The development of octagon Zr-4 alloy tube for heating reactors

International Nuclear Information System (INIS)

Yang Fanglin; Yang Yingli; Wang Guangshen

1989-10-01

The asymmetrical octagon Zr-4 alloy tubes which are used for fuel assembly in the heating reactor have been developed. The thickness of tube wall is 1.5 mm and the length is 1725 mm. The long side of the octagon is 138.7 0.3 +0.2 mm, the short side is 93.1 ± 0.1 mm. To manufacture these tubes a stretch draw forming processing method is adopted. The process is divided into two phases. In the first phase, a short draw mould is used to stretch the Zr-4 alloy tube. In the second phase, a long draw mould, its length is equal to the end-produt length, is used to complete the final processing. The size accuracy and repeatability of this method are excellent and can fully meet the design requirements

Process for forming seamless tubing of zirconium or titanium alloys from welded precursors

International Nuclear Information System (INIS)

Sabol, G.P.; Barry, R.F.

1987-01-01

A process is described for forming seamless tubing of a material selected from zirconium, zirconium alloys, titanium, and titanium alloys, from welded precursor tubing of the material, having a heterogeneous structure resulting from the welding thereof. The process consists of: heating successive axial segments of the welded tubing, completely through the wall thereof, including the weld, to uniformly transform the heterogeneous, as welded, material into the beta phase; quenching the beta phase tubing segments, the heating and quenching effected sufficiently rapid enough to produce a fine sized beta grain structure completely throughout the precursor tubing, including the weld, and to prevent growth of beta grains within the material larger than 200 micrometers in diameter; and subsequently uniformly deforming the quenched precursor tubing by cold reduction steps to produce a seamless tubing of final size and shape

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Science.gov (United States)

Shingledecker, John

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

Characterization of Tubing from Advanced ODS alloy (FCRD-NFA1)

International Nuclear Information System (INIS)

Maloy, Stuart Andrew; Aydogan, Eda; Anderoglu, Osman; Lavender, Curt; Anderson, Iver; Rieken, Joel; Lewandowski, John; Hoelzer, Dave; Odette, George R.

2016-01-01

Fabrication methods are being developed and tested for producing fuel clad tubing of the advanced ODS 14YWT and FCRD-NFA1 ferritic alloys. Three fabrication methods were based on plastically deforming a machined thick-wall tube sample of the ODS alloys by pilgering, hydrostatic extrusion or drawing to decrease the outer diameter and wall thickness and increase the length of the final tube. The fourth fabrication method consisted of the additive manufacturing approach involving solid-state spray deposition (SSSD) of ball milled and annealed powder of 14YWT for producing thin-wall tubes. Of the four fabrication methods, two methods were successful at producing tubing for further characterization: production of tubing by high-velocity oxy-fuel spray forming and production of tubing using high-temperature hydrostatic extrusion. The characterization described shows through neutron diffraction the texture produced during extrusion while maintaining the beneficial oxide dispersion. In this research, the parameters for innovative thermal spray deposition and hot extrusion processing methods have been developed to produce the final nanostructured ferritic alloy (NFA) tubes having approximately 0.5 mm wall thickness. Effect of different processing routes on texture and grain boundary characteristics has been investigated. It was found that hydrostatic extrusion results in combination of plane strain and shear deformations which generate rolling textures of ?- and ?-fibers on and together with a shear texture of ?-fiber on and . On the other hand, multi-step plane strain deformation in cross directions leads to a strong rolling textures of ?- and ?-fiber on together with weak ?-fiber on . Even though the amount of the equivalent strain is similar, shear deformation leads to much lower texture indexes compared to the plane strain deformations. Moreover, while 50% of hot rolling brings about a large number of high-angle grain boundaries (HAB), 44% of shear deformation results

Characterization of Tubing from Advanced ODS alloy (FCRD-NFA1)

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lavender, Curt [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Anderson, Iver [Ames Lab., Ames, IA (United States); Rieken, Joel [Ames Lab., Ames, IA (United States); Lewandowski, John [Case Western Reserve Univ., Cleveland, OH (United States); Hoelzer, Dave [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Odette, George R. [Univ. of California, Santa Barbara, CA (United States)

2016-09-20

Fabrication methods are being developed and tested for producing fuel clad tubing of the advanced ODS 14YWT and FCRD-NFA1 ferritic alloys. Three fabrication methods were based on plastically deforming a machined thick-wall tube sample of the ODS alloys by pilgering, hydrostatic extrusion or drawing to decrease the outer diameter and wall thickness and increase the length of the final tube. The fourth fabrication method consisted of the additive manufacturing approach involving solid-state spray deposition (SSSD) of ball milled and annealed powder of 14YWT for producing thin-wall tubes. Of the four fabrication methods, two methods were successful at producing tubing for further characterization: production of tubing by high-velocity oxy-fuel spray forming and production of tubing using high-temperature hydrostatic extrusion. The characterization described shows through neutron diffraction the texture produced during extrusion while maintaining the beneficial oxide dispersion. In this research, the parameters for innovative thermal spray deposition and hot extrusion processing methods have been developed to produce the final nanostructured ferritic alloy (NFA) tubes having approximately 0.5 mm wall thickness. Effect of different processing routes on texture and grain boundary characteristics has been investigated. It was found that hydrostatic extrusion results in combination of plane strain and shear deformations which generate rolling textures of α- and γ-fibers on {001}<110> and {111}<110> together with a shear texture of ζ-fiber on {011}<211> and {011}<011>. On the other hand, multi-step plane strain deformation in cross directions leads to a strong rolling textures of θ- and ε-fiber on {001}<110> together with weak γ-fiber on {111}<112>. Even though the amount of the equivalent strain is similar, shear deformation leads to much lower texture indexes compared to the plane strain deformations. Moreover, while 50% of hot rolling brings about a large number of

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

International Nuclear Information System (INIS)

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

2015-01-01

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

The resistance to PWSCC of explosively expanded Alloy 600 tube-to-tubesheet joints

International Nuclear Information System (INIS)

Gold, R.E.; Pement, F.W.; Tarabek, S.A.; Economy, G.

1992-01-01

Experimental evaluations were performed to determine the approximate magnitude of the residual stresses associated with explosively expanded steam generator tubing, and to assess the resistance to primary water stress corrosion cracking (PWSCC) of these expansions. Indexing of residual stresses was performed by means of magnesium chloride exposures of surrogate stainless steel mockups. The PWSCC resistance was evaluated by the testing of pressurized mockups of explosively expanded mill annealed Alloy 600 tubing in a highly accelerated Alloy 600 tubing in a highly accelerated steam test environment. Shot peening of the inside tube surfaces was demonstrated to be effective in modifying the residual stresses, providing additional resistance to PWSCC

Bottom nozzle to guide tube connection

International Nuclear Information System (INIS)

Bryan, W.J.

1991-01-01

This patent describes a nuclear fuel assembly which includes an upper end fitting and a lower end fitting spaced therefrom and connected thereto by elongated guide tubes of one alloy having an open upper end and a closed lower end with spaced fuel element retaining grids mounted on the guide tubes therebetween, the closed lower ends of the guide tubes including a threaded central passageway and the attachment of the guide tubes to the lower end fitting of another alloy. It comprises: an externally threaded bolt with a first end threadably received in the threaded central passageway of the lower end of the guide tube and a head at the other end of the side of the lower end fitting opposite the guide tube; an interruption in the external threads of the bolt which forms a groove which communicates the interior of the guide tube with the side of the lower end fitting opposite the guide tube and enhances its frictional engagement with the threaded central passageway, thereby to hold and attach the guide tube and lower end fitting firmly together, even through a series of temperature cycles

A Study on Corrosion and Fretting Wear Resistance of Alloy 690 Tubes

Energy Technology Data Exchange (ETDEWEB)

Won, Ju Jin; Min, Su Jung; Kim, Myeong Su; Kim, Kyu Tae [Dongguk Univ., Gyeongju (Korea, Republic of)

2013-10-15

In this article, the effects of such failures have on the materials of alloy 690 are assessed. The corroded volume variation and mass decreased continuously with time. However, the oxide volume changes in an irregular pattern since the oxide formed on the alloy 690 metal may be detached due to the flake formation. The amount of the fretting wear increased with time. It can be seen that the wear rate increased with time and reduced at the later time. The test results show that the ductility decreased as corrosion increases. Alloy 690 is broadly used as a material of nuclear power plant's steam generator tubes because of its excellent mechanical strength, corrosion properties, wear properties and stability at a high temperature. However, the tubes for nuclear power plant's steam generators become a major threat for lifetime management and efficient operation of nuclear power plant due to various corrosion and fretting wear failures caused by flow-induced vibration (FIV) that occurs between tubes.

Simulation studies on Tube End Expansion of AA2014 Alloy Tubes

Science.gov (United States)

Venugopal, L.; Prasad, N. E. C.; Geeta Krishna, P.; Praveen, L.

2018-03-01

End forming is defined as forming the end of tubular forms either by inverting the tube or by expanding it. It finds application in many fields such as in automotive and aerospace sectors as power transmission elements, fuel lines, exhaust pipes etc. The main aim of the present work is to expand the AA2014 alloy tubes with different die sets without any fracture. Deform 2D software was used for performing simulations on expanding the tubes with different die set (punch) values having differed forming angles (α = 15°, 30° and 45°) and expansion ratios (rp/r0 = 1.39, 1.53 and 1.67). Experiments were also conducted and the results correlate with the simulation results. The results shows that for the punch having less cone angle (α) values the linear displacement is more rather than higher cone angles. But in the case of higher cone angles the radial displacement is more than the linear displacement.

A Study On Critical Thinning In Thin-walled Tube Bending Of Al-Alloy 5052O Via Coupled Ductile Fracture Criteria

International Nuclear Information System (INIS)

Li Heng; Yang He; Zhan Mei

2010-01-01

Thin-walled tube bending(TWTB) method of Al-alloy tube has attracted wide applications in aerospace, aviation and automobile,etc. While, under in-plane double tensile stress states at the extrados of bending tube, the over-thinning induced ductile fracture is one dominant defect in Al-alloy tube bending. The main objective of this study is to predict the critical wall-thinning of Al-alloy tube bending by coupling two ductile fracture criteria(DFCs) into FE simulation. The DFCs include Continuum Damage Mechanics(CDM)-based model and GTN porous model. Through the uniaxial tensile test of the curved specimen, the basic material properties of the Al-alloy 5052O tube is obtained; via the inverse problem solution, the damage parameters of both the two fracture criteria are interatively determined. Thus the application study of the above DFCs in the TWTB is performed, and the more reasonable one is selected to obtain the critical thinning of Al-alloy tube in bending. The virtual damage initiation and evolution (when and where the ductile fracture occurs) in TWTB are investigated, and the fracture mechanisms of the voided Al-alloy tube in tube bending are consequently discussed.

Forming of Zr-4 alloy guide tube with varied diameters

International Nuclear Information System (INIS)

Wei Songyan; Tian Zhenye

1989-10-01

A new built-up mould method to manufacture Zr-4 alloy guide tubes with varied diameters at the middle of tube is introduced. The guide tube is used in nuclear power plants for guiding the control rods. This method has many advantages such as simple in forming, low cost of manufacturing, no need of special devices and favour of batch processing. The test results show that the accuracy of size, mechanical properties, resistance to corrosion, grain size and hydrogenate orientation of the end-products can meet the technical needs for nuclear reactor operation

Polycrystalline models for the calculation of residual stresses in zirconium alloys tubes

International Nuclear Information System (INIS)

Signorelli, J.W.; Turner, P.A.; Lebensohn, R.A.; Pochettino, A.A.

1995-01-01

Tubes made of different Zirconium alloys are used in various types of reactors. The final texture of tubes as well as the distribution of residual stresses depend on the mechanical treatments done during their manufacturing process. The knowledge and prediction of both the final texture and the distribution of residual stresses in a tube for nuclear applications are of outstanding importance in relation with in-reactor performance of the tube, especially in what concerns to its irradiation creep and growth behaviour. The viscoplastic and the elastoplastic self consistent polycrystal models are used to investigate the influence of different mechanical treatments, performed during rolling processes on the final distribution of intergranular residual stresses of zirconium alloys tubes. The residual strains predictions with both formulations show a non linear dependence with the orientation, but they are qualitatively different. This discrepancy could be explain in terms of the relative plastic activity between the -type and -type deformation modes predicted with the viscoplastic and elastoplastic models. (author). 10 refs., 4 figs., 1 tab

Rejuvenation of service exposed ammonia cracker tubes of cast Alloy 625 and their re-use

Energy Technology Data Exchange (ETDEWEB)

Singh, J.B., E-mail: jbsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Verma, A. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Jaiswal, D.M.; Kumar, N.; Patel, R.D. [Heavy Water Board, Department of Atomic Energy, Anushakti Nagar, Mumbai 400094 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

2015-09-17

This study is an extension of a previous study undertaken to rejuvenate ammonia cracker tubes of Alloy 625 alloy that have been service exposed in heavy water plants for their full service life of 100,000 h. The service exposure caused significant microstructural modifications and deterioration in mechanical properties, and a solution annealing treatment of 2 h at 1160 °C rejuvenated all properties similar to those of the virgin alloy. The present study reports the evolution of microstructure and mechanical properties of a full service exposed centrifugally cast Alloy 625 tube that was put into service again for 55,000 h after receiving a rejuvenation treatment. During the second service, microstructural modifications, increase in strength and loss of ductility were on the lines of the work reported earlier. However, it was encouraging to observe that degraded properties after the second service life remained within the bounds of those of virgin and full service exposed tubes. The good performance of the rejuvenated tube during the second service life has been attributed to good control of operation parameters that limited the precipitation of grain boundary carbides during the first service life, which otherwise would have had a direct bearing on premature failure of tubes during their second service life.

Externally fired gas turbine cycles with high temperature heat exchangers utilising Fe-based ODS alloy tubing

International Nuclear Information System (INIS)

Olsson, F.; Svensson, S.-A.; Duncan, R.

2001-01-01

This work is part of the BRITE / EuRAM Project 'Development of Torsional Grain Structures to Improve Biaxial Creep Performance of Fe-based ODS Alloy Tubing for Biomass Power Plant'. The main goal of this project is to heat exchanger tubes working at 1100 o C and above. The paper deals with design implications of a biomass power plant, using an indirectly fired gas turbine with a high temperature heat exchanger containing Fe-based ODS alloy tubing. In the current heat exchanger design, ODS alloy tubing is used in a radiant section, using a bayonet type tube arrangement. This enables the use of straight sections of ODS tubing and reduces the amount of material required. In order to assess the potential of the power plant system, thermodynamic calculations have been conducted. Both co-generation and condensing applications are studied and results so far indicate that the electrical efficiency is high, compared to values reached by conventional steam cycle power plants of the same size (approx. 5 MW e ). (author)

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

Science.gov (United States)

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

2018-01-02

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

Thermal creep behavior of N36 zirconium alloy cladding tube

International Nuclear Information System (INIS)

Wang, P.; Zhao, W.; Dai, X.

2015-01-01

N36 is an alloy containing Zr, Sn, Nb and Fe that is developed by China as a superior cladding material to meet the performance of PWR fuel assembly at the maximum fuel rod burn-up. The creep characteristics of N36 zirconium alloy cladding tube were investigated at temperature from 593 K to 723 K with stress ranging from 20 MPa to 160 MPa. Transitions in creep mechanisms were noted, showing the distinct three rate-controlled creep mechanisms for the alloy at test conditions. In the region of low stresses with stress exponent n ∼ 1 and activation energy Q ∼ (104±4) kJ.mol -1 , Coble creep, based on diffusion of materials through grain boundaries, is the dominant rate-controlling mechanism, which contributes to the creep deformation. The formation of slip bands acts as an accommodation mechanism. In the region of middle stress with stress exponent n ∼ 3 and activation energy Q ∼ (195±7) kJ.mol -1 , micro-creep, caused by viscous gliding of dislocations due to the interaction of O atoms with dislocations, controls the deformation. In the high stress region with stress exponent n ∼ 5-6 and activation energy Q ∼ (210±10) kJ.mol -1 , two mechanisms of the climb of edge dislocations (EDC) and the motion of jogged screw dislocation (MJS) contribute to rate controlling process. In test conditions N36 alloy cladding tube behaves a type of creep similar to that noted in class-I (A) alloys

Effect of flame-tube head structure on combustion chamber performance

Science.gov (United States)

Gu, Minqqi

1986-01-01

The experimental combustion performance of a premixed, pilot-type flame tube with various head structures is discussed. The test study covers an extensive area: efficiency of the combustion chamber, quality of the outlet temperature field, limit of the fuel-lean blowout, ignition performance at ground starting, and carbon deposition. As a result of these tests, a nozzle was found which fits the premixed pilot flame tube well. The use of this nozzle optimized the performance of the combustion chamber. The tested models had premixed pilot chambers with two types of air-film-cooling structures, six types of venturi-tube structures, and secondary fuel nozzles with two small spray-cone angles.

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

International Nuclear Information System (INIS)

Potz, F.; Bohmann, W.

1998-01-01

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

Evaluation of precipitates used in strainer head loss testing: Part II. Precipitates by in situ aluminum alloy corrosion

International Nuclear Information System (INIS)

Bahn, Chi Bum; Kasza, Ken E.; Shack, William J.; Natesan, Ken; Klein, Paul

2011-01-01

Graphical abstract: Display Omitted Research highlights: → Sump strainer head loss testing to evaluate chemical effects. → Aluminum hydroxide precipitates by in situ Al alloy corrosion caused head loss. → Intermetallic particles released from Al alloy can also cause significant head loss. → When evaluating Al effect on head loss, intermetallics should be considered. - Abstract: Vertical loop head loss tests were performed with 6061 and 1100 aluminum (Al) alloy plates immersed in borated solution at pH = 9.3 at room temperature and 60 o C. The results suggest that the potential for corrosion of an Al alloy to result in increased head loss across a glass fiber bed may depend on its microstructure, i.e., the size distribution and number density of intermetallic particles that are present in Al matrix and FeSiAl ternary compounds, as well as its Al release rate. Per unit mass of Al removed from solution, the WCAP-16530 aluminum hydroxide (Al(OH) 3 ) surrogate was more effective in increasing head loss than the Al(OH) 3 precipitates formed in situ by corrosion of Al alloy. However, in choosing a representative amount of surrogate for plant specific testing, consideration should be given to the potential for additional head losses due to intermetallic particles and the apparent reduction in the effective solubility of Al(OH) 3 when intermetallic particles are present.

Crack growth rates in vessel head penetration materials

International Nuclear Information System (INIS)

Gomez Briceno, D.; Lapena, J.; Blazquez, F.

1994-01-01

The cracks detected in reactor vessel head penetrations in certain European plants have been attributed to Primary Water Stress Corrosion Cracking (PWSCC). The penetrations in question are made from Inconel 600. The susceptibility of this alloy to PWSCC has been widely studied in relation to use of this material for steam generator tubes. When the first reactor vessel head penetration cracks were detected, most of the available data on crack propagation rates were from test specimens made from steam generator tubes and tested under conditions that questioned the validity of these data for assessment of the evolution of cracks in penetrations. For this reason, the scope of the Spanish Research Project on the Inspection and Repair of PWR reactor vessel head penetrations included the acquisition of data on crack propagation rates in Inconel 600, representative of the materials used for vessel head penetrations. (authors). 1 fig., 2 tabs., 6 refs

Influence of tube spinning on formability of friction stir welded aluminum alloy tubes for hydroforming application

Energy Technology Data Exchange (ETDEWEB)

Wang, X.S. [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Hu, Z.L., E-mail: zhilihuhit@163.com [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Hubei Key Laboratory of Advanced Technology of Automobile Parts, Wuhan University of Technology, Wuhan 430070 (China); State Key Laboratory of Materials Processing and Die and Mould Technology, Huazhong University of Science and Technology (China); Yuan, S.J. [State Key Laboratory of Advanced Welding and Joining, Harbin Institute of Technology, Harbin 150001 (China); Hua, L. [Hubei Key Laboratory of Advanced Technology of Automobile Parts, Wuhan University of Technology, Wuhan 430070 (China)

2014-06-01

Due to economic and ecological reasons, the application of tailor-welded blanks of aluminum alloy has gained more and more attention in manufacturing lightweight structures for automotives and aircrafts. In the study, the research was aimed to highlight the influence of spinning on the formability of FSW tubes. The microstructural characteristics of the FSW tubes during spinning were studied by electron backscattered diffraction (EBSD) and transmission electron microscopy (TEM). The formability of the FSW tubes with different spinning reduction was assessed by hydraulic bulge test. It is found that the spinning process shows a grain refinement of the tube. The grains of the FSW tube decrease with increasing thickness reduction, and the effect of grain refinement is more obvious for the BM compared to that of the weld. The difference of grain size and precipitates between the weld and BM leads to an asymmetric W-type microhardness distribution after spinning. The higher thickness reduction of the tube, the more uniform distribution of grains and precipitates it shows, and consequently results in more significant increase of strength. As compared with the result of tensile test, the tube after spinning shows better formability when the stress state changes from uniaxial to biaxial stress state.

