to assess the benefits of closely coupling thermalhydraulics modelling and fuel/fuel channel behaviour modelling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a one-dimensional, two-fluid thermalhydraulic simulation code designed by AECL to analyse two-phase flow and heat transfer in piping networks. The detailed heat transfer package in CATHENA allows a connection to be established from the multiple solid surfaces of tubes to the surrounding large amount of moderator water, which acts as a heat sink during a postulated loss of coolant event. The generalized heat transfer package within CATHENA allows the tube walls to be divided into several layers in the radial direction and several sectors in the circumferential direction, to account for heat transfer conditions in these two directions. The CATHENA code with the generalized heat transfer package is capable of capturing key pool-boiling phenomena such as nucleate, transition and film boiling heat transfer as well as an ability to model the rewet phenomenon to some extent. A CATHENA input model was generated and used in simulations of selected contact boiling experiment test cases. The transient wall temperatures have been calculated in different portions of the calandria tube. By using this model an adequate agreement was achieved between CATHENA calculation and experimental measurement The CATHENA code enables one to investigate the transient and local thermal-mechanical behaviour of the calandria tube

Development of containers sealing system like part of surveillance program of the vessel in nuclear power plants

International Nuclear Information System (INIS)

Romero C, J.; Hernandez C, R.; Fernandez T, F.; Rocamontes A, M.; Perez R, N.

2009-10-01

The owners of nuclear power plants should be demonstrate that the embrittlement effects by neutronic radiation do not commit the structural integrity from the pressure vessel of nuclear reactors, during conditions of routine operation and below postulate accident. For this reason, there are surveillance programs of vessels of nuclear power plants, in which are present surveillance capsules. A surveillance capsule is compound by the support, six containers for test tubes and dosimeters. The containers for test tubes are of two types: rectangular container for test tubes, Charpy V and Cylindrical Container for tension test tubes. These test tubes are subject to a same or bigger neutronic flow to that of vessel, being representative of vessel mechanical conditions. The test tubes are rehearsed to watch over the increase of embrittlement that presents the vessel. This work describes the development of welding system to seal the containers for test tubes, these should be filled with helium of ultra high purity, to a pressure of an atmosphere. In this system the welding process Gas Tungsten Arc Welding is used, a hermetic camera that allows to place the containers with three grades of freedom, a vacuum subsystem and pressure, high technology equipment's like: power source with integrated computer, arc starter of high frequency, helium flow controller, among others. Finally, the advances in the inspection system for the qualification of sealing system are mentioned, system that should measure the internal pressure of containers and the helium purity inside these. (Author)

STEFINS: a steel freezing integral simulation program

International Nuclear Information System (INIS)

Frank, M.V.

1980-09-01

STEFINS (STEel Freezing INtegral Simulation) is a computer program for the calculation of the rate of solidification of molten steel on solid steel. Such computations arize when investigating core melt accidents in fast reactors. In principle this problem involves a coupled two-dimensional thermal and hydraulic approach. However, by physically reasonable assumptions a decoupled approach has been developed. The transient solidification of molten steel on a cold wall is solved in the direction normal to the molten steel flow and independent from the solution for the molten steel temperature and Nusselt number along the direction of flow. The solutions to the applicable energy equations have been programmed in cylindrical and slab geometries. Internal gamma heating of steel is included

Improvement of pump tubes for gas guns and shock tube drivers

Science.gov (United States)

Bogdanoff, D. W.

1990-01-01

In a pump tube, a gas is mechanically compressed, producing very high pressures and sound speeds. The intensely heated gas produced in such a tube can be used to drive light gas guns and shock tubes. Three concepts are presented that have the potential to allow substantial reductions in the size and mass of the pump tube to be achieved. The first concept involves the use of one or more diaphragms in the pump tube, thus replacing a single compression process by multiple, successive compressions. The second concept involves a radical reduction in the length-to-diameter ratio of the pump tube and the pump tube piston. The third concept involves shock heating of the working gas by high explosives in a cyclindrical geometry reusable device. Preliminary design analyses are performed on all three concepts and they appear to be quite feasible. Reductions in the length and mass of the pump tube by factors up to about 11 and about 7, respectively, are predicted, relative to a benchmark conventional pump tube.

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

Role of innovative institutional structures in integrated governance. A case study of integrating health and nutrition programs in Chhattisgarh, India.