Head losses prediction and analysis in a bulb turbine draft tube under different operating conditions using unsteady simulations

Science.gov (United States)

Wilhelm, S.; Balarac, G.; Métais, O.; Ségoufin, C.

2016-11-01

Flow prediction in a bulb turbine draft tube is conducted for two operating points using Unsteady RANS (URANS) simulations and Large Eddy Simulations (LES). The inlet boundary condition of the draft tube calculation is a rotating two dimensional velocity profile exported from a RANS guide vane- runner calculation. Numerical results are compared with experimental data in order to validate the flow field and head losses prediction. Velocity profiles prediction is improved with LES in the center of the draft tube compared to URANS results. Moreover, more complex flow structures are obtained with LES. A local analysis of the predicted flow field using the energy balance in the draft tube is then introduced in order to detect the hydrodynamic instabilities responsible for head losses in the draft tube. In particular, the production of turbulent kinetic energy next to the draft tube wall and in the central vortex structure is found to be responsible for a large part of the mean kinetic energy dissipation in the draft tube and thus for head losses. This analysis is used in order to understand the differences in head losses for different operating points. The numerical methodology could then be improved thanks to an in-depth understanding of the local flow topology.

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

Institute of Scientific and Technical Information of China (English)

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

2017-01-01

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

Observations and insights into Pb-assisted stress corrosion cracking of alloy 600 steam generator tubes

International Nuclear Information System (INIS)

Thomas, L.; Bruemmer, Stephen M.

2005-01-01

Pb-assisted stress-corrosion cracking (PbSCC) of Alloy 600 steam-generator tubing in high-temperature-water service and laboratory tests were studied by analytical transmission electron microscopy of cross-sectioned samples. Examinations of pulled tubes from many pressurized water reactors revealed lead in cracks from 11 of 17 samples. Comparisons of the degraded intergranular structures with ones produced in simple laboratory tests with PbO in near-neutral AVT water showed that the PbSCC characteristics in service tubing could be reproduced without complex chemistries and heat-flow conditions that can occur during plant operation. Observations of intergranular and transgranular cracks promoted by Pb in the test samples also provided new insights into the mechanisms of PbSCC in mill-annealed and thermally treated Alloy 600

SCC analysis of Alloy 600 tubes from a retired steam generator

Science.gov (United States)

Hwang, Seong Sik; Kim, Hong Pyo

2013-09-01

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

Effects of Oxidation and fractal surface roughness on the wettability and critical heat flux of glass-peened zirconium alloy tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Nitheanandan, T.; Bullock, C.D.; Slater, L.F.; McRae, G.A.

2003-05-01

Glass-bead peening the outside surfaces of zirconium alloy tubes has been shown to increase the Critical Heat Flux (CHF) in pool boiling of water. The CHF is found to correlate with the fractal roughness of the metal tube surfaces. In this study on the effect of oxidation on glass-peened surfaces, test measurements for CHF, surface wettability and roughness have been evaluated using various glass-peened and oxidized zirconium alloy tubes. The results show that oxidation changes the solid-liquid contact angle (i.e., decreases wettability of the metal-oxide surface), but does not change the fractal surface roughness, appreciably. Thus, oxidation of the glass-peened surfaces of zirconium alloy tubes is not expected to degrade the CHF enhancement obtained by glass-bead peening. (author)

Delayed hydride cracking behavior of Zr-2.5Nb alloy pressure tubes for PHWR700

Energy Technology Data Exchange (ETDEWEB)

Sunil, S.; Bind, A.K.; Khandelwal, H.K.; Singh, R.N., E-mail: rnsingh@barc.gov.in; Chakravartty, J.K.

2015-11-15

In order to attain improved in-reactor performance few prototypes pressure tubes of Zr-2.5Nb alloy were manufactured by employing forging to break the cast structure and to obtain more homogeneous microstructure. Both double forging and single forging were employed. The forged material was further processed by employing hot extrusion, cold pilgering and autoclaving. A detailed characterization in terms of mechanical properties and microstructure of the prototype tubes were carried for qualifying it for intended use as pressure tubes in PHWR700 reactors. In this work, Delayed Hydride Cracking (DHC) behavior of the forged Zr-2.5Nb pressure tube material characterized in terms of DHC velocity and threshold stress intensity factor associated with DHC (K{sub IH}) was compared with that of conventionally manufactured material in the temperature range of 200–283 °C. Activation energy associated with the DHC in this alloy was found to be ∼60 kJ/mol for the forged materials.

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

Tensile properties of Zr-2.5 Nb pressure tube alloy between 25 and 800 degC

International Nuclear Information System (INIS)

Singh, R.N.; Kishore, R.; Sinha, T.K.; Banerjee, S.

2000-10-01

Tensile properties of zirconium-2.5 wt. % niobium pressure tube material were evaluated by uniaxial tension tests at temperatures between 25 and 800 degC and under strain-rates varying from 3.3 x 10 -5 to 3.3 x 10 -3 /s. Tests were carried out on specimens fabricated from the sections of finished (autoclaved) tubes as well as on those machined from the sections of cold worked (2 nd pilgered) tubes. Moreover, specimens fabricated from finished tubes belonging to twenty different heats were tested at 300 degC to study the heat to heat variation in tensile properties of this alloy. In order to study the effect of the crystallographic texture on the tensile properties, specimens oriented in longitudinal as well as, in transverse directions of the tubes were also tested. Results showed that both yield and ultimate tensile strengths of this alloy decreased monotonically with increasing test temperatures, with a rapid fall in strengths above a temperature of 350 degC (623 K). The tensile ductility did not change appreciably up to 400 degC (673K) but increased rapidly above this temperature. The observed results on the temperature dependence of the strength and ductility indicated the possible occurrence of dynamic strain-ageing in this alloy in the temperature range of 200-300 degC (473 to 573 K). The transverse specimens showed higher strengths and lower ductility as compared to those of the longitudinal specimens up to a temperature of 350 degC (623 K). Above 350 degC, the difference in the strengths and the ductility of the two types of the specimens, became negligibly small indicating that the texture did not appreciably influence the tensile properties of this alloy at temperatures exceeding 350 degC. The alloy developed extensive superplasticity (ductility exceeding 100 %), when tested in the temperature range of 650-800 degC. Maximum ductility values of 650 % for longitudinal and 900 % for the transverse orientation with strain-rate sensitivity (m) exceeding 0

Residual life assessment of French PWR vessel head penetrations through metallurgical analysis

International Nuclear Information System (INIS)

Pichon, C.; Boudot, R.; Benhamou, C.; Gelpi, A.

1994-01-01

In September 1991, a vessel head penetration was found leaking at Bugey 3 plant during the hydrotest included in the framework of decennial In Service Inspections. Non destructive examinations performed afterwards on several other plants have shown some cracked penetrations. Destructive expertise confirmed quickly that again this new problem is related to stress corrosion cracking of Alloy 600 used as base material. During the last 15 years, similar cracking have been met in steam generator tubes and secondly in pressurizer instrumentation tubes. In spite of all the work performed since that time an extension appears to be necessary for explaining the features of this new event; however material sensitivity, stress and temperature still remain the key parameters governing the behavior of Alloy 600 in PWR environment. In this paper, only the material sensitivity of vessel head penetrations is examined through metallurgical analysis in relation with SCC tests. On the basis of vessel head field experience in combination with thermomechanical process used for fabrication of original bars criteria for a sensitivity ranking of penetrations are proposed. Metallurgical investigations and SCC tests were carried out to support this sensitivity ranking. The final aim is to use such information among those quoted above for assessment of vessel heads residual life. This document is an overview of the work performed in France concerning the material sensitivity of forged Alloy 600. It represents an important part of the assessments and investigations undertaken in France on the stress corrosion cracking phenomenon affecting the reactor vessel head penetrations in PWR's

Relaxation and corrosion resistance of alloy 800 used for steam generator tubes of ship borne boilers

International Nuclear Information System (INIS)

Corrieu, J.M.; Cortial, F.; Maillard, J.L.; Vernot-Loier, C.; Lebeau, M.

1994-01-01

The INCO ''INCOLOY 800'' trademark groups the Fe-Cr-Ni alloys containing 30 to 35% nickel, 19 to 23% chromium, 0,15 to 0,60% aluminium, 0,15 to 0,60% titanium and less than 0,10% carbon contents, used as construction materials for condenser and heat exchanger tubes. In parallel with water chemistry control and studies aimed at reducing the residual stresses resulting from tube expansion, studies have been conducted to a better understanding of this alloy, its metallurgy and its corrosion behaviour under accurately defined fabrication and heat treatment conditions. The purpose of this paper is to present the results of a behaviour study of INDRET alloy 800 concerning isothermal relaxation and effects of the said relaxation heat treatments on alloy microstructure studied with a transmission electron-chemical method to determine the sensitiveness to intergranular corrosion, and by electrochemistry in pressurized hot water. (authors). 4 figs., 5 tabs., 7 refs

Examination of steam generator alloy 800 NG tube from the Almaraz unit 2 NPP

International Nuclear Information System (INIS)

Diego, G. de; Gomez Briceno, D.; Maffiotte, C.; Baladia, M.; Arias, C.J.

2015-01-01

The steam generators of Almaraz Unit 2 were replaced in 1997 by the model 61W/D3 (Siemens) with Alloy 800NG steam generator tubes. Denting indications were firstly detected in 2006 in the SG-3. Crack indications were identified in 2009. At the end of 2011, three tubes were recovered from this steam generator to carry out destructive examination in order to identify the root cause of the tubes degradation. Analysis of deposits point out the existence of multiples elements in the removed OD (Outer Diameter) deposits as well as in the deposits at the free tube under sludge and at the transition zone. Deposits are more abundant at the transition zone than at free tube. About 10% Na concentration has been detected, whereas S and Cl appear in small concentrations. Si appears regularly and Cr, Ni concentrations in the deposits are similar. Multiple intergranular cracks have been detected at 3 mm above the last contact point between the tube and the TS (tube support), in a band of around 5 mm, practically in the whole perimeter of the tube. Fracture surface of crack-B was partially covered by a Si rich layer, whereas fracture surface of crack-A seems to be cleaner. However, no significant differences in composition, except higher amount of S in crack-B, were found in the deposits of both cracks. EDX mapping and Auger profiles point out Ni enrichment with slight Cr enrichment or depletion and Fe depletion. The comparison of Auger profiles with available results for Alloy 800 tested in caustic and acid sulfate environments seems to indicate that the environment inside the cracks detected in the tube R67C48 is neutral or moderately caustic

An evaluation of the statistical variability in thermal expansion properties of steam generator tubesheet (SA-508) and tubing (Alloy-600TT)

International Nuclear Information System (INIS)

Riccardella, P.C.; Staples, J.F.; Kandra, J.T.

2009-01-01

Inspections of steam generator tubing are performed in U.S. PWRs as part of the Steam Generator Management Program. Westinghouse has recently completed a technical justification demonstrating that in steam generators with thermally treated Ni-Cr Alloy (Alloy 600TT) tubes that are hydraulically expanded into low alloy steel (SA-508) tubesheets, flaws in the region of the tubes below a certain distance from the top of the tubesheet, denoted H * , will not result in reactor coolant pressure boundary breach nor unacceptable primary-to-secondary leakage. This is because, even if a flaw in this region were to result in complete tube sever, if the length of undegraded tube in the tubesheet exceeds H*, neither operating nor accident loadings create sufficient pull-out forces to overcome the frictional forces between the tube and tubesheet. One key component of this technical justification is the differential thermal expansion between the tube and tubesheet, since a significant portion of the pullout strength of the hydraulically expanded tube-to-tubesheet joint is due to mechanical interference resulting from the larger expansion of the tubing relative to the tubesheet at a given temperature. To address this phenomenon, a detailed statistical evaluation of coefficient of thermal expansion (CTE) data for the tubesheet material (SA-508) and the tube material (thermally treated Alloy-600) was performed. Data used in the evaluation included existing test results obtained from a number of sources as well as extensive new laboratory data developed specifically for this purpose. The evaluation resulted in recommended statistical distributions of this property for the two materials including their means and probabilistic variability. In addition, it was determined that the CTE values reported in the ASME Code (Section II) represent reasonably conservative mean values for both the tubesheet and tubing material. (author)

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

Energy Technology Data Exchange (ETDEWEB)

Chen, G.; Zhang, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Xu, D.K. [Environmental Corrosion Center, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, D.H. [Hunan Taohuajiang Nuclear Power Co., Ltd, Yiyang, 413000 (China); Chen, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Zhang, Z., E-mail: zhe.zhang@tju.edu.cn [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China)

2017-06-15

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ{sub x} did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ{sub xa}. For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ{sub xa} and the internal pressure p{sub i}. The hoop ratcheting strain ɛ{sub θ} increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ{sub x} was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

International Nuclear Information System (INIS)

Chen, G.; Zhang, X.; Xu, D.K.; Li, D.H.; Chen, X.; Zhang, Z.

2017-01-01

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ x did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ xa . For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ xa and the internal pressure p i . The hoop ratcheting strain ɛ θ increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ x was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Fabrication of Aluminum Tubes Filled with Aluminum Alloy Foam by Friction Welding

Directory of Open Access Journals (Sweden)

Yoshihiko Hangai

2015-10-01

Full Text Available Aluminum foam is usually used as the core of composite materials by combining it with dense materials, such as in Al foam core sandwich panels and Al-foam-filled tubes, owing to its low tensile and bending strengths. In this study, all-Al foam-filled tubes consisting of ADC12 Al-Si-Cu die-cast aluminum alloy foam and a dense A1050 commercially pure Al tube with metal bonding were fabricated by friction welding. First, it was found that the ADC12 precursor was firmly bonded throughout the inner wall of the A1050 tube without a gap between the precursor and the tube by friction welding. No deformation of the tube or foaming of the precursor was observed during the friction welding. Next, it was shown that by heat treatment of an ADC12-precursor-bonded A1050 tube, gases generated by the decomposition of the blowing agent expand the softened ADC12 to produce the ADC12 foam interior of the dense A1050 tube. A holding time during the foaming process of approximately tH = 8.5 min with a holding temperature of 948 K was found to be suitable for obtaining a sound ADC12-foam-filled A1050 tube with sufficient foaming, almost uniform pore structures over the entire specimen, and no deformation or reduction in the thickness of the tube.

Hydride-induced degradation of hoop ductility in textured zirconium-alloy tubes: A theoretical analysis

International Nuclear Information System (INIS)

Qin, W.; Szpunar, J.A.; Kozinski, J.

2012-01-01

Hydride-induced degradation of hoop ductility in Zr-alloy tubular components has been studied for many years because of its importance in the nuclear industry. In this paper the role of intergranular and intragranular δ-hydrides in the degradation of ductility of the textured Zr-alloy tubes is investigated. The correlation among hydride distribution, orientation and morphology in the tubes is formulated based on thermodynamic modeling, and then analyzed. The results show that the applied stress, the crystallographic texture of α-Zr matrix, the grain-boundary structure, and the morphology and size of Zr grains simultaneously govern the site preference and the orientation of hydrides. A criterion is proposed to determine the threshold stress of hydride reorientation. The hoop ductility of the hydrided Zr tubes is discussed using the concept of macroscopic fracture strain. It is shown that the intergranular hydrides may be more deleterious to ductility than the intragranular ones. This work defines a general framework for understanding the relation of the microstructure of hydride-forming materials to embrittlement.

Investigation of impingement attack mechanism of copper alloy condenser tubes

Energy Technology Data Exchange (ETDEWEB)

Fukumura, Takuya; Nakajima, Nobuo; Arioka, Koji; Totsuka, Nobuo; Nakagawa, Tomokazu [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

In order to investigate generation and growth mechanisms of impingement attacks of sea water against copper alloy condenser tubes used in condensers of nuclear power plants, we took out condenser tubes from actual condensers, cut them into several pieces and carried out several material tests mainly for impinged spots. In addition water flow inside of a pit was analyzed. From the results of the investigation, it was found that all of impingement attacks were found in the marks left by sessile organisms and none were found in downstream of the marks as frequently proposed so far. At the pits generated inside the marks, iron coating was striped and zinc content was deficient in some cases. Combining these data and the result of flow analysis, we considered the following mechanism of the impingement attacks: sessile organisms clinging to the surface of the condenser tube and growth, occlusion of the tube, extinction and decomposition of sessile organisms, pollution corrosion under the organisms and cavity formation, occlusion removal by the cleaning, generation of impingement attacks by flow collision inside the cavity, growth of the impingement attacks. (author)

Brazing open cell reticulated copper foam to stainless steel tubing with vacuum furnace brazed gold/indium alloy plating

Science.gov (United States)

Howard, Stanley R [Windsor, SC; Korinko, Paul S [Aiken, SC

2008-05-27

A method of fabricating a heat exchanger includes brush electroplating plated layers for a brazing alloy onto a stainless steel tube in thin layers, over a nickel strike having a 1.3 .mu.m thickness. The resultant Au-18 In composition may be applied as a first layer of indium, 1.47 .mu.m thick, and a second layer of gold, 2.54 .mu.m thick. The order of plating helps control brazing erosion. Excessive amounts of brazing material are avoided by controlling the electroplating process. The reticulated copper foam rings are interference fit to the stainless steel tube, and in contact with the plated layers. The copper foam rings, the plated layers for brazing alloy, and the stainless steel tube are heated and cooled in a vacuum furnace at controlled rates, forming a bond of the copper foam rings to the stainless steel tube that improves heat transfer between the tube and the copper foam.

Thermal creep properties of alloy D9 stainless steel and 316 stainless steel fuel clad tubes

International Nuclear Information System (INIS)

Latha, S.; Mathew, M.D.; Parameswaran, P.; Bhanu Sankara Rao, K.; Mannan, S.L.

2008-01-01

Uniaxial thermal creep rupture properties of 20% cold worked alloy D9 stainless steel (alloy D9 SS) fuel clad tubes for fast breeder reactors have been evaluated at 973 K in the stress range 125-250 MPa. The rupture lives were in the range 90-8100 h. The results are compared with the properties of 20% cold worked type 316 stainless steel (316 SS) clad tubes. Alloy D9 SS were found to have higher creep rupture strengths, lower creep rates and lower rupture ductility than 316 SS. The deformation and damage processes were related through Monkman Grant relationship and modified Monkman Grant relationship. The creep damage tolerance parameter indicates that creep fracture takes place by intergranular cavitation. Precipitation of titanium carbides in the matrix and chromium carbides on the grain boundaries, dislocation substructure and twins were observed in transmission electron microscopic investigations of alloy D9 SS. The improvement in strength is attributed to the precipitation of fine titanium carbides in the matrix which prevents the recovery and recrystallisation of the cold worked microstructure

Computational study of a low head draft tube and validation with experimental data

Energy Technology Data Exchange (ETDEWEB)

Henau, V De; Payette, F A; Sabourin, M [Alstom Power Systems, Hydro 1350 chemin Saint-Roch, Sorel-Tracy (Quebec), J3R 5P9 (Canada); Deschenes, C; Gagnon, J M; Gouin, P, E-mail: vincent.dehenau@power.alstom.co [Hydraulic Machinery Laboratory, Laval University 1065 ave. de la Medecine, Quebec (Canada)

2010-08-15

The objective of this paper is to investigate methodologies to improve the reliability of CFD analysis of low head turbine draft tubes. When only the draft tube performance is investigated, the study indicates that draft tube only simulations with an adequate treatment of the inlet boundary conditions for velocity and turbulence are a good alternative to rotor/stator (stage) simulations. The definition of the inlet velocity in the near wall regions is critical to get an agreement between the stage and draft tube only solutions. An average turbulent kinetic energy intensity level and average turbulent kinetic energy dissipation length scale are sufficient as turbulence inlet conditions as long as these averages are coherent with the stage solution. Comparisons of the rotor/stator simulation results to the experimental data highlight some discrepancies between the predicted draft tube flow and the experimental observations.

Study on Hydroforming of Magnesium Alloy Tube under Temperature Condition

Science.gov (United States)

Wang, Xinsong; Wang, Shouren; Zhang, Yongliang; Wang, Gaoqi; Guo, Peiquan; Qiao, Yang

2018-01-01

First of all, under 100 °C, 150 °C, 200 °C, 250 °C, 300 °C and 350 °C, respectively do the test of magnesium alloy AZ31B temperature tensile and the fracture of SEM electron microscopic scanning, studying the plastic forming ability under six different temperature. Secondly, observe and study the real stress-strain curves and fracture topography. Through observation and research can concluded that with the increase of temperature, the yield strength and tensile strength of AZ31B was increased, and the elongation rate and the plastic deformation capacity are increased obviously. Taking into account the actual production, energy consumption, and mold temperature resistance, 250 °Cwas the best molding temperature. Finally, under the temperature condition of 250 °C, the finite element simulation and simulation of magnesium alloy profiled tube were carried out by Dynaform, and the special wall and forming limit diagram of magnesium alloy were obtained. According to the forming wall thickness and forming limit diagram, the molding experiment can be optimized continuously.

Evaluation of High Temperature Corrosion Resistance of Finned Tubes Made of Austenitic Steel And Nickel Alloys

Directory of Open Access Journals (Sweden)

Turowska A.

2016-06-01

Full Text Available The purpose of the paper was to evaluate the resistance to high temperature corrosion of laser welded joints of finned tubes made of austenitic steel (304,304H and nickel alloys (Inconel 600, Inconel 625. The scope of the paper covered the performance of corrosion resistance tests in the atmosphere of simulated exhaust gases of the following chemical composition: 0.2% HCl, 0.08% SO2, 9.0% O2 and N2 in the temperature of 800°C for 1000 hours. One found out that both tubes made of austenitic steel and those made of nickel alloy displayed good resistance to corrosion and could be applied in the energy industry.

Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

Energy Technology Data Exchange (ETDEWEB)

Bahn, Chi Bum, E-mail: bahn@anl.gov [Argonne National Laboratory, Argonne, IL 60439 (United States); Majumdar, Saurin [Argonne National Laboratory, Argonne, IL 60439 (United States); Harris, Charles [United States Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

2011-10-15

Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low {Delta}K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed. - Highlights: > Leak rate of flawed Alloy 600 tubing increased at constant pressure condition. > Fractography revealed two cases: ductile tearing and crystallographic facets. > Crystallographic facets are typical features of fatigue crack growth at low {Delta}K. > Fatigue source could be water jet-induced vibration and/or high-pressure pump pulsation.

Proving the viability of manufacturing of multi-layer steel/vanadium alloy/steel composite tubes by numerical simulations and experiment

Science.gov (United States)

Nechaykina, T.; Nikulin, S.; Rozhnov, A.; Molotnikov, A.; Zavodchikov, S.; Estrin, Y.

2018-05-01

Vanadium alloys are promising structural materials for fuel cladding tubes for fast-neutron reactors. However, high solubility of oxygen and nitrogen in vanadium alloys at operating temperatures of 700 °C limits their application. In this work, we present a novel composite structure consisting of vanadium alloy V-4Ti-4Cr (provides high long-term strength of the material) and stainless steel Fe-0.2C-13Cr (as a corrosion resistant protective layer). It is produced by co-extrusion of these materials forming a three-layered tube. Finite element simulations were utilised to explore the influence of the various co-extrusion parameters on manufacturability of multi-layered tubes. Experimental verification of the numerical modelling was performed using co-extrusion with the process parameters suggested by the numerical simulations. Scanning electron microscopy and microhardness measurements revealed a defect-free diffusion layer at the interfaces between both materials indicating a good quality bonding for these co-extrusion conditions.

Comparative Evaluation of Cast Aluminum Alloys for Automotive Cylinder Heads: Part II—Mechanical and Thermal Properties

Science.gov (United States)

Roy, Shibayan; Allard, Lawrence F.; Rodriguez, Andres; Porter, Wallace D.; Shyam, Amit

2017-05-01

The first part of this study documented the as-aged microstructure of five cast aluminum alloys namely, 206, 319, 356, A356, and A356+0.5Cu, that are used for manufacturing automotive cylinder heads (Roy et al. in Metall Mater Trans A, 2016). In the present part, we report the mechanical response of these alloys after they have been subjected to various levels of thermal exposure. In addition, the thermophysical properties of these alloys are also reported over a wide temperature range. The hardness variation due to extended thermal exposure is related to the evolution of the nano-scale strengthening precipitates for different alloy systems (Al-Cu, Al-Si-Cu, and Al-Si). The effect of strengthening precipitates (size and number density) on the mechanical response is most obvious in the as-aged condition, which is quantitatively demonstrated by implementing a strength model. Significant coarsening of precipitates from long-term heat treatment removes the strengthening efficiency of the nano-scale precipitates for all these alloys systems. Thermal conductivity of the alloys evolve in an inverse manner with precipitate coarsening compared to the strength, and the implications of the same for the durability of cylinder heads are noted.

Crack growth of throughwall flaw in Alloy 600 tube during leak testing

International Nuclear Information System (INIS)

Bahn, Chi Bum; Majumdar, Saurin

2015-01-01

Graphical abstract: - Highlights: • A series of leak testing was conducted at a constant pressure and room temperature. • The time-dependent increase in the leak rate was observed. • The fractography revealed slip offsets and crystallographic facets. • Time-dependent plasticity at the crack tip caused the slip offsets. • Fatigue by jet/structure interaction caused the crystallographic facets. - Abstract: We examined the issue of whether crack growth in a full thickness material can occur in a leaking crack. A series of leak tests was conducted at a room temperature and constant pressure (17.3 MPa) with Alloy 600 tube specimens containing a tight rectangular throughwall axial fatigue crack. To exclude a potential pulsation effect by a high pressure pump, the test water was pressurized by using high pressure nitrogen gas. Fractography showed that crack growth in the full thickness material can occur in the leaking crack by two mechanisms: time-dependent plasticity at the crack tip and fatigue induced by jet/structure interaction. The threshold leak rate at which the jet/structure interaction was triggered was between 1.3 and 3.3 L/min for the specific heat of the Alloy 600 tube tested

Trunnion Failure of the Recalled Low Friction Ion Treatment Cobalt Chromium Alloy Femoral Head.

Science.gov (United States)

Urish, Kenneth L; Hamlin, Brian R; Plakseychuk, Anton Y; Levison, Timothy J; Higgs, Genymphas B; Kurtz, Steven M; DiGioia, Anthony M

2017-09-01

Gross trunnion failure (GTF) is a rare complication in total hip arthroplasty (THA) reported across a range of manufacturers. Specific lots of the Stryker low friction ion treatment (LFIT) anatomic cobalt chromium alloy (CoCr) V40 femoral head were recalled in August 2016. In part, the recall was based out of concerns for disassociation of the femoral head from the stem and GTF. We report on 28 patients (30 implants) with either GTF (n = 18) or head-neck taper corrosion (n = 12) of the LFIT CoCr femoral head and the Accolade titanium-molybdenum-zirconium-iron alloy femoral stems. All these cases were associated with adverse local tissue reactions requiring revision of the THA. In our series, a conservative estimate of the incidence of failure was 4.7% (n = 636 total implanted) at 8.0 ± 1.4 years from the index procedure. Failures were associated with a high-offset 127° femoral stem neck angle and increased neck lengths; 43.3% (13 of 30) of the observed failures included implant sizes outside the voluntary recall (27.8% [5 of 18] of the GTF and 75.0% [8 of 12] of the taper corrosion cases). Serum cobalt and chromium levels were elevated (cobalt: 8.4 ± 7.0 μg/mL; chromium: 3.4 ± 3.3 μ/L; cobalt/chromium ratio: 3.7). The metal artifact reduction sequence magnetic resonance imaging demonstrated large cystic fluid collections typical with adverse local tissue reactions. During revision, a pseudotumor was observed in all cases. Pathology suggested a chronic inflammatory response. Impending GTF could be diagnosed based on aspiration of black synovial fluid and an oblique femoral head as compared with the neck taper on radiographs. In our series of the recalled LFIT CoCr femoral head, the risk of impending GTF or head-neck taper corrosion should be considered as a potential diagnosis in a painful LFIT femoral head and Accolade titanium-molybdenum-zirconium-iron alloy THA with unknown etiology. Almost half of the failures we observed included sizes outside of the

Fuel assembly and fuel cladding tube

International Nuclear Information System (INIS)

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

1996-01-01

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

The Study of Heat Treatment Effects on Chromium Carbide Precipitation of 35Cr-45Ni-Nb Alloy for Repairing Furnace Tubes

Directory of Open Access Journals (Sweden)

Nakarin Srisuwan

2016-01-01

Full Text Available This paper presents a specific kind of failure in ethylene pyrolysis furnace tubes. It considers the case in which the tubes made of 35Cr-45Ni-Nb high temperature alloy failed to carburization, causing creep damage. The investigation found that used tubes became difficult to weld repair due to internal carburized layers of the tube. The microstructure and geochemical component of crystallized carbide at grain boundary of tube specimens were characterized by X-ray diffractometer (XRD, scanning electron microscopy (SEM with back-scattered electrons mode (BSE, and energy dispersive X-ray spectroscopy (EDS. Micro-hardness tests was performed to determine the hardness of the matrix and the compounds of new and used tube material. The testing result indicated that used tubes exhibited a higher hardness and higher degree of carburization compared to those of new tubes. The microstructure of used tubes also revealed coarse chromium carbide precipitation and a continuous carbide lattice at austenite grain boundaries. However, thermal heat treatment applied for developing tube weld repair could result in dissolving or breaking up chromium carbide with a decrease in hardness value. This procedure is recommended to improve the weldability of the 35Cr-45Ni-Nb used tubes alloy.

The use of radiologically placed gastroctomy tubes in head and neck cancer patients receiving radiotherapy

International Nuclear Information System (INIS)

Tyldesley, Scott; Sheehan, Finbarr; Munk, Peter; Tsang, Victor; Skarsgard, David; Bowman, Carol A.; Hobenshield, Shirley E.

1996-01-01

Purpose: Patients undergoing radiotherapy to the head and neck area frequently experience radiation reactions that can markedly restrict oral intake, require hospitalization, and occasionally cause treatment interruptions. The Vancouver Cancer Center (VCC) has recently employed radiologically placed gastrostomy tubes (G-tubes) in the management of this problem. A review of the patients on whom this procedure had been performed is the subject of this review. Methods and Materials: Thirty-four patients had gastrostomy tubes inserted under radiologic guidance. This group is compared to a control group matched for age, sex, irradiated volume, and radiation dose, who did not have gastrostomy tubes. Patients with gastrostomy tubes were divided into two categories: (a) patients who had tubes inserted in anticipation of severe reactions, and (b) patients who developed severe radiation reactions necessitating nutritional support. Results: The gastrostomy group consisted of 65% males with an average age of 59 years and stage range of II (12%), III (24%), and IV (65%). In both the elective group and the nonelective group, patients maintained their weight at 95 to 97% of the pretreatment weight, at follow-up of 6 weeks and 3 months. This compared with an average weight loss in the control group of 9% at 6 weeks and 12% at 3 months. The length of hospitalization was a mean of 4.9 days in the elective group and 19 days in the nonelective group. Complication were low compared to those documented in the literature, but included two tube migrations, two aspirations, and one gastrointestinal bleed. Conclusions: We believe that gastrostomy tubes contribute significantly to the management of patients with head and neck cancer, particularly in maintenance of nutrition, and they may decrease the need for hospitalization

The elastic properties of zirconium alloy fuel cladding and pressure tubing materials

International Nuclear Information System (INIS)

Rosinger, H.E.; Northwood, D.O.

1979-01-01

A knowledge of the elastic properties of zirconium alloys is required in the mathematical modelling of cladding and pressure tubing performance. Until recently, little of this type of data was available, particularly at elevated temperatures. The dynamic elastic moduli of zircaloy-2, zircaloy-4, the alloys Zr-1.0 wt%Nb, Zr-2.5 wt%Nb and Marz grade zirconium have therefore been determined over the temperature range 275 to 1000 K. Young's modulus and shear modulus for all the zirconium alloys decrease with temperature and are expressed by empirical relations fitted to the data. The elastic properties are texture dependent and a detailed study has been conducted on the effect of texture on the elastic properties of Zr-1.0 wt% Nb over the temperature range 275 to 775 K. The results are compared with polycrystalline elastic constants computed from single crystal elastic constants, and the effect of texture on the dynamic elastic moduli is discussed in detail. (Auth.)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Ion-induced swelling of ODS ferritic alloy MA957 tubing to 500 dpa

Energy Technology Data Exchange (ETDEWEB)

Toloczko, M.B., E-mail: mychailo.toloczko@pnnl.gov [Pacific Northwest National Laboratory, Richland, WA 99354 (United States); Garner, F.A. [Radiation Effects Consulting, Richland, WA 99354 (United States); Voyevodin, V.N.; Bryk, V.V.; Borodin, O.V.; Mel’nychenko, V.V.; Kalchenko, A.S. [Kharkov Institute of Physics and Technology, Kharkov (Ukraine)

2014-10-15

In order to study the potential swelling behavior of the ODS ferritic alloy MA957 at very high dpa levels, specimens were prepared from pressurized tubes that were unirradiated archives of tubes previously irradiated in FFTF to doses as high as 110 dpa. These unirradiated specimens were irradiated with 1.8 MeV Cr{sup +} ions to doses ranging from 100 to 500 dpa and examined by transmission electron microscopy. No co-injection of helium or hydrogen was employed. It was shown that compared to several tempered ferritic/martensitic steels irradiated in the same facility, these tubes were rather resistant to void swelling, reaching a maximum value of only 4.5% at 500 dpa and 450 °C. In this fine-grained material, the distribution of swelling was strongly influenced by the presence of void denuded zones along the grain boundaries.

Superelastic NiTi memory alloy micro-tube under tension - nucleation and propagation of martensite band

International Nuclear Information System (INIS)

Li, Z.Q.; Sun, Q.P.

2000-01-01

The superelastic behavior of polycrystalline NiTi shape memory alloy micro-tube under tension is studied experimentally. The nominal stress-strain curve of the micro-tube is recorded. By using a special surface coating it is found that the deformation of the tube is via the nucleation and propagation of stress-induced martensite band. The experiments show that the martensite nucleates in the form of a spiral lens-shaped narrow band that is inclined at 61 to the axis of loading when the stress reaches the peak of stress-strain curve. The width and the length of the band grew gradually with increase of loading and finally joined and merged into a single band. The subsequent deformation of the tube is realized by the propagation of this cylindrical martensite band. (orig.)

Microstructure control of Zr-Nb-Sn alloy with Mo addition for HWR pressure tube application

International Nuclear Information System (INIS)

Hwang, S. K.; Kim, M. H.; Kim, J. H.; Kwon, S. I.; Kim, Y. S.

1997-01-01

As a basic research to develop the material for heavy water reactor pressure tube application the effect of Mo addition to Zr-Nb-Sn alloy was studied for the purpose of minimizing the amount of cold working while maintaining a high strength. To select the target alloy system we first designed various alloy compositions and chose Zr-Nb-Sn and Zr-Nb-Mo through multi-regression analysis of the relationship between the basic properties and the compositions. Plasma arc melting was used to produce the alloys and the microstructure change introduced by the processing steps including hot forging, beta-heat treatment, hot rolling, cold rolling and recrystallization heat treatment was investigated. Recrystallization of Zr-Nb-Sn was retarded by adding Mo and this resulted in a fine grain structure in Zr-Nb-Sn-Mo alloy. Beside the retarding effect recrystallization, Mo increased the amount of residual beta phase and showed an indication of precipitation hardening, which added up to the possibility of applying the alloy for the desired usage. (author)

Comparative Evaluation of Cast Aluminum Alloys for Automotive Cylinder Heads: Part I—Microstructure Evolution

Science.gov (United States)

Roy, Shibayan; Allard, Lawrence F.; Rodriguez, Andres; Watkins, Thomas R.; Shyam, Amit

2017-05-01

The present study stages a comparative evaluation of microstructure and associated mechanical and thermal response for common cast aluminum alloys that are used for manufacturing automotive cylinder heads. The systems considered are Al-Cu (206-T6), Al-Si-Cu (319-T7), and Al-Si (356-T6, A356-T6, and A356 + 0.5Cu-T6). The focus of the present manuscript is on the evaluation of microstructure at various length scales after aging, while the second manuscript will deal with the mechanical and thermal response of these alloys due to short-term (aging) and long-term (pre-conditioning) heat treatments. At the grain-scale, the Al-Cu alloy possessed an equiaxed microstructure as opposed to the dendritic structure for the Al-Si-Cu or Al-Si alloys which is related to the individual solidification conditions for these alloy systems. The composition and morphology of intermetallic precipitates within the grain and at the grain/dendritic boundary are dictated by the alloy chemistry, solidification, and heat treatment conditions. At the nanoscale, these alloys contain various metastable strengthening precipitates (GPI and θ^'' in Al-Cu alloy, θ^' in Al-Si-Cu alloy, and β^' in Al-Si alloys) with varying size, morphology, coherency, and thermal stability.

Viability of thin wall tube forming of ATF FeCrAl

Energy Technology Data Exchange (ETDEWEB)

Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Anderoglu, Osman [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lavender, Curt [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Yamamoto, Yukinori [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-09-16

Fabrication of thin walled tubing of FeCrAl alloys is critical to its success as a candidate enhanced accident-tolerant fuel cladding material. Alloys that are being investigated are Generation I and Generation II FeCrAl alloys produced at ORNL and an ODS FeCrAl alloy, MA-956 produced by Special Metals. Gen I and Gen II FeCrAl alloys were provided by ORNL and MA-956 was provided by LANL (initially produced by Special Metals). Three tube development efforts were undertaken. ORNL led the FeCrAl Gen I and Gen II alloy development and tube processing studies through drawing tubes at Rhenium Corporation. LANL received alloys from ORNL and led tube processing studies through drawing tubes at Century Tubing. PNNL led the development of tube processing studies on MA-956 through pilger processing working with Sandvik Corporation. A summary of the recent progress on tube development is provided in the following report and a separate ORNL report: ORNL/TM-2015/478, “Development and Quality Assessments of Commercial Heat Production of ATF FeCrAl Tubes”.