Science.gov (United States)

Kalita, Anuska; Mondal, Shinjini

2012-01-01

The aim of this paper is to highlight the significance of integrated governance in bringing about community participation, improved service delivery, accountability of public systems and human resource rationalisation. It discusses the strategies of innovative institutional structures in translating such integration in the areas of public health and nutrition for poor communities. The paper draws on experience of initiating integrated governance through innovations in health and nutrition programming in the resource-poor state of Chhattisgarh, India, at different levels of governance structures--hamlets, villages, clusters, blocks, districts and at the state. The study uses mixed methods--i.e. document analysis, interviews, discussions and quantitative data from facilities surveys--to present a case study analyzing the process and outcome of integration. The data indicate that integrated governance initiatives improved convergence between health and nutrition departments of the state at all levels. Also, innovative structures are important to implement the idea of integration, especially in contexts that do not have historical experience of such partnerships. Integration also contributed towards improved participation of communities in self-governance, community monitoring of government programs, and therefore, better services. As governments across the world, especially in developing countries, struggle towards achieving better governance, integration can serve as a desirable process to address this. Integration can affect the decentralisation of power, inclusion, efficiency, accountability and improved service quality in government programs. The institutional structures detailed in this paper can provide models for replication in other similar contexts for translating and sustaining the idea of integrated governance. This paper is one of the few to investigate innovative public institutions of a and community mobilisation to explore this important, and under

Production of ceramic-metal joints for high-vacuum applications and development of simulation program for discharge tube

Energy Technology Data Exchange (ETDEWEB)

Kang, S. H.; Chung, K. H. [Seoul National University, Seoul (Korea)

2000-04-01

To develop a ceramic-metal jointed tube for high-vacuum applications, metalizing process and active metal brazing were investigated. Active metal brazing was adopted as a joining process to produce a high-vacuum tube which had high joint strength and reliability. A possibility for the development of new composition of Mo-Mn paste was studied. Also, to improve the strength and reliability of active metal brazed joint, TiN coating was introduced as a diffusion barrier. It was revealed that TiN coating could improve the joint strength and reliability. 100mm {phi} tube joint was produced using incusil ABA brazing alloy. The strength and reliability of manufactured tube showed higher value than commercial one. The electric field distribution in ceramic tube under high voltage was analyzed. Two dimensional electric field distribution was investigated under the existence of charged particles. From this result, electric field distribution at the surface of ceramic tube and the location of high electric field was predicted. Finally, Arc discharge was simulated to analyze the effect of arc discharge on the discharge tube wall. The maximum temperature of arc was 12000-13000K. The wall temperature was increased 100-170K by the arc discharge. 45 refs., 57 figs., 4 tabs. (Author)

Inelastic analysis of finite length and depth cracked tubes

International Nuclear Information System (INIS)

Reich, M.; Gardner, D.; Prachuktam, S.; Chang, T.Y.

1977-01-01

Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdowns. This paper concerns itself with the prediction of the failure pressures for typical PWR steam generator tubes with longitudinal finite length and finite depth cracks. Only local plastic overload failure is considered since the material is non-notch sensitive. Non-linear finite element analyses are carried out to determine the burst pressures of steam generator tubes containing longitudinal cracks located on the outer surface of the tubes. The non-linearities considered herein include elastic-plastic material behavior and large deformations. A non-proprietary general purpose non-linear finite element program, NFAP was adopted for the analysis. Due to the asymmetric nature of the cracks, two-dimensional, as well as three-dimensional finite element analyses, were performed. The two-dimensional element and its formulations are similar to those of NONSAP. The three-dimensional isoparametric element with elastic-plastic material characteristics together with the large deformation formulations used in NFAP are described in the Report BNL-20684. The numerical accuracy of the program was investigated and checked with known solutions of benchmark problems. In addition to the three-dimensional element which was specifically inserted into NFAP for this problem, other features such as direct pressure inputs for isoparametric elements, automatic load increment adjustments for convergent non-linear solutions, and automatic bandwidth reduction schemes are incorporated into the program thus allowing for a more economical evaluation of three-dimensional inelastic analysis. In summary the analysis clearly shows that for short cracks axial effects play a significant role. For long cracks, they are not important since two-dimensional conditions predominate and failure is governed by circumferential or hoop stress conditions

Status of the tube elongation problem as of June 1976

International Nuclear Information System (INIS)

Alexander, W.K.