Design of a Kaplan turbine for a wide range of operating head -Curved draft tube design and model test verification-

Science.gov (United States)

KO, Pohan; MATSUMOTO, Kiyoshi; OHTAKE, Norio; DING, Hua

2016-11-01

As for turbomachine off-design performance improvement is challenging but critical for maximising the performing area. In this paper, a curved draft tube for a medium head Kaplan type hydro turbine is introduced and discussed for its significant effect on expanding operating head range. Without adding any extra structure and working fluid for swirl destruction and damping, a carefully designed outline shape of draft tube with the selected placement of center-piers successfully supresses the growth of turbulence eddy and the transport of the swirl to the outlet. Also, more kinetic energy is recovered and the head lost is improved. Finally, the model test results are also presented. The obvious performance improvement was found in the lower net head area, where the maximum efficiency improvement was measured up to 20% without compromising the best efficiency point. Additionally, this design results in a new draft tube more compact in size and so leads to better construction and manufacturing cost performance for prototype. The draft tube geometry parameter designing process was concerning the best efficiency point together with the off-design points covering various water net heads and discharges. The hydraulic performance and flow behavior was numerically previewed and visualized by solving Reynolds-Averaged Navier-Stokes equations with Shear Stress Transport turbulence model. The simulation was under the assumption of steady-state incompressible turbulence flow inside the flow passage, and the inlet boundary condition was the carefully simulated flow pattern from the runner outlet. For confirmation, the corresponding turbine efficiency performance of the entire operating area was verified by model test.

SeZnSb alloy and its nano tubes, graphene composites properties

Directory of Open Access Journals (Sweden)

Abhay Kumar Singh

2013-04-01

Full Text Available Composite can alter the individual element physical property, could be useful to define the specific use of the material. Therefore, work demonstrates the synthesis of a new composition Se96-Zn2-Sb2 and its composites with 0.05% multi-walled carbon nano tubes and 0.05% bilayer graphene, in the glassy form. The diffused amorphous structure of the multi walled carbon nano tubes and bilayer gaphene in the Se96-Zn2-Sb2 alloy have been analyzed by using the Raman, X-ray photoluminescence spectroscopy, Furrier transmission infrared spectra, photoluminescence, UV/visible absorption spectroscopic measurements. The diffused prime Raman bands (G and D have been appeared for the multi walled carbon nano tubes and graphene composites, while the X-ray photoluminescence core energy levels peak shifts have been observed for the composite materials. Subsequently the photoluminescence property at room temperature and a drastic enhancement (upto 80% in infrared transmission percentage has been obtained for the bilayer graphene composite, along with optical energy band gaps for these materials have been evaluated 1.37, 1.39 and 1.41 eV.

Demonstration of a shape memory alloy torque tube-based morphing radiator

Science.gov (United States)

Chong, Jorge B.; Walgren, Patrick; Hartl, Darren J.

2018-03-01

Long-distance crewed space exploration will require advanced thermal control systems (TCS) with the ability to handle a wide range of thermal loads. The ability of a TCS to adapt to the thermal environment is described by the turndown ratio. Developing radiators with high turndown ratios is critical for improving TCS technology. This paper describes a novel morphing radiator designed to achieve a high turndown ratio by varying its own radiative view factor and effective emissivity through the use of shape memory alloys (SMAs). This radiator features two SMA torque tubes cantilevered to a rigid fixture. The working fluid is transported within the SMA tubes through an annular flow system. In a cold environment, radiator panels fixed to the free ends of the tubes are oriented vertically in a parallel-plate fashion, where the high-emissivity interior faces have restricted views to the environment and heat rejection is minimized. When the system heats up, the tubes actuate by twisting in opposing directions, bringing the panels to a horizontal position with the interior faces exposed to maximize heat rejection. When the system cools down, the tubes twist in reverse, restoring the panels to the vertical orientation where heat rejection is again minimized. This variable heat rejection system has the potential for achieving higher turndown ratios than those of current state-of-the-art systems. A benchtop prototype has been designed and tested to demonstrate actuation and to explore internal heat transfer effects. Prototype design, testing, and results are herein described.

Environmentally assisted fatigue evaluation model of alloy 690 steam generator tube in high temperature water

International Nuclear Information System (INIS)

Tan Jibo; Wu Xinqiang; Han Enhou; Wang Xiang; Liu Xiaoqiang; Xu Xuelian

2015-01-01

Nickel-based alloy 690 has been widely used as steam generator tube in light water reactor (LWR) nuclear power plants, which may suffer from corrosion fatigue during long-term service. Many researches and operating experience indicated that the effect of LWR environment could significantly reduce the fatigue life of structural materials. However. such an environmental degradation effect was not fully addressed in the current ASME code design fatigue curves. Therefore, the Regulatory Guide 1.207 issued by US NRC required a new NPP have to incorporate the environment effects into fatigue analyses. In the last few decades, researchers in USA and Japan systematically investigated the corrosion fatigue behavior of nuclear-grade structural materials in LWR environment. Then, ANL model and JSME model were proposed, which incorporated environmental effects, including temperature, dissolved oxygen (DO) and strain rate for the nickel-based alloys. Due to lack of experiment data on domestic materials, there is no related environmental fatigue design model in China. In the present work, based on the corrosion fatigue tests of a kind of boat-shaped specimen in borated and lithiated high temperature water, the corrosion fatigue behavior and environmentally assisted cracking mechanism of domestic Alloy 690 steam generator tube have been investigate. An IMR model for the nickel-based alloy was proposed. The environmental fatigue life correction factor (F en ) was established, which addressed the environmental factors, including temperature, strain rate and dissolved oxygen. The method to evaluate environmental fatigue damage of structural materials in NPPs was proposed. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

2010-07-01

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

Welding qualification procedure for fuel rods tubes of Zr-Sn alloys by the TIG automatic process

International Nuclear Information System (INIS)

1984-11-01

It is presented the requirements to be used in the Welding qualification procedure for tubes of Zr-Sn alloys, specified in the ASTM B353 regulatory guide, used in the fabrication of fuel rods PWR reactors by the automatic TIG process. (E.G.) [pt

Comparative evaluation of cast aluminum alloys for automotive cylinder heads: Part I Microstructure evolution

International Nuclear Information System (INIS)

Roy, Shibayan; Allard, Lawrence Frederick Jr; Rodriguez, Andres; Watkins, Thomas R.; Shyam, Amit

2017-01-01

The present study stages a comparative evaluation of microstructure and associated mechanical and thermal response for common cast aluminum alloys that are used for manufacturing automotive cylinder heads. The systems considered are Al-Cu (206-T6), Al-Si-Cu (319-T7), and Al-Si (356-T6, A356-T6, and A356 + 0.5Cu-T6). The focus of the present manuscript is on the evaluation of microstructure at various length scales after aging, while the second manuscript will deal with the mechanical and thermal response of these alloys due to short-term (aging) and long-term (pre-conditioning) heat treatments. At the grain-scale, the Al-Cu alloy possessed an equiaxed microstructure as opposed to the dendritic structure for the Al-Si-Cu or Al-Si alloys which is related to the individual solidification conditions for these alloy systems. The composition and morphology of intermetallic precipitates within the grain and at the grain/dendritic boundary are dictated by the alloy chemistry, solidification, and heat treatment conditions. At the nanoscale, these alloys contain various metastable strengthening precipitates (GPI and θ''θ'' in Al-Cu alloy, θ'θ' in Al-Si-Cu alloy, and β'β' in Al-Si alloys) with varying size, morphology, coherency, and thermal stability.

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

CERN Document Server

American Society for Testing and Materials. Philadelphia

2011-01-01

Standard specification for Nickel-Chromium-Iron alloys (UNS N06600, N06601, N06603, N06690, N06693, N06025, N06045, and N06696), Nikel-Chromium-Cobalt-Molybdenum alloy (UNS N06617), and Nickel-Iron-Chromium-Tungsten alloy (UNS N06674) seamless pipe and tube

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

International Nuclear Information System (INIS)

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

1993-01-01

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

Advanced nondestructive examination of the reactor vessel head penetration tube welds

International Nuclear Information System (INIS)

Cvitanovic, M.; Zado, V.

1996-01-01

Beside a referent code examination requirements, appearance of the service induced flaws on the Reactor Vessel Head (RVH) penetration tube welds forced development of remotely operated examination tools and techniques. Several systems were developed for examination of RVH PWR type while only one system for examination of VVER - 440 type RVH has been developed by Inetec. In this article the most advanced RVH VVER - 440 type examination techniques such as ultrasonic, eddy current and visual testing techniques as well as remotely operated tool are described. (author)

SCC of Alloy 600 in PWR steam generator tubes

International Nuclear Information System (INIS)

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

1986-01-01

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

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

International Nuclear Information System (INIS)

Mathews, A.J.

1988-01-01

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

Bottom head assembly

International Nuclear Information System (INIS)

Fife, A.B.

1998-01-01

A bottom head dome assembly is described which includes, in one embodiment, a bottom head dome and a liner configured to be positioned proximate the bottom head dome. The bottom head dome has a plurality of openings extending there through. The liner also has a plurality of openings extending there through, and each liner opening aligns with a respective bottom head dome opening. A seal is formed, such as by welding, between the liner and the bottom head dome to resist entry of water between the liner and the bottom head dome at the edge of the liner. In the one embodiment, a plurality of stub tubes are secured to the liner. Each stub tube has a bore extending there through, and each stub tube bore is coaxially aligned with a respective liner opening. A seat portion is formed by each liner opening for receiving a portion of the respective stub tube. The assembly also includes a plurality of support shims positioned between the bottom head dome and the liner for supporting the liner. In one embodiment, each support shim includes a support stub having a bore there through, and each support stub bore aligns with a respective bottom head dome opening. 2 figs

X-ray study of texture in zirconium alloy tubes and in graphite

International Nuclear Information System (INIS)

Skvortsov, V.V.; Alekseev, S.I.

1987-01-01

X-ray study of texture in zirconium alloy tubes and in graphite has been developed. The method is based on constructing coordinate grid of stereographic projection determining quantity and coordinates of points where measurements should be performed depending on a specimen slope pitch. Complete stereographic projection obtained so is a base both for constructing pole figures showing distribution normales of plane system being studied and for calculating texture coefficients determining property anisotropy in materials under investigation. This method can be applied to study texture in items of any materials independent of the item shape

Microstructure, elastic deformation behavior and mechanical properties of biomedical β-type titanium alloy thin-tube used for stents.

Science.gov (United States)

Tian, Yuxing; Yu, Zhentao; Ong, Chun Yee Aaron; Kent, Damon; Wang, Gui

2015-05-01

Cold-deformability and mechanical compatibility of the biomedical β-type titanium alloy are the foremost considerations for their application in stents, because the lower ductility restricts the cold-forming of thin-tube and unsatisfactory mechanical performance causes a failed tissue repair. In this paper, β-type titanium alloy (Ti-25Nb-3Zr-3Mo-2Sn, wt%) thin-tube fabricated by routine cold rolling is reported for the first time, and its elastic behavior and mechanical properties are discussed for the various microstructures. The as cold-rolled tube exhibits nonlinear elastic behavior with large recoverable strain of 2.3%. After annealing and aging, a nonlinear elasticity, considered as the intermediate stage between "double yielding" and normal linear elasticity, is attributable to a moderate precipitation of α phase. Quantitive relationships are established between volume fraction of α phase (Vα) and elastic modulus, strength as well as maximal recoverable strain (εmax-R), where the εmax-R of above 2.0% corresponds to the Vα range of 3-10%. It is considered that the "mechanical" stabilization of the (α+β) microstructure is a possible elastic mechanism for explaining the nonlinear elastic behavior. Copyright © 2015 Elsevier Ltd. All rights reserved.

Cladding tube manufacturing technology

Energy Technology Data Exchange (ETDEWEB)

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

1999-04-01

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

Experience in quality assurance of alloy D9 clad tubes for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Kapoor, K.; Prahlad, B.

2012-01-01

Stainless Steel Alloy D9 is the material for cladding in various sub-assemblies of Prototype Fast Breeder Reactor (PFBR). The fabrication, inspection, testing and supply of the clad tubes for the first core of PFBR is nearly completed. The paper also compares the specification requirements and the achieved results for some of the critical aspects which is arrived after completing supply against the first core requirement

Steam generator tube integrity program

International Nuclear Information System (INIS)

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

1996-01-01

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

Development of the advanced nuclear materials -Development of Inconel alloys-

International Nuclear Information System (INIS)

Kuk, Il Hyun; Chang, Jin Sung; Lee, Chang Kyu; Park, Soon Dong; Kim, Woo Kon; Jeong, Man Kyo; Woo, Yoon Myung; Han, Chang Hee

1995-07-01

The performance and the integrity of the steam generator U-tubes directly affects the efficiency and economics of nuclear power plant because they are closely interrelated with the maintenance and repair. Also the steam generator U-tubes have been one of world-wide hot issues in nuclear power plants for long time because of their continuing corrosion-related degradation. Right after stress corrosion cracking of Alloy 600 tubes are reported at primary side, in which the environment is believed to be tightly controlled all the time, in mid 80's, alloy 690 has started to replace alloy 600. Alloy 690 is basically same with alloy 600 except more Cr content. Firstly minor elements in alloy 690 (C, B, N, Y, Mo) were added or controlled to improve hot workability and corrosion resistance. It would be much more desirable if the mechanism or basic understanding of the degradation phenomena of steam generator U-tubes in operation conditions can be illuminated through the alloy modification research. Alloy 600 tubes which were preproduced in cooperation with Sammi Special Steel were evaluated, being compared with imported one. Also alloy 600 and alloy 690 tubes were produced from Inconel 600 and 690 INCO- forged bar. These will be closely evaluated with purely Korean-made alloy 600 and 690 tubes. 22 tabs., 93 figs., 14 refs. (Author)

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.

Synthesis and characterization of Ni-Mo filler brazing alloy for Mo-W joining for microwave tube technology

Directory of Open Access Journals (Sweden)

Frank Ferrer Sene

2013-04-01

Full Text Available A brazing process based on Ni-Mo alloy was developed to join porous tungsten cathode bottom and dense molybdenum cathode body for microwave tubes manufacture. The Ni-Mo alloy was obtained by mixing and milling powders in the eutectic composition, and applied on the surface of the components. The brazing was made at 1400 °C by using induction heating in hydrogen for 5 minutes. Alumina surfaces were coated with the binder and analyzed by Energy Dispersive X-rays Fluorescence. The brazed samples were analyzed by Scanning Electron Microscopy coupled to Energy Dispersive Spectroscopy. Stress-strain tests were performed to determine the mechanical behavior of the joining. The quality of the brazing was evaluated by assuring the presence of a "meniscus" formed by the Ni-Mo alloy on the border of the tungsten and molybdenum joint, the absence of microstructural defects in the interface between the tungsten and molybdenum alloys, and the adhesion of the brazed components.

Oxide Dispersion Strengthened Fe(sub 3)Al-Based Alloy Tubes: Application Specific Development for the Power Generation Industry

Energy Technology Data Exchange (ETDEWEB)

Kad, B.K.

1999-07-01

A detailed and comprehensive research and development methodology is being prescribed to produce Oxide Dispersion Strengthened (ODS)-Fe3Al thin walled tubes, using powder extrusion methodologies, for eventual use at operating temperatures of up to 1100C in the power generation industry. A particular 'in service application' anomaly of Fe3Al-based alloys is that the environmental resistance is maintained up to 1200C, well beyond where such alloys retain sufficient mechanical strength. Grain boundary creep processes at such high temperatures are anticipated to be the dominant failure mechanism.

Study on frictional pressure drop of steam-water two phase flow in optimized four-head internal-ribbed tube

International Nuclear Information System (INIS)

Wang Weishu; Zhu Xiaojing; Bi Qincheng; Wu Gang; Yu Shuiqing

2012-01-01

The optimized internal-ribbed tube is different from the normal internal-ribbed tube on the frictional pressure drop characteristics. The frictional pressure drop characteristics of steam-water two phase flow in horizontal four-head optimized internal-ribbed were studied under adiabatic condition. According to the experimental and calculation results, the two-phase multiplier is greatly affected by the steam quality and pressure. The two-phase multiplier increases with increasing quality, and decreases with increasing pressure. In the near-critical pressure region, the two-phase multiplier is close to 1. The frictional pressure drop of two phase flow in optimized tube is less than that in the normal tube under the same work condition. The good hydrodynamic condition could be achieved when the optimized internal-ribbed tube is used in the heat transfer equipment because the self-compensating characteristics exist due to the reduction of frictional pressure drop. (authors)

Strong, corrosion-resistant aluminum tubing

Science.gov (United States)

Reed, M. W.; Adams, F. F.

1980-01-01

When aluminum tubing having good corrosion resistance and postweld strength is needed, type 5083 alloy should be considered. Chemical composition is carefully controlled and can be drawn into thin-wall tubing with excellent mechanical properties. Uses of tubing are in aircraft, boats, docks, and process equipment.

Heat exchanger tubing materials for CANDU nuclear generating stations

International Nuclear Information System (INIS)

Taylor, G.F.

1977-07-01

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Scalable shape- and size-controlled synthesis of metal nano-alloys

KAUST Repository

Bakr, Osman M.

2016-01-21

Embodiments of the present disclosure provide for a continuous-flow reactor, methods of making metal nano-alloys, and metal nano-alloys. An embodiment of the continuous-flow reactor includes a first tubular component having a tubular inlet and a tubular outlet, and a heated tube-in-tube gas reactor fluidly connected to the first tubular component, wherein the heated tube-in-tube gas reactor comprises an inner tube having a gas permeable surface and an outer tube. An embodiment of the method of producing metal nano-alloys, includes contacting a reducible metal precursor and a reducing fluid in a continuous-flow reactor to form a mixed solution; and flowing the mixed solution through the continuous-flow reactor for a residence time to form the metal nano-alloys. An embodiment of the composition includes a plurality of metal nano-alloys having a monodisperse size distribution and a uniform shape distribution.

Spring/dimple instrument tube restraint

International Nuclear Information System (INIS)

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

1993-01-01

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

Mechanical splicing of superelastic Cu–Al–Mn alloy bars with headed ends

Science.gov (United States)

Kise, S.; Mohebbi, A.; Saiidi, M. S.; Omori, T.; Kainuma, R.; Shrestha, K. C.; Araki, Y.

2018-06-01

This paper examines the feasibility of mechanical splicing using a steel coupler to connect headed ends of superelastic Cu–Al–Mn alloy (Camalloy) bars and steel reinforcing bars to be used in concrete structures. Although threading of Camalloy is as easy as that of steel, mechanical splicing using threaded ends requires machining of Camalloy bars into dog-bone shape to avoid brittle fracture at the threaded ends. The machining process requires significant time and cost and wastes substantial amount of the material. This paper attempts to resolve this issue by applying mechanical splicing using steel couplers to connect headed ends of Camalloy and steel reinforcing bars. To study its feasibility, we prepare 3 specimens wherein both ends of each Camalloy bar (13 mm diameter and 300 mm length) are connected to steel reinforcing bars. The specimens are tested under monotonic, single-cycle, and full-cycle tension loading conditions. From these tests, we observed (1) excellent superelasticity with recoverable strain of around 6% and (2) large ductility with fracture strain of over 19%. It should be emphasized here that, in all the specimens, ductile fracture occurred at the locations apart from the headed ends. This is in sharp contrast with brittle fracture of headed superelastic Ni–Ti SMA bars, most of which took place around the headed ends. From the results of the microstructural analysis, we identified the following reasons for avoiding brittle fracture at the headed ends: (1) Precipitation hardening increases the strength around the boundary between the straight and headed (tapered) portions, where stress concentration takes place. (2) The strength of the straight portion does not increase significantly up to the ductile fracture if its grain orientation is close to 〈0 0 1〉.

Evolution of grain boundary character distributions in alloy 825 tubes during high temperature annealing: Is grain boundary engineering achieved through recrystallization or grain growth?

International Nuclear Information System (INIS)

Bai, Qin; Zhao, Qing; Xia, Shuang; Wang, Baoshun; Zhou, Bangxin; Su, Cheng

2017-01-01

Grain boundary engineering (GBE) of nickel-based alloy 825 tubes was carried out with different cold drawing deformations by using a draw-bench on a factory production line and subsequent annealing at various temperatures. The microstructure evolution of alloy 825 during thermal-mechanical processing (TMP) was characterized by means of the electron backscatter diffraction (EBSD) technique to study the TMP effects on the grain boundary network and the evolution of grain boundary character distributions during high temperature annealing. The results showed that the proportion of ∑ 3 n coincidence site lattice (CSL) boundaries of alloy 825 tubes could be increased to > 75% by the TMP of 5% cold drawing and subsequent annealing at 1050 °C for 10 min. The microstructures of the partially recrystallized samples and the fully recrystallized samples suggested that the proportion of low ∑ CSL grain boundaries depended on the annealing time. The frequency of low ∑ CSL grain boundaries increases rapidly with increasing annealing time associating with the formation of large-size highly-twinned grains-cluster microstructure during recrystallization. However, upon further increasing annealing time, the frequency of low ∑ CSL grain boundaries decreased markedly during grain growth. So it is concluded that grain boundary engineering is achieved through recrystallization rather than grain growth. - Highlights: •The grain boundary engineering (GBE) is applicable to 825 tubes. •GBE is achieved through recrystallization rather than grain growth. •The low ∑ CSL grain boundaries in 825 tubes can be increased to > 75%.

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)

Corrosion in PWR steam generator tubes made of alloy 600TT: overview of operating experience, NDE and safety issues

International Nuclear Information System (INIS)

Curieres, I. de; Sollier, T.; Delaval, C.

2015-01-01

About 60 PWR plants worldwide are operating with steam generator tubes made of alloy 600TT, among which 27 are located in France. This alloy is susceptible to corrosion, both on the primary and secondary side in every fleet, though with different kinetics or extent. It is noteworthy that many of the primary side corrosion issues can be clearly explained by design or operating conditions. However, studies show that all the secondary side issues are much hardly explained by simple considerations. This paper will give an overview of the international operating experience of this alloy and indicate the associated controllability and safety-related issues. An emphasis will be put on the manufacturing, chemistry and specificities of the different fleets. The French situation will be reviewed in this frame. (authors)

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1996-08-01

In Phase 1 of this project, a variety of developmental and commercial tubing alloys and claddings was exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 modified, NF 709, 690 clad, and 671 clad for over 10,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy are being exposed for 4,000, 12,000, and 16,000 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after approximately 4,400 hours of exposure.

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.

Method of forming magnetostrictive rods from rare earth-iron alloys

Science.gov (United States)

McMasters, O. Dale

1986-09-02

Rods of magnetrostructive alloys of iron with rare earth elements are formed by flowing a body of rare earth-iron alloy in a crucible enclosed in a chamber maintained under an inert gas atmosphere, forcing such molten rare-earth-iron alloy into a hollow mold tube of refractory material positioned with its lower end portion within the molten body by means of a pressure differential between the chamber and mold tube and maintaining a portion of the molten alloy in the crucible extending to a level above the lower end of the mold tube so that solid particles of higher melting impurities present in the alloy collect at the surface of the molten body and remain within the crucible as the rod is formed in the mold tube.

Importance of crevices formed between tubes and tube plate for the operational behaviour of heat exchangers

International Nuclear Information System (INIS)

Achten, N.; Herbsleb, G.; Wieling, N.

1986-01-01

It must be guaranteed by construction and manufacture of heat exchangers that primary and secondary medium are completely separated from each other. When this requirement is fullfilled, the operational use of heat exchangers can be impaired by corrosion reactions within the crevice formed between tube and tube plate which may result in corrosion damage. The various techniques which are in use to connect tubes and tube plate and which are described in the present report, must be valued with respect to the tightness of the connection as well as to the formation of crevices between tubes and tube plate. Corrosion resistant copperbase alloys and stainless steels are the most important materials which are in use for the construction of heat exchangers. The mechanisms of crevice corrosion with unalloyed and low alloy carbon steels, stainless steels, and mixed connections between tube and tube plate with these materials are described in detail. Crevice corrosion may be caused also by the formation of galvanic cells between materials of differing electrochemical response. Furthermore, the concentration of aggressive media in crevices between tubes and tube plate can lead to corrosion damage of heat exchanger tubes. For the service operation of heat exchangers without any hazard of corrosion damage in crevices between tubes and tube plate, such crevices must be avoided by proper construction and manufacture. As a model for suitable measures to avoid crevices, the manufacture of steam generators for PWR's is described. (orig.) [de

Repair technology for steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

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

2001-02-01

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

Repair technology for steam generator tubes

International Nuclear Information System (INIS)

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

2001-02-01

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

Cladding tube manufacturing technology

International Nuclear Information System (INIS)

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

1999-04-01

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

Leakproof Swaged Joints in Thin-Wall Tubing

Science.gov (United States)

Stuckenberg, F. H.; Crockett, L. K.; Snyder, W. E.

1986-01-01

Tubular inserts reinforce joints, reducing incidence of leaks. In new swaging technique, tubular inserts placed inside ends of both tubes to be joined. Made from thicker-wall tubing with outside diameter that matches inside diameter of thin tubing swaged, inserts support tube ends at joint. They ensure more uniform contact between swage fitting and tubing. New swaging technique developed for Al/Ti/V-alloy hydraulic supply lines.

Oxide Dispersion Strengthened Fe3Al-Based Alloy Tubes: Application Specific Development for the Power Generation Industry

Energy Technology Data Exchange (ETDEWEB)

Kad, B.K.

2002-02-08

A detailed and comprehensive research and development methodology is being prescribed to produce Oxide Dispersion Strengthened (ODS)-Fe{sub 3}Al thin walled tubes, using powder extrusion methodologies, for eventual use at operating temperatures of up to 1100% in the power generation industry. A particular ''in service application'' anomaly of Fe{sub 3}Al-based alloys is that the environmental resistance is maintained up to 1200 C, well beyond where such alloys retain sufficient mechanical strength. Grain boundary creep processes at such high temperatures are anticipated to be the dominant failure mechanism. Thus, the challenges of this program are manifold: (1) to produce thin walled ODS-Fe{sub 3}Al tubes, employing powder extrusion methodologies, with (2) adequate increased strength for service at operating temperatures, and (3) to mitigate creep failures by enhancing the as-processed grain size in ODS-Fe{sub 3}Al tubes. Our research progress till date has resulted in the successful batch production of typically 8 Ft. lengths of 1-3/8 inch diameter, 1/8 inch wall thickness, ODS-Fe{sub 3}Al tubes via a proprietary single step extrusion consolidation process. The process parameters for such consolidation methodologies have been prescribed and evaluated as being routinely reproducible. Such processing parameters (i.e., extrusion ratios, temperature, can design etc.) were particularly guided by the need to effect post-extrusion recrystallization and grain growth at a sufficiently low temperature, while still meeting the creep requirement at service temperatures. Static recrystallization studies show that elongated grains (with their long axis parallel to the extrusion axis), typically 200-2000 {micro}m in diameter, and several millimeters long can be obtained routinely, at 1200 C. The growth kinetics are affected by the interstitial impurity content in the powder batches. For example complete recrystallization, across the tube wall thickness, is

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Randomized study of percutaneous endoscopic gastrostomy versus nasogastric tubes for enteral feeding in head and neck cancer patients treated with (chemo)radiation

International Nuclear Information System (INIS)

Corry, J.; Poon, W.; McPhee, N.; Milner, A. D.; Cruickshank, D.; Rischin, D.; Peters, L. J.

2008-01-01

Full text: Percutaneous endoscopic gastrostomy (PEG) tubes have largely replaced nasogastric tubes (NGT) for nutritional support of patients with head and neck cancer undergoing curative (chemo)radiotherapy without any good scientific basis. A randomized trial was conducted to compare PEG tubes and NGT in terms of nutritional outcomes, complications, patient satisfaction and cost. The study was closed early because of poor accrual, predominantly due to patients' reluctance to be randomized. There were 33 patients eligible for analysis. Nutritional support with both tubes was good. There were no significant differences in overall complication rates, chest infection rates or in patients' assessment of their overall quality of life. The cost of a PEG tube was 10 times that of an NGT. The duration of use of PEG tubes was significantly longer, a median 139 days compared with a median 66 days for NGT. We found no evidence to support the routine use of PEG tubes over NGT in this patient group

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

Deposition of magnetite particles onto alloy-800 steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

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

1998-07-01

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

Deposition of magnetite particles onto alloy-800 steam generator tubes

International Nuclear Information System (INIS)

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

1998-01-01

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

Influences of Alloying Element and Annealing on the Microstructure and Corrosion Resistance of Steam Generator Tubing Materials of Nuclear Power Plant (I)

International Nuclear Information System (INIS)

Kim, Young Sik; Pari, Yong Soo; Kuk, Il Hiun

1996-01-01

Influences of alloying elements and annealing heat treatments on Alloy 690 and Alloy 600 for steam generator tubing materials of nuclear power plants were studied. OM, SEM, TEM, and XRD analyses were used to study the microstructural changes of the alloys. Mechanical properties were investigated by means of tension tests and Rockwell hardness tests, and corrosion resistance was evaluated using the anodic polarization tests and the 65% boiling nitric acid immersion tests. Increasing the carbon content of Alloy 690, the hardness and tensile strength were increased, but the elongation and grain size were decreased. However, increasing the annealing temperature, the tensile strength and hardness were decreased, but the elongation and grain size were increased. Increasing the carbon content of Alloy 690, the results of the anodic polarization tests and the nitric acid immersion tests showed that the annealing temperature to reveal a minimum corrosion rate was increased. This behavior seemed to be due to the combination of the solid solution of carbon in the matrix and grain growth with annealing. In this work, the corrosion properties of Alloy 690 were better than that of Alloy 600, and the range of the optimum annealing temperature of Alloy 690 was from 1100 .deg. C to 1150 .deg. C

Methods for batch fabrication of cold cathode vacuum switch tubes

Science.gov (United States)

Walker, Charles A [Albuquerque, NM; Trowbridge, Frank R [Albuquerque, NM

2011-05-10

Methods are disclosed for batch fabrication of vacuum switch tubes that reduce manufacturing costs and improve tube to tube uniformity. The disclosed methods comprise creating a stacked assembly of layers containing a plurality of adjacently spaced switch tube sub-assemblies aligned and registered through common layers. The layers include trigger electrode layer, cathode layer including a metallic support/contact with graphite cathode inserts, trigger probe sub-assembly layer, ceramic (e.g. tube body) insulator layer, and metallic anode sub-assembly layer. Braze alloy layers are incorporated into the stacked assembly of layers, and can include active metal braze alloys or direct braze alloys, to eliminate costs associated with traditional metallization of the ceramic insulator layers. The entire stacked assembly is then heated to braze/join/bond the stack-up into a cohesive body, after which individual switch tubes are singulated by methods such as sawing. The inventive methods provide for simultaneously fabricating a plurality of devices as opposed to traditional methods that rely on skilled craftsman to essentially hand build individual devices.

Enteral Feeding Tubes in Patients Undergoing Definitive Chemoradiation Therapy for Head-and-Neck Cancer: A Critical Review

Energy Technology Data Exchange (ETDEWEB)

Koyfman, Shlomo A., E-mail: koyfmas@ccf.org [Departments of Radiation and Solid Tumor Oncology, Taussig Cancer Institute, Cleveland Clinic, Cleveland, Ohio (United States); Adelstein, David J. [Departments of Radiation and Solid Tumor Oncology, Taussig Cancer Institute, Cleveland Clinic, Cleveland, Ohio (United States)

2012-11-01

Definitive chemoradiation therapy has evolved as the preferred organ preservation strategy in the treatment of locally advanced head-and-neck cancer (LA-HNC). Dry mouth and dysphagia are among the most common and most debilitating treatment-related toxicities that frequently necessitate the placement of enteral feeding tubes (FT) in these patients to help them meet their nutritional requirements. The use of either a percutaneous endoscopic gastrostomy tube or a nasogastric tube, the choice of using a prophylactic vs a reactive approach, and the effects of FTs on weight loss, hospitalization, quality of life, and long-term functional outcomes are areas of continued controversy. Considerable variations in practice patterns exist in the United States and abroad. This critical review synthesizes the current data for the use of enteral FTs in this patient population and clarifies the relative advantages of different types of FTs and the timing of their use. Recent developments in the biologic understanding and treatment approaches for LA-HNC appear to be favorably impacting the frequency and severity of treatment-related dysphagia and may reduce the need for enteral tube feeding in the future.

Stress relief to prevent stress corrosion in the transition region of expanded Alloy 600 steam-generator tubing. Final report

International Nuclear Information System (INIS)

Woodward, J.; van Rooyen, D.

1983-05-01

The feasibility of preventing primary side roll transition cracking has been investigated, using induction heating to attain stress relief of expanded Ni-Cr-Fe Alloy 600 steam generator tubing. Work on rolled tubing and U-bends has shown that temperatures with which stress relief can be obtained range from 700 to 850 0 C, with lower temperatures in this range requiring longer times at temperature to provide the requisite reduction in residual stresses. No work has yet been done outside this range. Preliminary tests, using induction heating, have been carried out on a mock tube sheet assembly, designed to the dimensions of a typical steam generator, and have identified the type of heating/cooling cycle that would occur in the tube sheet during a stress relief operation. Preliminary results show that the times to reach the higher temperatures in the range observed to give stress relief, of the order of 850 0 C, can be as short as 8 seconds, and less with optimum coil design and power control

Heat removal capability of divertor coaxial tube assembly

International Nuclear Information System (INIS)

Shibui, Masanao; Nakahira, Masataka; Tada, Eisuke; Takatsu, Hideyuki

1994-05-01

To deal with high power flowing in the divertor region, an advanced divertor concept with gas target has been proposed for use in ITER/EDA. The concept uses a divertor channel to remove the radiated power while allowing neutrals to recirculate. Candidate channel wall designs include a tube array design where many coaxial tubes are arranged in the toroidal direction to make louver. The coaxial tube consists of a Be protection tube encases many supply tubes wound helically around a return tube. V-alloy and hardened Cu-alloy have been proposed for use in the supply and return tubes. Some coolants have also been proposed for the design including pressurized He and liquid metals, because these coolants are consistent with the selection of coolants for the blanket and also meet the requirement of high temperature operation. In the coaxial tube design, the coolant area is restricted and brittle Be material is used under severe thermal cyclings. Thus, to obtain the coaxial tube with sufficient safety margin for the expected fusion power excursion, it is essential to understand its applicability limit. The paper discusses heat removal capability of the coaxial tube and recommends some design modifications. (author)

The effect lead impurities on the corrosion resistance of alloy 600 and alloy 690 in high temperature water

International Nuclear Information System (INIS)

Sakai, T.; Nakagomi, N.; Kikuchi, T.; Aoki, K.; Nakayasu, F.; Yamakawa, K.

1998-01-01

Degradation of nickel-based alloy steam generator (SG) tubing caused by lead-induced corrosion has been reported recently in some PWR plants. Several laboratory studies also have shown that lead causes intergranular or transgranular stress corrosion cracking (IGSCC or TGSCC) of the tubing materials. Information from previous studies suggests two possible explanations for the mechanism of lead-induced corrosion. One is selective dissolution of tube metal elements, resulting in formation of a lead-containing nickel-depleted oxide film as observed in mildly acidic environments. The other explanation is an increase in potential, as has been observed in lead-contaminated caustic environments, although not in all volatile treatment (AVT) water such as the ammonium-hydrazine water chemistry. These observation suggest that an electrochemical reaction between metal elements and dissolved lead might be the cause of lead-induced corrosion. The present work was undertaken to clarify the lead-induced corrosion mechanism of nickel-based alloys from an electrochemical viewpoint, focusing on mildly acidic and basic environments. These are the probable pH conditions in the crevice region between the tube and tube support plate of the SG where corrosion damage could occur. Measurements of corrosion potential and electrochemical polarization of nickel-based alloys were performed to investigate the effect of lead on electrochemical behavior of the alloys. Then, constant extension rate tests (CERT) were carried out to determine the corrosion susceptibility of the alloys in a lead-contaminated environment. (J.P.N.)

Characteristics of hydrostatically extruded Zr-2.5Nb alloy

International Nuclear Information System (INIS)

Jie, Z.; Jiaqi, D.; Tieqi, Y.; Wenxian, H.; Yan, L.; Yunxia, Z.; Zhenhe, L.

1984-01-01

Hydrostatic extrusion is a new production technology. Zr-2.5Nb alloy tubes cold hydrostatically extruded possess excellent mechanical properties similar to heat-treated tubes and better than cold-worked tubes. Examination by transmission electron microscope shows that the alloy is of a uniform cell substructure containing the (α + β) phases, which is one of important factors improving properties of the alloy. The study of texture, stress, and reorientation of the hydride shows that hydrostatically extruded tubes with basal plane normals in the radial direction have obviously higher hydride reorientation threshold stress than tubes with basal plane normals in the circumferential direction. Moreover, investigation of fracture toughness reveals that hydride distributed perpendicular to the crack propagation direction restrains further propagation of the crack. It is favorable for preserving the fracture resistance of the material

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

Science.gov (United States)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Schaefer, L.; Kempe, H.

1983-12-01

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

The influence of lead on stress corrosion cracking of steam generator tubing

International Nuclear Information System (INIS)

Ryan Curtis Wolfe

2015-01-01

Lead (Pb) is present at low concentrations on the secondary side of steam generators, but is known to accumulate in steam generator sludge and become concentrated in crevices and cracks. Pb is known to have played a role in the degradation of Alloy 600MA tubing, necessitating the replacement of those steam generators. There is new evidence which indicates that Pb has also played a role in the stress corrosion cracking (SCC) of Alloy 600TT. Furthermore. laboratory testing indicates that advanced tubing alloys such as Alloy 690TT and Alloy 800NG area also susceptible to this attack. In response to these vulnerabilities, utilities are attempting to manufacture tubing using processes which will impart optimal corrosion resistance, fabricate and operate SG's to minimize stress in the tubing, undertake efforts to identify and remove the sources of Pb, reduce the existing inventory of Pb using chemical or mechanical cleaning processes, and maintain rigorous chemistry controls. Research is warranted to qualify chemical methods to mitigate PbSCC that may be observed in service. This presentation will review work performed through the Electric Power Research Institute (EPRI) to address the issue of Pb-assisted stress corrosion cracking of steam generator tubing. (author)

Microstructure and textural characterization of hot extruded Zr-2.5Nb alloy PHWR pressure tube fabricated by various ingot processing route

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Jha, S.K.; Saibaba, N.; Neogy, S.; Mani Krishna, K.V.; Srivastava, D.; Dey, G.K.

2011-01-01

Zr-2.5 Nb alloys finds its applications as a pressure tube component in pressure tube type thermal reactors such as PHWRs and RBMK due to properties attributed such as low neutron absorption cross section, high temperature strength and corrosion resistance etc. Manufacturing of this life time components involves series of thermo-mechanical processes of hot working and cold working with intermediate annealing. The life time of Pressure tube are limited due to their diametral creep properties which is governed by metallurgical characteristics such as texture, microstructure dislocation density etc. The primary breakdown of cast structure in Vacuum Arc Melted ingot can be effected by either hot extrusion or forging in single or multiple stages before final hot extrusion step into the blank for manufacturing of seamless pressure tube. Elevated temperature deformation carried out in hot working above the recrystallization temperature would enable impositions of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on process parameters such as extrusion ratio, temperature and strain rate. Basic microstructure developed at this deformation stage has significant bearing on the final properties of the material fabricated with subsequent cold working steps. The major texture in α+β Zr-2.5 Nb alloy is established during final extrusion to blank which does not change significantly during subsequent cold pilgering. However, microstructure is modified significantly in subsequent cold working which can be effected by cold pilgering or cold drawing in single or multiple steps. Present paper brings out the various ingot processing routes using forging and or extrusion followed for fabrication of pressure tubes. The development of texture and microstructures has been discussed at the blank stage from these processing routes and also with respect to varying extrusion variable such as extrusion ratio

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.