1976-01-01

It was discovered in May of 1971 that the N Reactor process tubes had apparently increased in length by as much as one inch. Preliminary observations and measurements led to the tentative conclusion that this observed elongation was linear with accumulated tube exposure and also that it was related in some manner to the tube fabrication process. It appeared that the observed elongation was approximately proportional to the degree of cold work retained in the finished tubes. This latter conclusion was based on the observation that those tubes with approximately 17-18 percent cold work had elongated only about half as much as the standard 30-percent-cold-worked tubes. It was immediately recognized that if such elongation was to continue unchecked it could pose a limit to reactor life since total possible tube expansion, from all causes, is limited to 1.75 inches by nozzle design considerations as shown in Figure 1. Thermal and hydraulic expansion were calculated to total approximately 0.75 inches which left only one inch available to accommodate tube growth or creep. Since discovery of this phenomenon, an extensive measurements program has been carried out to evaluate the extent and rate of tube elongation. Two corrective approaches have been developed and a small number of tubes were modified by each method during the 1976 summer outage. During the 1974, 1975 and 1976 Summer Outages, measurements were made on all tubes to determine the clearance remaining between the nozzle keys and the gas packing ring. These readings not only give an overall picture of the extent of elongation, but also provide immediate data indicating which tubes are about out of clearance. The report presents an evaluation of the measurements taken to date

The Eclipse system for surveying the guide tubes of control rod clusters

International Nuclear Information System (INIS)

Pace, Y.M.

2008-01-01

Eclipse is a new system developed by Areva to assess the wear of the guide tubes of control rod clusters. This system is based on the projection of a shadow on a light plan in order to record the profile and the internal diameter of a hollow tube. This system allows us to quantify the wear and it can be included in a program dedicated to monitor the wear and master its kinetics. This system has been validated on the guide tubes from the Ringhals units. (A.C.)

DITTY - a computer program for calculating population dose integrated over ten thousand years

International Nuclear Information System (INIS)

Napier, B.A.; Peloquin, R.A.; Strenge, D.L.

1986-03-01

The computer program DITTY (Dose Integrated Over Ten Thousand Years) was developed to determine the collective dose from long term nuclear waste disposal sites resulting from the ground-water pathways. DITTY estimates the time integral of collective dose over a ten-thousand-year period for time-variant radionuclide releases to surface waters, wells, or the atmosphere. This document includes the following information on DITTY: a description of the mathematical models, program designs, data file requirements, input preparation, output interpretations, sample problems, and program-generated diagnostic messages

The Transuranic Waste Program's integration and planning activities and the contributions of the TRU partnership

International Nuclear Information System (INIS)

Harms, T.C.; O'Neal, W.; Petersen, C.A.; McDonald, C.E.

1994-02-01

The Technical Support Division, EM-351 manages the integration and planning activities of the Transuranic Waste Program. The Transuranic Waste Program manager provides transuranic waste policy, guidance, and issue resolution to Headquarters and the Operations Offices. In addition, the program manager is responsible for developing and implementing an integrated, long-range waste management plan for the transuranic waste system. A steering committee, a core group of support contractors, and numerous interface working groups support the efforts of the program manager. This paper provides an overview of the US Department of Energy's transuranic waste integration activities and a long-range planning process that includes internal and external stakeholder participation. It discusses the contributions and benefits provided by the Transuranic Partnership, most significantly, the integration activities and the body of data collected and assembled by the Partnership

Integrated Guidelines for Management of Alloy 600 Locations

Energy Technology Data Exchange (ETDEWEB)

Na, Kyung-Hwan; Chung, Hansub; Yang, Jun-Seog; Lee, Kyoung-Soo [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

2015-05-15

The locations experiencing PWSCC include steam generator tubes, pressurizer instrumental nozzles, control rod driving mechanism(CRDM) penetration nozzles, reactor outlet nozzles, and bottom mounted instrumental(BMI) nozzles. Korea Hydro and Nuclear Power Co.(KHNP) has developed integrated guidelines for management of alloy 600 locations and the guidelines are under review by the regulator. The guidelines consist of alloy 600 location database, inspection program, maintenance/preventive maintenance method, and finally water chemistry management for PWSCC mitigation. In this paper, the detailed contents are presented. The integrated guidelines collected all relevant information on the management of alloy 600 locations. This information may be useful for establishing the most effective preventive maintenance strategies by prioritization in addition to maintenance strategies. Table II summarize maintenance strategies for alloy 600 locations.

Integrated modeling and characterization of local crack chemistry

International Nuclear Information System (INIS)

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

1996-01-01

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

Integrated modeling and characterization of local crack chemistry

International Nuclear Information System (INIS)

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

1995-01-01

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

Bender/Coiler for Tubing

Science.gov (United States)

Stoltzfus, J. M.