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)

Comparison of evaluation method for planar flaw in pressure tube

International Nuclear Information System (INIS)

Choi, Sung Nam; Kim, Hyung Nam; Yoo, Hyun Joo; Hwang, Won Gul

2009-01-01

CSA N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA N285.8-05, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of FFSG(Fitness For Service Guideline for Zirconium alloy pressure in operation CANDU) used now. The object of this paper is to address the fracture initiation and plastic collapse evaluation for the planar flaw as it applies to the pressure tube on Wolsong NPP.

Inelastic analysis of finite length and depth cracked tubes

International Nuclear Information System (INIS)

Reich, M.; Prachuktam, S.; Gardner, D.

1977-01-01

Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdowns. Typical PWR steam generator units contain thousands of long straight tubes with U-bend sections. These tubes are primarily made from alloy 600 and their sizes vary between 3 / 4 '' and 7 / 8 '' (1.905 cm and 2.223 cm) in diameter with nominal thicknesses of 0.043'' to 0.053'' (0.109 cm to 0.135 cm). Since alloy 600 (and the previously used 304-SS tubes) are ductile, high toughness materials LEFM (linear elastic fracture mechanics) criteria do not apply. This paper concerns itself with the prediction of the failure pressures for typical PWR steam generator tubes with longitudinal finite length and finite depth cracks. Only local plastic overload failure is considered

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2 field testing

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L.; Seitz, W.W.; Girshik, A. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1998-06-01

In Phase 1 of this project, laboratory experiments were performed on a variety of developmental and commercial tubing alloys and claddings by exposing them to fireside corrosion tests which simulated a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347, RA85H, HR3C, RA253MA, Fe{sub 3}Al + 5Cr, Ta-modified 310, NF 709, 690 clad, 671 clad, and 800HT for up to approximately 16,000 hours to the actual operating conditions of a 250-MW, coal-fired boiler. The samples were installed on air-cooled, retractable corrosion probes, installed in the reheater cavity, and controlled to the operating metal temperatures of an existing and advanced-cycle, coal-fired boiler. Samples of each alloy were exposed for 4,483, 11,348, and 15,883 hours of operation. The present results are for the metallurgical examination of the corrosion probe samples after the full 15,883 hours of exposure. A previous topical report has been issued for the 4,483 hours of exposure.

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Evaluation of the use of automatic exposure control and automatic tube potential selection in low-dose cerebrospinal fluid shunt head CT

Energy Technology Data Exchange (ETDEWEB)

Wallace, Adam N.; Bagade, Swapnil; Chatterjee, Arindam; Hicks, Brandon; McKinstry, Robert C. [Barnes Jewish Hospital, Mallinckrodt Institute of Radiology, St. Louis, MO (United States); Washington University School of Medicine, St. Louis, MO (United States); Vyhmeister, Ross [Washington University School of Medicine, St. Louis, MO (United States); Ramirez-Giraldo, Juan Carlos [Siemens Healthcare, Malvern, PA (United States)

2015-03-17

Cerebrospinal fluid shunts are primarily used for the treatment of hydrocephalus. Shunt complications may necessitate multiple non-contrast head CT scans resulting in potentially high levels of radiation dose starting at an early age. A new head CT protocol using automatic exposure control and automated tube potential selection has been implemented at our institution to reduce radiation exposure. The purpose of this study was to evaluate the reduction in radiation dose achieved by this protocol compared with a protocol with fixed parameters. A retrospective sample of 60 non-contrast head CT scans assessing for cerebrospinal fluid shunt malfunction was identified, 30 of which were performed with each protocol. The radiation doses of the two protocols were compared using the volume CT dose index and dose length product. The diagnostic acceptability and quality of each scan were evaluated by three independent readers. The new protocol lowered the average volume CT dose index from 15.2 to 9.2 mGy representing a 39 % reduction (P < 0.01; 95 % CI 35-44 %) and lowered the dose length product from 259.5 to 151.2 mGy/cm representing a 42 % reduction (P < 0.01; 95 % CI 34-50 %). The new protocol produced diagnostically acceptable scans with comparable image quality to the fixed parameter protocol. A pediatric shunt non-contrast head CT protocol using automatic exposure control and automated tube potential selection reduced patient radiation dose compared with a fixed parameter protocol while producing diagnostic images of comparable quality. (orig.)

Effect of Ti3+ ion on the Corrosion Behavior of Alloy 600

International Nuclear Information System (INIS)

Lee, Chang Bong; Lim, Han Gwi; Kim, Bok Hee; Kim, Ki Ju

1999-01-01

Alloy 600 has been widely used as a steam generator tubing material in pressurized water reactors(PWRs) nuclear power plants. Corrosion of steam generator tubing mainly occurs on the secondary water side. The purpose of this work is primarily concerned with examining the effect of Ti 3+ ion concentrations on the corrosion behavior of the Alloy 600 steam generator tubing material. Corrosion behavior of the Alloy 600 steam generator tubing material was studied in aqueous solutions with varying Ti 3+ ion concentration at room temperature. Potentiodynamic and potentiostatic polarization techniques were used to determine the corrosion and pitting potentials for the Alloy 600 test material. The addition of Ti 3+ ion to 1000ppm, showed inhibition effect on the corrosion of Alloy 600. But the corrosion of Alloy 600 was accelerated when the concentration of Ti 3+ ion exceeded 1000ppm, it is assumed that the effect of general corrosion of Alloy 600 is more sensitive than pitting corrosion. It is considered that the passive film which was formed on the Alloy 600 surface in the 100ppm Ti 3+ ion containing solution is mainly consisted of TiO 2

Study on corrosion of thermal power plant condenser tubes

Energy Technology Data Exchange (ETDEWEB)

Mohammadi, Abdolreza Rashidi; Zhaam, Ali Akbar [Niroo Research Institute, end of Poonak Bakhtari blvd., Shahrak Ghods, Tehran (Iran)

2004-07-01

The aim of this investigation is to study kinds of corrosion mechanisms in thermal power plant condenser tubes. Condenser is a shell and tube heat exchanger in which cooling water flows through its tubes. While the steam from low pressure turbine passes within condenser tubes, it is condensed by cooling water. The exhausted steam from low pressure turbine is condensed on external surface of condenser tubes and heat is transferred to cooling water which flow into tubes. Tubes composition is usually copper-based alloys, stainless steel or titanium. Annual damages due to corrosion cause much cost for replacement and repairing metallic equipment and installations in electric power industry. Because of existence of different contaminants in water and steam cycle, condenser tubes surfaces are exposed to corrosion. Contaminants like oxygen, carbon dioxide, chloride ion and ammonia in water and steam cycle originate several damages such as pitting and crevice corrosion, erosion, galvanic attack, SCC, condensed corrosion, de-alloying in thermal power plant condenser. The paper first states how corrosion damage takes place in condensers and then introduces types of usual alloys used in condensers and also their corrosion behavior. In continuation, a brief explanation is presented about kinds of condenser failures due to corrosion. Then, causes and locations of different mechanisms of corrosion events on condenser tubes and effects of different parameters such as composition, temperature, chloride and sulfide ion concentration, pH, water velocity and biological precipitation are examined and finally protection methods are indicated. Also some photos of tubes specimens related to power plants are studied and described in each case of mentioned mechanisms. (authors)

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

Energy Technology Data Exchange (ETDEWEB)

Huang, Jun-Lin; Zhou, Ke-Yi, E-mail: boiler@seu.edu.cn; Xu, Jian-Qun [Key Laboratory of Energy Thermal Conversion and Control of Ministry of Education, School of Energy and Environment, Southeast University, Nanjing 210096, Jiangsu Province (China); Wang, Xin-Meng; Tu, Yi-You [School of Materials Science and Engineering, Southeast University, Nanjing 210096, Jiangsu Province (China)

2014-07-28

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

Science.gov (United States)

Huang, Jun-Lin; Zhou, Ke-Yi; Wang, Xin-Meng; Tu, Yi-You; Xu, Jian-Qun

2014-07-01

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

Acoustic emission analysis on tensile failure of steam-side oxide scales formed on T22 alloy superheater tubes

International Nuclear Information System (INIS)

Huang, Jun-Lin; Zhou, Ke-Yi; Xu, Jian-Qun; Wang, Xin-Meng; Tu, Yi-You

2014-01-01

Failure of steam-side oxide scales on boiler tubes can seriously influence the safety of coal-fired power plants. Uniaxial tensile tests employing acoustic emission (AE) monitoring were performed, in this work, to investigate the failure behavior of steam-side oxide scales on T22 alloy boiler superheater tubes. The characteristic frequency spectra of the captured AE signals were obtained by performing fast Fourier transform. Three distinct peak frequency bands, 100-170, 175-250, and 280-390 kHz, encountered in different testing stages were identified in the frequency spectra, which were confirmed to, respectively, correspond to substrate plastic deformation, oxide vertical cracking, and oxide spalling with the aid of scanning electronic microscopy observations, and can thus be used for distinguishing different oxide failure mechanisms. Finally, the critical cracking strain of the oxide scale and the interfacial shear strength of the oxide/substrate interface were estimated, which are the critical parameters urgently desired for modeling the failure behavior of steam-side oxide scales on boiler tubes of coal-fired power plants.

A computational study of low-head direct chill slab casting of aluminum alloy AA2024

Science.gov (United States)

Hasan, Mainul; Begum, Latifa

2016-04-01

The steady state casting of an industrial-sized AA2024 slab has been modeled for a vertical low-head direct chill caster. The previously verified 3-D CFD code is used to investigate the solidification phenomena of the said long-range alloy by varying the pouring temperature, casting speed and the metal-mold contact heat transfer coefficient from 654 to 702 °C, 60-180 mm/min, and 1.0-4.0 kW/(m2 K), respectively. The important predicted results are presented and thoroughly discussed.

Neutronographic Texture Analysis of Zirconium Based Alloys

International Nuclear Information System (INIS)

Kruz'elová, M; Vratislav, S; Kalvoda, L; Dlouhá, M

2012-01-01

Neutron diffraction is a very powerful tool in texture analysis of zirconium based alloys used in nuclear technique. Textures of five samples (two rolled sheets and three tubes) were investigated by using basal pole figures, inversion pole figures, and ODF distribution function. The texture measurement was performed at diffractometer KSN2 on the Laboratory of Neutron Diffraction, Department of Solid State Engineering, Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague. Procedures for studying textures with thermal neutrons and procedures for obtaining texture parameters (direct and inverse pole figures, three dimensional orientation distribution function) are also described. Observed data were processed by software packages HEXAL and GSAS. Our results can be summarized as follows: i) All samples of zirconium alloys show the distribution of middle area into two maxima in basal pole figures. This is caused by alloying elements. A characteristic split of the basal pole maxima tilted from the normal direction toward the transverse direction can be observed for all samples, ii) Sheet samples prefer orientation of planes (100) and (110) perpendicular to rolling direction and orientation of planes (002) perpendicular to normal direction, iii) Basal planes of tubes are oriented parallel to tube axis, meanwhile (100) planes are oriented perpendicular to tube axis. Level of resulting texture and maxima position is different for tubes and for sheets. The obtained results are characteristic for zirconium based alloys.

Energy conservation in two wheeler head tube welding fixture by modeling

Science.gov (United States)

Kamalanathan, S.; Guruprasad, B.; Elango, A.

2018-05-01

Energy conservation in automotive industry plays a significant role to increase the productivity which reduces the Men, Material, and Machinery. The Automotive industry sector is one of the major sector which works in two or more times of loading of work pieces in welding fixture. It consumes more energy. This project focuses on reduce the energy consumption by applying single time loading of work pieces with minimum number of labours, machine and reduces scrap. Welding fixtures are designed for the components which are difficult to weld in normal way or without any holding unit. The fixture is to be designed for the two wheeler head tube assembly which is to be welded with its companion for its application. It is demonstrated by modelling in Uni-graphics Software and FEA analysis will be done by ANSYS and experimentally products are tested and execute in industry. A code of practice suggested establishing acceptable standard for energy used in Automotive industry.

Microstructure and Mechanical Properties of Heat-Treated B319 Alloy Diesel Cylinder Heads

Science.gov (United States)

Chaudhury, S. K.; Apelian, D.; Meyer, P.; Massinon, D.; Morichon, J.

2015-07-01

Microstructure and mechanical properties of B319 alloy diesel cylinder heads were investigated in this study. Cylinder heads were heat treated to T5, T6, and T7 tempers using fluidized bed technology. Three different fluidized beds were used, each to solutionize, quench, and age the castings. For comparative purposes, castings were also aged using conventional forced-air circulation electric-resistance furnace. Effects of processing parameters such as temperature, time, and heating rate on microstructural evolution and mechanical properties namely tensile properties and hardness of B319 alloy castings were studied. The number density and size range of precipitates were measured. Results show that the T5 temper has no effect on eutectic phases such as Si- and Fe-rich intermetallic, and Al2Cu. On contrary, both T6 and T7 tempers result in spherodization of the eutectic Si and partial dissolution of the Al2Cu phase. Prolonged solution heat treatment for 8 hours in fluidized bed results in limited dissolution of the secondary eutectic Al2Cu phase. Aging (T6, T7, and T5) results in precipitation of Al5Cu2Mg8Si6 and Al2Cu phases in B319 alloy. The number density of precipitates in T6 temper is greater than in T7 and T5 tempers. The number density of precipitates is also affected by the duration of solution heat treatment. In general, long solution heat treatment (8 hours) results in greater precipitate density than short solution treatment (2 hours). The distribution of precipitates is inhomogeneous and varied across the dendritic structure. In general, precipitation rate of Al5Cu2Mg8Si6 phase is greater near the periphery of the dendrite as compared to the center. This is because Al5Cu2Mg8Si6 nucleates on Si particle, grain boundaries, and triple junction between recrystallized Al grains and Si particles. Similarly, heterogeneous sites such as grain boundaries and Al/Si interface also act as nucleating sites for the precipitation of Al2Cu phase. In general, the

Final Technical Report - High-Performance, Oxide-Dispersion-Strengthened Tubes for Production of Ethylene adn Other Industrial Chemicals

Energy Technology Data Exchange (ETDEWEB)

McKimpson, Marvin G.

2006-04-06

This project was undertaken by Michigan Technological University and Special Metals Corporation to develop creep-resistant, coking-resistant oxide-dispersion-strengthened (ODS) tubes for use in industrial-scale ethylene pyrolysis and steam methane reforming operations. Ethylene pyrolysis tubes are exposed to some of the most severe service conditions for metallic materials found anywhere in the chemical process industries, including elevated temperatures, oxidizing atmospheres and high carbon potentials. During service, hard deposits of carbon (coke) build up on the inner wall of the tube, reducing heat transfer and restricting the flow of the hydrocarbon feedstocks. About every 20 to 60 days, the reactor must be taken off-line and decoked by burning out the accumulated carbon. This decoking costs on the order of $9 million per year per ethylene plant, accelerates tube degradation, and requires that tubes be replaced about every 5 years. The technology developed under this program seeks to reduce the energy and economic cost of coking by creating novel bimetallic tubes offering a combination of improved coking resistance, creep resistance and fabricability not available in current single-alloy tubes. The inner core of this tube consists of Incoloy(R) MA956, a commercial ferritic Fe-Cr-Al alloy offering a 50% reduction in coke buildup combined with improved carburization resistance. The outer sheath consists of a new material - oxide dispersion strengthened (ODS) Alloy 803(R) developed under the program. This new alloy retains the good fireside environmental resistance of Alloy 803, a commercial wrought alloy currently used for ethylene production, and provides an austenitic casing to alleviate the inherently-limited fabricability of the ferritic Incoloy(R) MA956 core. To provide mechanical compatibility between the two alloys and maximize creep resistance of the bimetallic tube, both the inner Incoloy(R) MA956 and the outer ODS Alloy 803 are oxide dispersion

Development of process route for the production of Fe-0.12C-9CR-2W-0.35Y2O3 ODS alloy tubing for Indian FBR application

International Nuclear Information System (INIS)

Lakshminarayana, B.; Tonpe, S.; Jha, S.K.; Kapoor, Komal; Dubey, A.K.; Gurunadh, J.; Surender, A.; Deshpande, K.V.K.; Maity, P.K.

2011-01-01

In the wake of Nuclear Renaissance, India is playing key role in generation of clean and green Nuclear Energy. It has entered into its second stage Nuclear Power Program on commercial scale with the commencement of construction of 500 MWe Prototype Fast Breeder Reactor (PFBR) at Kalpakkam. Nuclear Fuel Complex (NFC), Hyderabad is playing a crucial role in the manufacture of all the critical sub-assemblies in SS (D9) grade materials for this reactor. The SS(D9) material with controlled cold work is having very good void swelling resistance and high temperature properties, which can sustain fluence of 100 dpa. The paper covers the manufacturing process and characterization of the ODS tubes for fuel clad application. Manufacturing of 9 Cr 2W Y 2 O 3 - ODS martensitic steel fuel cladding tube has been taken up in Nuclear Fuel Complex, Hyderabad with mechanical alloying followed by MS canning of mechanically alloyed powder, upsetting and hot extrusion and subsequently thermo mechanical process. Manufacturing technology of ODS steel tube is critical with respect its chemical composition, dimensional tolerances, Y 2 O 3 particle size and its distribution and achievement of mechanical properties with proper combination of cold working and heat treatment. The paper covers the manufacturing process and characterization of the ODS tubes for fuel clad application. Manufacturing process for the production of ODS alloy (9 Cr 2W Y 2 O 3 - ODS) has been optimized for mass scale production at NFC

Tubing crimping pliers

Science.gov (United States)

Lindholm, G.T.

1981-02-27

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

Delayed hydride cracking in Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Mieza, Juan I.; Domizzi, Gladys; Vigna, Gustavo L.

2007-01-01

Zr-2.5 Nb alloy from CANDU pressure tubes are prone to failure by hydrogen intake. One of the degradation mechanisms is delayed hydride cracking, which is characterized by the velocity of cracking. In this work, we study the effect of beta zirconium phase transformation over delayed hydride cracking velocity in Zr-2.5 Nb alloy from pressure tubes. Acoustic emission technique was used for cracking detection. (author) [es

Prospects for zirconium structural alloys at high temperatures

International Nuclear Information System (INIS)

Thomas, W.R.

1969-05-01

Improved station efficiencies and lower capital costs provide incentives for the development of zirconium alloys for pressure tubes which can operate at temperatures above 450 o C. The experience of the Ti industry indicates that a complex alloy containing solution hardeners of Sn or Al and precipitation hardeners of Mo and Nb and perhaps Si will be required. The thermal neutron cross-section of the alloy will be about 10% higher than Zircaloy-2 and because of its poor corrosion resistance will require cladding with a corrosion resistant alloy such as Zr-Cr. Results to date indicate that such a pressure tube is feasible. (author)

Development of microstructure in thermomechanical processing of zirconium alloys

International Nuclear Information System (INIS)

Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2009-01-01

Zirconium based alloys are used for the manufacture of fuel tubes pressure tubes calandria tubes and other components of Pressurized Heavy Water Reactors (PHWRS). In single or two phase zirconium alloy system a variety of microstructure can be generated by suitable heat treatments by the process of equilibrium and non equilibrium phase transformations Microstructure can also be modified by alloying with α and β stabilizers. The microstructure in Zr alloys could be single hexagonal phase (α alloys) two phase bcc and hexagonal (α + β alloys) phase, single metastable martensitic microstructure and β with ω phase. The microstructural and micro textural evolution during thermo mechanical treatments depends strongly on such initial microstructure. Hot extrusion is a significant bulk deformation step which decides the initial microstructure of the alloy. It is carried out at elevated temperature i e above the recrystallization temperature, which enable imposition of large strains in single step. This deformation causes a significant change in the microstructure of the material and depends on extrusion process parameters such as temperature, strain rate (Ram speed), reduction ratio etc. In the present paper development of microstructures, microtexture and texture have been examined. An attempt is also made to optimise the hot working parameters for different Zirconium alloys with help of these studies. (author)

Evaluation of the use of automatic exposure control and automatic tube potential selection in low-dose cerebrospinal fluid shunt head CT.

Science.gov (United States)

Wallace, Adam N; Vyhmeister, Ross; Bagade, Swapnil; Chatterjee, Arindam; Hicks, Brandon; Ramirez-Giraldo, Juan Carlos; McKinstry, Robert C

2015-06-01

Cerebrospinal fluid shunts are primarily used for the treatment of hydrocephalus. Shunt complications may necessitate multiple non-contrast head CT scans resulting in potentially high levels of radiation dose starting at an early age. A new head CT protocol using automatic exposure control and automated tube potential selection has been implemented at our institution to reduce radiation exposure. The purpose of this study was to evaluate the reduction in radiation dose achieved by this protocol compared with a protocol with fixed parameters. A retrospective sample of 60 non-contrast head CT scans assessing for cerebrospinal fluid shunt malfunction was identified, 30 of which were performed with each protocol. The radiation doses of the two protocols were compared using the volume CT dose index and dose length product. The diagnostic acceptability and quality of each scan were evaluated by three independent readers. The new protocol lowered the average volume CT dose index from 15.2 to 9.2 mGy representing a 39 % reduction (P < 0.01; 95 % CI 35-44 %) and lowered the dose length product from 259.5 to 151.2 mGy/cm representing a 42 % reduction (P < 0.01; 95 % CI 34-50 %). The new protocol produced diagnostically acceptable scans with comparable image quality to the fixed parameter protocol. A pediatric shunt non-contrast head CT protocol using automatic exposure control and automated tube potential selection reduced patient radiation dose compared with a fixed parameter protocol while producing diagnostic images of comparable quality.

Advanced pressure tube sampling tools

International Nuclear Information System (INIS)

Wittich, K.C.; King, J.M.

2002-01-01

Deuterium concentration is an important parameter that must be assessed to evaluate the Fitness for service of CANDU pressure tubes. In-reactor pressure tube sampling allows accurate deuterium concentration assessment to be made without the expenses associated with fuel channel removal. This technology, which AECL has developed over the past fifteen years, has become the standard method for deuterium concentration assessment. AECL is developing a multi-head tool that would reduce in-reactor handling overhead by allowing one tool to sequentially sample at all four axial pressure tube locations before removal from the reactor. Four sets of independent cutting heads, like those on the existing sampling tools, facilitate this incorporating proven technology demonstrated in over 1400 in-reactor samples taken to date. The multi-head tool is delivered by AECL's Advanced Delivery Machine or other similar delivery machines. Further, AECL has developed an automated sample handling system that receives and processes the tool once out of the reactor. This system retrieves samples from the tool, dries, weighs and places them in labelled vials which are then directed into shielded shipping flasks. The multi-head wet sampling tool and the automated sample handling system are based on proven technology and offer continued savings and dose reduction to utilities in a competitive electricity market. (author)

Heat treated tube for cladding nuclear fuel element

International Nuclear Information System (INIS)

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

1983-01-01

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

Influence of microstructure modification on the circumferential creep of Zr–Nb–Sn–Fe cladding tubes

International Nuclear Information System (INIS)

Jeong, Gu Beom; Kim, In Won; Hong, Sun Ig

2016-01-01

Out-of-reactor, non-irradiated thermal creep performances and lives of annealed and stress-relieved Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were studied and compared. The creep rates of annealed Zr-1.02Nb-0.69Sn-0.12Fe cladding tubes were appreciably slower than those of stress-relieved annealed counterpart. The stress exponent increased slightly from 5.1 to 6.1 in the stress-relieved cladding to 5.3–6.3 in the annealed cladding. The creep activation energy of the annealed Zr-1.02Nb-0.69Sn-0.12Fe alloy (300–330 kJ/mol) was larger compared to that of the stress-relieved alloy (210–260 kJ/mol). The creep activation energy of annealed alloy is close to that of self-diffusion in α-Zr (336 kJ/mol). The smaller activation energy in the stress-relieved alloy is attributed to the increasing contribution of faster diffusion path such as grain boundaries and dislocations. The presence of dislocation arrays with higher dislocation density and smaller grain size in the stress-relived alloy was confirmed by TEM analysis. The creep rupture time increased dramatically in the annealed Zr–1Nb- 0.7Sn-0.1Fe alloy compared to that of stress-relieved alloy, supporting the decrease of creep rate by annealing. The creep life of Zr-1.02Nb-0.69Sn-0.12Fe claddings can be extended through microstructure modification by annealing at intermediate temperatures in which dislocation creep dominates. - Highlights: • Effect of microstructure modification on creep in Zr–Nb–Sn–Fe tubes was studied. • Creep activation energy in annealed tubes was larger than in stress-relieved tubes. • Lower dislocation density in lager grains was observed after creep in annealed tubes. • Larson–Miller parameter of annealed tube was larger than that of stress-relieved one. • Creep life of tubes was extended through microstructure modification by annealing.

Analysis of Reactor Vessel Lower Head Penetration Tube Failure

International Nuclear Information System (INIS)

Stempniewicz, Marek

1999-01-01

This paper presents results of two studies, performed to investigate the behavior of the reactor vessel penetration tubes in case of relocation of molten material into the tubes. The first study is on the CORVIS drain line experiment 03/1. Results of pre-test calculations are presented, and compared to the later obtained experimental data. The timing of the drain line melting and the velocity of the debris flowing inside the drain line were predicted correctly, but the penetration depth was clearly underestimated. If the calculations are done using different correlation for the melt-to-wall convective heat transfer, the results are closer to the experiment. It cannot however be concluded that the alternative correlation is more appropriate until other uncertainties are clarified. The second study presents calculations performed for GKN Dodewaard CRD, instrument tubes and drain line. Calculations were performed to estimate whether the tubes have a chance to withstand the first attack of the melt and thus postpone vessel failure until the water in the lower plenum evaporates. Calculations were performed assuming that the melt can move into the tubes without any resistance, e.g. presence of water in the tubes was not taken into account. The results indicate that the critical penetration of the GKN vessel, which is most likely to fail, is the drain line. Results also indicate that external flooding should prevent early tube failure, at least in case of low vessel pressure. (author)

Evaluation of crack propagation of alloy 600 tube in high temperature water, (1)

International Nuclear Information System (INIS)

Hirano, Hideo; Kawamura, H.; Kawamura, Kohji; Matsubara, Masaaki

1990-01-01

This report describes the analysis of stress intensity factors at cracks in alloy 600 steam generator tubes. Based on the results of the analysis, IGA/SCC tests were carried out to examine the effect of stress intensity and water quality on the crack propagation rate. The main test result are as follows: (1) Hoop stress was caused by the pressure difference between the internal and external surface of the steam generator tube. The calculated hoop stress was about 7 kg/mm 2 . In addition, the temperature difference between the internal and external surface caused thermal stress. The thermal stress was about 10 kg/mm 2 at the external surface and the one at the internal surface was about -10 kg/mm 2 . Total stress at the external and internal surface was 17 kg/mm 2 and -3 kg/mm 2 , respectively. (2) The stress intensity factor at the crack tip increased with increasing crack length. For a long crack, the stress intensity factor decreased with increasing crack number. However, for a short crack, the stress intensity factor decreased little with increasing crack number. (3) Under high stress-intensity conditions, i.e. 40∼50 kg·mm -3/2 , the IGA/SCC test showed that IGA/SCC propagated in AVT and AVT/boric-acid solution at 320degC and 350degC. However, the propagation rate was low. (author)

Regional neural tube closure defined by the Grainy head-like transcription factors.

Science.gov (United States)

Rifat, Yeliz; Parekh, Vishwas; Wilanowski, Tomasz; Hislop, Nikki R; Auden, Alana; Ting, Stephen B; Cunningham, John M; Jane, Stephen M

2010-09-15

Primary neurulation in mammals has been defined by distinct anatomical closure sites, at the hindbrain/cervical spine (closure 1), forebrain/midbrain boundary (closure 2), and rostral end of the forebrain (closure 3). Zones of neurulation have also been characterized by morphologic differences in neural fold elevation, with non-neural ectoderm-induced formation of paired dorso-lateral hinge points (DLHP) essential for neural tube closure in the cranial and lower spinal cord regions, and notochord-induced bending at the median hinge point (MHP) sufficient for closure in the upper spinal region. Here we identify a unifying molecular basis for these observations based on the function of the non-neural ectoderm-specific Grainy head-like genes in mice. Using a gene-targeting approach we show that deletion of Grhl2 results in failed closure 3, with mutants exhibiting a split-face malformation and exencephaly, associated with failure of neuro-epithelial folding at the DLHP. Loss of Grhl3 alone defines a distinct lower spinal closure defect, also with defective DLHP formation. The two genes contribute equally to closure 2, where only Grhl gene dosage is limiting. Combined deletion of Grhl2 and Grhl3 induces severe rostral and caudal neural tube defects, but DLHP-independent closure 1 proceeds normally in the upper spinal region. These findings provide a molecular basis for non-neural ectoderm mediated formation of the DLHP that is critical for complete neuraxis closure. (c) 2010 Elsevier Inc. All rights reserved.

Experimental study and modeling of high-temperature oxidation and phase transformation of cladding-tubes made in zirconium alloy

International Nuclear Information System (INIS)

Mazeres, Benoit

2013-01-01

One of the hypothetical accident studied in the field of the safety studies of Pressurized light Water Reactor (PWR) is the Loss-Of-Coolant-Accident (LOCA). In this scenario, zirconium alloy fuel claddings could undergo an important oxidation at high temperature (T≅ 1200 C) in a steam environment. Cladding tubes constitute the first confinement barrier of radioelements and then it is essential that they keep a certain level of ductility after quenching to ensure their integrity. These properties are directly related to the growth kinetics of both the oxide and the αZr(O) phase and also to the oxygen diffusion profile in the cladding tube after the transient. In this context, this work was dedicated to the understanding and the modeling of the both oxidation phenomenon and oxygen diffusion in zirconium based alloys at high temperature. The numerical tool (EKINOX-Zr) used in this thesis is based on a numerical resolution of a diffusion/reaction problem with equilibrium-conditions on three moving boundaries: gas/oxide, oxide/αZr(O), αZr(O)/βZr. EKINOX-Zr kinetics model is coupled with ThermoCalc software and the Zircobase database to take into account the influence of the alloying elements (Sn, Fe, Cr, Nb) but also the influence of hydrogen on the solubility of oxygen. This study focused on two parts of the LOCA scenario: the influence of a pre-oxide layer (formed in-service) and the effects of hydrogen. Thanks to the link between EKINOX-Zr and the thermodynamic database Zircobase, the hydrogen effects on oxygen solubility limit could be considered in the numerical simulations. Thus, simulations could reproduce the oxygen diffusion profiles measured in pre-hydrided samples. The existence of a thick pre-oxide layer on cladding tubes can induce a reduction of this pre-oxide layer before the growth of a high-temperature one during the high temperature dwell under steam. The first simulations performed using the numerical tool EKINOX-Zr showed that this particular

SCC of Alloy 600 components in PWR primary loop

International Nuclear Information System (INIS)

Gomez-Briceno, Dolores; Lapena, Jesus; Castano, M. Luisa; Blazquez, Fernando

2002-01-01

initiation time has been determined. A detailed fractographic study of the fracture surface points out that the appearance of the fracture, intergranular in all the cases, is related to the susceptibility of the material. For the crack growth rate test, CT specimens tested under constant load were used. Specimens were fabricated from five Alloy 600 heats (two forged bars, cold work and hot work tubes, and a plate) with yield strength ranging from 280 to 413 MPa. Crack growth rate data were obtained at temperatures between 290 and 330 deg. C. Activation energy for both processes, crack initiation and propagation has been determined. On the other hand, in January 1994, during a refueling outage, an ID axial throughwall crack was detected in one of the RVH nozzle of Jose Cabrera Nuclear Plant in Spain. Extensive NDE examination of all the vessel head penetrations confirmed ID axially oriented indications in several of the nozzles. The cause of the extensive cracking detected was identified as an IGA/SCC process in primary water contaminated with sulphur species due to a cation resin ingress in the primary loop during the early 1980s. In order to confirm the postulated degradation process and to assess its relevance for other alloy 600 components in the reactor primary loop, an experimental program was performed. The scope of this program included to study the behaviour of sensitised alloy 600 in the water conditions postulated as the cause of the cracking and to obtain crack growth rate data in similar conditions, at 285 and 325 deg. C. In addition, the behaviour of the sensitised alloy 600 in shutdown conditions was also studied. In this paper the main results of these experimental programs, including no published data, will be presented and discussed in the light of the available results from other laboratories. (author)

Titanium condenser tubes. Problems and their solution for wider application to large surface condensers. [PWR

Energy Technology Data Exchange (ETDEWEB)

Sato, S; Sugiyama, S; Nagata, K; Nanba, K; Shimono, M [Sumitomo Light Metal Industries Ltd., Tokyo (Japan)

1977-06-01

The corrosion resistance of titanium in sea water is extremely excellent, but titanium tubes are expensive, and the copper alloy tubes resistant in polluted sea water were developed, therefore they were not used practically. In 1970, ammonia attack was found on the copper alloy tubes in the air-cooled portion of condensers, and titanium tubes have been used as the countermeasure. As the result of the use, the galvanic attack on copper alloy tube plates with titanium tubes as cathode and the hydrogen absorption at titanium tube ends owing to excess electrolytic protection was observed, but the corrosion resistance of titanium tubes was perfect. These problems can be controlled by the application of proper electrolytic protection. The condensers with all titanium tubes adopted recently in USA are intended to realize perfectly no-leak condensers as the countermeasure to the corrosion in steam generators of PWR plants. Regarding large condensers of nowadays, three problems are pointed out, namely the vibration of condenser tubes, the method of joining tubes and tube plates, and the tubes of no coolant leak. These three problems in case of titanium tubes were studied, and the problem of the fouling of tubes was also examined. The intervals of supporting plates for titanium tubes should be narrowed. The joining of titanium tubes and titanium tube plates by welding is feasible and promising. The cleaning with sponge balls is effective to control fouling.

Titanium condenser tubes--problems and their solutions for wider application to large surface condensers

Energy Technology Data Exchange (ETDEWEB)

Sato, S; Sugiyama, Y; Nagata, K; Namba, K; Shimono, M

1978-01-01

To meet the demand for high reliability condensers for thermal and nuclear power plants, especially for PWR plants, the condensers installed entirely with titanium tubes have been investigated and used. Some difficulties from conventional copper alloy tubes exist. Further investigations are necessary on three items: (1) tube vibration; (2) joining tubes to tube plate; (3) fouling (bio-fouling) control. Literature survey on the tube vibration suggests that the probability of tube vibration due to decreased stiffness of titanium tubes in comparison with conventional copper alloy tubes can be decreased by designing the proper span length between supports. Experiments on seal welding of tubes to a tube plate have successfully proved that pulsed TIG arc welding is applicable to get reliable and strong joints, even on site, by suitable countermeasures. Experiments on the fouling (bio-fouling) of titanium tubes in marine application reveal that the increased fouling of titanium tubes could be controlled by proper application of sponge ball cleaning.

A Eutectic Melting Study of Double Wall Cladding Tubes of FeCrAl and Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Lim, Woojin; Son, Seongmin; Lee, You Ho; Lee, Jeong Ik; Ryu, Ho Jin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jeong, Eun [Kyunghee University, Yongin (Korea, Republic of)

2015-10-15

The eutectic melting behavior of FeCrAl/Zircaloy-4 double wall cladding tubes was investigated by annealing at various temperatures ranging from 900 .deg. C to 1300 .deg. C. It was found that significant eutectic melting occurred after annealing at temperatures equal to or higher than 1150 .deg. C. It means that an additional diffusion barrier layer is necessary to limit the eutectic melting between FeCrAl and Zircaloy-4 alloy cladding tubes. Coating of FeCrAl layers on the Zr alloy cladding tube is being investigated for the development of accident tolerant fuel by exploiting of both the oxidation resistance of FeCrAl alloys and the neutronic advantages of Zr alloys. Coating of FeCrAl alloys on Zr alloy cladding tubes can be performed by various techniques including thermal spray, laser cladding, and co-extrusion. Son et al. also reported the fabrication of FeCrAl/Zr ally double wall cladding by the shrink fit method. For the double layered cladding tubes, the thermal expansion mismatch between the dissimilar materials, severe deformation or mechanical failure due to the evolution of thermal stresses can occur when there is a thermal cycling. In addition to the thermal stress problems, chemical compatibilities between the two different alloys should be investigated in order to check the stability and thermal margin of the double wall cladding at a high temperature. Generally, it is considered that Zr alloy cladding will maintain its mechanical integrity up to 1204 .deg. C (2200 .deg. F) to satisfy the acceptance criteria for emergency core cooling systems.

Lower head failure analysis

International Nuclear Information System (INIS)

Rempe, J.L.; Thinnes, G.L.; Allison, C.M.; Cronenberg, A.W.

1991-01-01

The US Nuclear Regulatory Commission is sponsoring a lower vessel head research program to investigate plausible modes of reactor vessel failure in order to determine (a) which modes have the greatest likelihood of occurrence during a severe accident and (b) the range of core debris and accident conditions that lead to these failures. This paper presents the methodology and preliminary results of an investigation of reactor designs and thermodynamic conditions using analytic closed-form approximations to assess the important governing parameters in non-dimensional form. Preliminary results illustrate the importance of vessel and tube geometrical parameters, material properties, and external boundary conditions on predicting vessel failure. Thermal analyses indicate that steady-state temperature distributions will occur in the vessel within several hours, although the exact time is dependent upon vessel thickness. In-vessel tube failure is governed by the tube-to-debris mass ratio within the lower head, where most penetrations are predicted to fail if surrounded by molten debris. Melt penetration distance is dependent upon the effective flow diameter of the tube. Molten debris is predicted to penetrate through tubes with a larger effective flow diameter, such as a boiling water reactor (BWR) drain nozzle. Ex-vessel tube failure for depressurized reactor vessels is predicted to be more likely for a BWR drain nozzle penetration because of its larger effective diameter. At high pressures (between ∼0.1 MPa and ∼12 MPa) ex-vessel tube rupture becomes a dominant failure mechanism, although tube ejection dominates control rod guide tube failure at lower temperatures. However, tube ejection and tube rupture predictions are sensitive to the vessel and tube radial gap size and material coefficients of thermal expansion

Complete Status Report Documenting Development of Friction Stir Welding for Joining Thin Wall Tubing of ODS Alloys

Energy Technology Data Exchange (ETDEWEB)

Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bunn, Jeffrey R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gussev, Maxim N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-08-01

The development of friction stir welding (FSW) for joining thin sections of the advanced oxide dispersion strengthened (ODS) 14YWT ferritic alloy was initiated in Fuel Cycle Research and Development (FCRD), now the Nuclear Technology Research and Development (NTRD), in 2015. The first FSW experiment was conducted in late FY15 and successfully produced a bead-on-plate stir zone (SZ) on a 1 mm thick plate of 14YWT (SM13 heat). The goal of this research task is to ultimately demonstrate that FSW is a feasible method for joining thin wall (0.5 mm thick) tubing of 14YWT.

Device and method for shortening reactor process tubes

Science.gov (United States)

Frantz, Charles E.; Alexander, William K.; Lander, Walter E. B.

1980-01-01

This disclosure describes a device and method for in situ shortening of nuclear reactor zirconium alloy process tubes which have grown as a result of radiation exposure. An upsetting technique is utilized which involves inductively heating a short band of a process tube with simultaneous application of an axial load sufficient to cause upsetting with an attendant decrease in length of the process tube.

Inhomogeneous deformation of aluminum alloy AA-2219 Rivet heads during manufacturing-A case study

International Nuclear Information System (INIS)

Ahmad, S.; Subhani, T.; Haq, A.U.

2006-01-01

AA -2219 aluminum alloy wire of diameter 4mm was used for the manufacturing of rivets used in aerospace industry. The problem arose as the heads of many of the rivets were inhomogenously deformed during manufacturing and were declared reject by the customer. Investigation was carried out to uncover the cause of the problem and to find its solution. Chemical analysis and mechanical testing were performed to ascertain the chemical composition and mechanical properties of the rivet wire. Metallographic examination was also carried out to study the macro and micro structural details. Metallographic and mechanical results were also compared with that of qualified rivets. It was revealed that the rivet wire was not annealed after wire drawing process- and was in hardened state, which gave rise to a non-uniform plastic flow during manufacturing. The presence of second phase coherent precipitates increased the hardness and tensile strength of rivet wire and was found to be the possible cause of inhomogenously deformed rivet heads; which was rectified by partial annealing the rivet wire at suitable temperature. (author)

Experimental observations on uniaxial whole-life transformation ratchetting and low-cycle stress fatigue of super-elastic NiTi shape memory alloy micro-tubes

Science.gov (United States)

Song, Di; Kang, Guozheng; Kan, Qianhua; Yu, Chao; Zhang, Chuanzeng

2015-07-01

In this work, the low-cycle fatigue failure of super-elastic NiTi shape memory alloy micro-tubes with a wall thickness of 150 μm is investigated by uniaxial stress-controlled cyclic tests at human body temperature 310 K. The effects of mean stress, peak stress, and stress amplitude on the uniaxial whole-life transformation ratchetting and fatigue failure of the NiTi alloy are observed. It is concluded that the fatigue life depends significantly on the stress levels, and the extent of martensite transformation and its reverse play an important role in determining the fatigue life. High peak stress or complete martensite transformation shortens the fatigue life.

Experimental observations on uniaxial whole-life transformation ratchetting and low-cycle stress fatigue of super-elastic NiTi shape memory alloy micro-tubes

International Nuclear Information System (INIS)

Song, Di; Kang, Guozheng; Kan, Qianhua; Yu, Chao; Zhang, Chuanzeng

2015-01-01

In this work, the low-cycle fatigue failure of super-elastic NiTi shape memory alloy micro-tubes with a wall thickness of 150 μm is investigated by uniaxial stress-controlled cyclic tests at human body temperature 310 K. The effects of mean stress, peak stress, and stress amplitude on the uniaxial whole-life transformation ratchetting and fatigue failure of the NiTi alloy are observed. It is concluded that the fatigue life depends significantly on the stress levels, and the extent of martensite transformation and its reverse play an important role in determining the fatigue life. High peak stress or complete martensite transformation shortens the fatigue life. (paper)

Flow forming of Al-Zn alloys 7075

International Nuclear Information System (INIS)

Abbas, G.

1997-01-01

Feasibility of flow forming aluminium alloy 7075 for manufacturing the tubes of about 70 mm diameter and different lengths has been studied. The elongation of the material was increased by increasing the working temperature of the process. Effect of various process parameters like roller rpm, roller offset, roller feed, spindle speed etc. on the flow ability of ally 7075 and the hardness and surface furnish of the tube has been determined. It was found that the roller bite and roller speed, of the flow forming machine, were two most important process parameters, which affected the hardness and the surface finish of the tube. By establishing the optimum process parameters, it was possible to produce aluminium alloy 7075 tubes of lengths ranging from 500 cm to 1 m with 70 mm outer diameter. The maximum hardness achieved on the surface of the tube was 146 Hv with good surface finish. (author)

Effect of Elemental Sulfur and Sulfide on the Corrosion Behavior of Cr-Mo Low Alloy Steel for Tubing and Tubular Components in Oil and Gas Industry.

Science.gov (United States)

Khaksar, Ladan; Shirokoff, John

2017-04-20

The chemical degradation of alloy components in sulfur-containing environments is a major concern in oil and gas production. This paper discusses the effect of elemental sulfur and its simplest anion, sulfide, on the corrosion of Cr-Mo alloy steel at pH 2 and 5 during 10, 20 and 30 h immersion in two different solutions. 4130 Cr-Mo alloy steel is widely used as tubing and tubular components in sour services. According to the previous research in aqueous conditions, contact of solid sulfur with alloy steel can initiate catastrophic corrosion problems. The corrosion behavior was monitored by the potentiodynamic polarization technique during the experiments. Energy dispersive X-ray spectroscopy (EDS) and scanning electron microscopy (SEM) have been applied to characterize the corrosion product layers after each experiment. The results show that under the same experimental conditions, the corrosion resistance of Cr-Mo alloy in the presence of elemental sulfur is significantly lower than its resistance in the presence of sulfide ions.

Design and Computational Fluid Dynamics Optimization of the Tube End Effector for Reactor Pressure Vessel Head Type VVER-1000

International Nuclear Information System (INIS)

Novosel, D.

2006-01-01

In this paper is presented development and optimization of the tube end effector design which should consist of 4 ultrasonic transducers, 4 Eddy Current's transducers and Radiation Proof Dot Camera. Basically, designing was conducted by main input requests, such as: inner diameter of a tested reactor pressure vessel head penetration tube, dimensions of a transducers and maximum allowable vertical movement of a manipulator connection rod in order to cover all inner tube surface. As is obvious, for ultrasonic testing should be provided the thin layer of liquid material (in our case water was chosen) which is necessary to make physical contact between transducer surface and investigated inner tube surface. By help of Computational Fluid Dynamics, determined were parameters of geometry, as the most important factor of transducer housing, hydraulically parameters for water supply and primary drain together implemented into this housing, movement of the end effectors (vertical and cylindrical) and finally, necessary equipment which has to provide all hydraulically and pneumatic requirements. As the cylindrical surface of the inner tube diameter was liquefied and contact between transducer housing and tested tube wasn't ideally covered, water leakage could occur in downstream direction. To reduce water leakage, which is highly contaminated, developed was second water drain by diffuser assembly which is driven by Venturi pipe, commercially called vacuum generator. Using the Computational Fluid Dynamic, obtained was optimized geometry of diffuser control volume with the highest efficiency, in other words, unobstructed fluid flux. Afterwards, the end effectors system was synchronized to the existing operable system for NDT methods all invented and designed by INETEC. (author)

Development of a multivariable normal tissue complication probability (NTCP) model for tube feeding dependence after curative radiotherapy/chemo-radiotherapy in head and neck cancer

International Nuclear Information System (INIS)

Wopken, Kim; Bijl, Hendrik P.; Schaaf, Arjen van der; Laan, Hans Paul van der; Chouvalova, Olga; Steenbakkers, Roel J.H.M.; Doornaert, Patricia; Slotman, Ben J.; Oosting, Sjoukje F.; Christianen, Miranda E.M.C.; Laan, Bernard F.A.M. van der; Roodenburg, Jan L.N.; René Leemans, C.; Verdonck-de Leeuw, Irma M.; Langendijk, Johannes A.

2014-01-01

Background and purpose: Curative radiotherapy/chemo-radiotherapy for head and neck cancer (HNC) may result in severe acute and late side effects, including tube feeding dependence. The purpose of this prospective cohort study was to develop a multivariable normal tissue complication probability (NTCP) model for tube feeding dependence 6 months (TUBE M6 ) after definitive radiotherapy, radiotherapy plus cetuximab or concurrent chemoradiation based on pre-treatment and treatment characteristics. Materials and methods: The study included 355 patients with HNC. TUBE M6 was scored prospectively in a standard follow-up program. To design the prediction model, the penalized learning method LASSO was used, with TUBE M6 as the endpoint. Results: The prevalence of TUBE M6 was 10.7%. The multivariable model with the best performance consisted of the variables: advanced T-stage, moderate to severe weight loss at baseline, accelerated radiotherapy, chemoradiation, radiotherapy plus cetuximab, the mean dose to the superior and inferior pharyngeal constrictor muscle, to the contralateral parotid gland and to the cricopharyngeal muscle. Conclusions: We developed a multivariable NTCP model for TUBE M6 to identify patients at risk for tube feeding dependence. The dosimetric variables can be used to optimize radiotherapy treatment planning aiming at prevention of tube feeding dependence and to estimate the benefit of new radiation technologies

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)

Darlington NGD fuel handling head eight acceptance program

International Nuclear Information System (INIS)

Skelton, P.H.; Sie, T.

1996-01-01

Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier. During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site. Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels. An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site. This paper focuses on the results of the head/screw test program. Results of the acceptance testing are also discussed. (author). 2 refs., 4 figs

Darlington NGD fuel handling head eight acceptance program

Energy Technology Data Exchange (ETDEWEB)

Skelton, P H; Sie, T [Ontario Hydro, Bowmanville (Canada). Darlington Nuclear Generating Station; Pilgrim, J [Canadian General Electric Co. Ltd., Toronto, ON (Canada)

1997-12-31

Darlington NGD requires eight fuelling machine heads to fuel the four 932 MW reactors. Six heads are used on the three fuelling machine trolleys for normal fuelling operations. A further two heads are required to allow for maintenance and to provide for such reactor face activities as PIPE and CIGAR. Seven heads were successfully delivered to site from the head supplier. During acceptance testing, stalls on the charge tube screw assembly of the eighth and final head prevented its delivery to site. Replacement of the charge tube screw with a spare screw did not alleviate the problem. An in depth series of tests were undertaken at site, at the supplier and at the screw sub-supplier to determine the root cause of the problem. These tests included taking torque measurements under different operating conditions and using different components to assess the effects of the changes on torque levels. An assessment of the effects of changing chemical conditions (particularly crud levels) was also made. To ensure that the results of the testing were well understood, additional torque testing was also completed on a head and screw assembly at site that was known to work well. Based on all of the above series of tests, a recommendation was made to re-machine the charge tube screw(s). The original charge tube screw from Head eight was subsequently returned to the sub-supplier for re-work. Follow-up torque measurements and acceptance testing showed that the screw rework was effective and that Head eight could be successfully delivered to site. This paper focuses on the results of the head/screw test program. Results of the acceptance testing are also discussed. (author). 2 refs., 4 figs.

Vessel head penetrations: French approach for maintenance in the PLIM program

International Nuclear Information System (INIS)

Champigny, F.

2002-01-01

Full text: In 1991, in the Bugey nuclear power plant, for the first time a leak occurred at the level of a vessel head penetration made with base nickel alloy (Inconel 600). This leak was caused by a primary stress corrosion cracking coming from inside the penetration tube. The crack was trough wall extent and primary fluid went out from the top of the vessel head. Immediately, Electricite de France launched important research programs and expertise in order to understand the root causes and propose solutions to this problem. The root causes confirmed PWSCC, and in the same time solutions for repair were studied and an inspection program was established to check the base metal of other vessel head penetrations. After several tests, repair solutions were abandoned because of their high costs (financial and dosimetry). EDF decided to replace all the vessel heads with Inconel 600 penetrations. Non destructive developments leaded to use eddy currents for detection and characterization but also televisual techniques to confirm. In a second step, in order to inspect without removing the inside thermal sleeve, eddy current and ultrasonic sword probes were achieved and used to inspect all vessel heads penetrations. Up to now, 75% of the vessel head have been replaced on the 900 MW and 1300 MW fleets but to replace wisely the last vessel heads EDF continues to perform NDE of the penetrations on the basis of safety criteria. This paper describes the different steps of the applied policy in France, NDE methods, criteria and the results obtained. (author)

Alloy chemistry and microstructural control to meet the demands of the automotive Stirling engine

Science.gov (United States)

Stephens, J. R.

1986-01-01

The automotive Stirling engine now under development by DOE/NASA as an alternative to the internal combustion engine, imposes severe materials requirements for the hot portion of the engine. Materials selected must be low cost and contain a minimum of strategic elements so that availability is not a problem. Heater head tubes contain high pressure hydrogen on the inside and are exposed to hot combustion gases on the outside surface. The cylinders and regenerator housings must be readily castable into complex shapes having varying wall thicknesses and be amenable to brazing and welding operations. Also, high strength, oxidation resistance, resistance to hydrogen permeation, cyclic operation, and long-life are required. A research program conducted by NASA Lewis focused on alloy chemistry and microstructural control to achieve the desired properties over the life of the engine. Results of alloy selection, characterization, evaluation, and actual engine testing of selected materials are presented.

Delayed hydrogen cracking of zirconium alloy pressure tubes

International Nuclear Information System (INIS)

Jackman, A.H.; Dunn, J.T.

1976-10-01

After several years of almost continuous service, Pickering Units 3 and 4 have both experienced long outages to replace cracked pressure tubes. This report summarizes the status of the investigation into the cause of the cracks as of May 1976. The basic cause of the cracking was the presence of very high residual tensile stresses in the pressure tubes due to improper rolling procedures. These residual stresses are being reduced to acceptable levels by local stress relieving techniques at Bruce G.S. and in future reactors improvements in rolling procedures and changes in pressure tube specifications will prevent a recurrence of this problem. (author)

Kinked and retained nasogastric tube in polytrauma patient; a case report

Directory of Open Access Journals (Sweden)

Kumar Ashok

2017-06-01

Full Text Available Enteral feeding is an important and preferred technique of feeding in head injury patient to provide nutrition. As inadequate nutrition causes decrease in physical ability, neurological impairment and takes a long time for improvement or delayed deterioratation. With our best knowledge kinked and retained nasogastric tube in stomach is a very rare complication of feeding in head injuries patients. Predisposing factors that can cause kinking is excess tube length, tube in situ for long time and small bore tube. We are reporting one such case of kinked and retained nasogastric tube in the stomach of a polytrauma patient which was retrieved by upper GI endoscope.

Development and validation of a prediction model for tube feeding dependence after curative (chemo- radiation in head and neck cancer.

Directory of Open Access Journals (Sweden)

Kim Wopken

Full Text Available BACKGROUND: Curative radiotherapy or chemoradiation for head and neck cancer (HNC may result in severe acute and late side effects, including tube feeding dependence. The purpose of this prospective cohort study was to develop a prediction model for tube feeding dependence 6 months (TUBEM6 after curative (chemo- radiotherapy in HNC patients. PATIENTS AND METHODS: Tube feeding dependence was scored prospectively. To develop the multivariable model, a group LASSO analysis was carried out, with TUBEM6 as the primary endpoint (n = 427. The model was then validated in a test cohort (n = 183. The training cohort was divided into three groups based on the risk of TUBEM6 to test whether the model could be extrapolated to later time points (12, 18 and 24 months. RESULTS: Most important predictors for TUBEM6 were weight loss prior to treatment, advanced T-stage, positive N-stage, bilateral neck irradiation, accelerated radiotherapy and chemoradiation. Model performance was good, with an Area under the Curve of 0.86 in the training cohort and 0.82 in the test cohort. The TUBEM6-based risk groups were significantly associated with tube feeding dependence at later time points (p<0.001. CONCLUSION: We established an externally validated predictive model for tube feeding dependence after curative radiotherapy or chemoradiation, which can be used to predict TUBEM6.

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

Probabilistic analysis of degradation incubation time of steam generator tubing materials

International Nuclear Information System (INIS)

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

2012-01-01

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

Inconel alloys development -Development of the advanced nuclear materials-

International Nuclear Information System (INIS)

Kuk, Il Hiun; Jang, Jin Sung; Rhee, Chang Kyu; Chung, Man Kyo; Woo, Yun Myeoung; Han, Chang Hee

1994-07-01

We surveyed the current status and problems in S/G U-tubes in Korea and worldwide. Also we gathered manufacturing specifications of S/G U-tubes and compared/analyzed the differences in them company by company. We produced alloy 600 tubes (in cooperation with Sammi Special Steels) through V.I.M. (Vacuum Induction Melting; 2 ton capacity), 4 steps of hot press forging, hot extrusion (10:1 of reduction ratio), 3 steps of cold pilgerings and so on. We will continue to characterize the tubes and 2nd time preproduce the tubes using the feed-back data. With regard to alloy 690, which is getting popular for S/G U-tubes worldwide, we cast four 60 Kg ingots and two 6 Kg ingots by V.I.M.. We analyzed the chemical composition, macrostructures, hot workability, and so on ; all ingots were good except on 60 Kg ingot. Finally we produced high quality alloy 690 ingot (about 1 Kg) by E.S.R. (Electroslag Remelting) method (in cooperation with Yeoungnam University). We used CaF/CaO/Al2O3/MgO quartenary slag system. We have made directionally grown good ingots by E.S.R. and especially the hot workability at 1100 deg C - the temperature at which V.I.M. ingots showed very poor hot workability - was very much improved (from 30 to 90 % of reduction of area). We continue to analyze the effects of E.S.R. to the structure and properties of alloy 690 (grain size, morphology, and directionality; any changes of inclusions and so on). (Author)

In Vivo Damage of the Head-Neck Junction in Hard-on-Hard Total Hip Replacements: Effect of Femoral Head Size, Metal Combination, and 12/14 Taper Design

Directory of Open Access Journals (Sweden)

Massimiliano Baleani

2017-07-01

Full Text Available Recently, concerns have been raised about the potential effect of head-neck junction damage products at the local and systemic levels. Factors that may affect this damage process have not been fully established yet. This study investigated the possible correlations among head-neck junction damage level, implant design, material combination, and patient characteristics. Head-neck junctions of 148 retrieved implants were analysed, including both ceramic-on-ceramic (N = 61 and metal-on-metal (N = 87 bearings. In all cases, the male taper was made of titanium alloy. Damage was evaluated using a four-point scoring system based on damage morphology and extension. Patient age at implantation, implantation time, damage risk factor, and serum ion concentration were considered as independent potential predicting variables. The damage risk factor summarises head-neck design characteristics and junction loading condition. Junction damage correlated with both implantation time and damage factor risk when the head was made of ceramic. A poor correlation was found when the head was made of cobalt alloy. The fretting-corrosion phenomenon seemed mainly mechanically regulated, at least when cobalt alloy components were not involved. When a component was made of cobalt alloy, the role of chemical phenomena increased, likely becoming, over implantation time, the damage driving phenomena of highly stressed junctions.

Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

International Nuclear Information System (INIS)

Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

1977-01-01

The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

Delayed Hydride Cracking Mechanism in Zirconium Alloys and Technical Requirements for In-Service Evaluation of Zr-2.5Nb Tubes with Flaws

International Nuclear Information System (INIS)

Kim, Young Suk

2007-01-01

In association with periodic inspection of CANDU nuclear power plant components, Canadian Standards Association issued CSA N285.8 in 2005 as technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors. This first version, CSA N285.8 involves procedures for, firstly, the evaluation of pressure tube flaws, secondly, the evaluation of pressure tube to calandria tube contact and, thirdly, the assessment of a reactor core, and material properties and derived quantities. The evaluation of pressure tube flaws includes delayed hydride cracking evaluation the procedures of which are stipulated based on the existing delayed hydride cracking models. For example, the evaluation of flaw-tip hydride precipitation during reactor cooldown involves a procedure to calculate the equilibrium hydrogen equivalent concentration in solution at the flaw tip, Htipas follows: Htip=Hfexp[- (VH delta no.)/RT], where Hf is the total bulk hydrogen equivalent concentration, VH partial molar volume of hydrogen in zirconium, δ a difference in hydrostatic stress between the bulk and the crack tip. When Htip ≥TSSP at temperature, then flaw-tip hydride is predicted to precipitate. Eq. (1) suggests that hydrogen concentration at the crack tip would increase due to an work energy given by the difference in the hydrostatic stress

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Evaluation on mechanical and corrosion properties of steam generator tubing materials

International Nuclear Information System (INIS)

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

1998-06-01

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

Evaluation on mechanical and corrosion properties of steam generator tubing materials

Energy Technology Data Exchange (ETDEWEB)

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

1998-06-15

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Expanded heat treatment to form residual compressive hoop stress on inner surface of zirconium alloy tubing

International Nuclear Information System (INIS)

Megata, Masao

1997-01-01

A specific heat treatment process that introduces hoop stress has been developed. This technique can produce zirconium alloy tubing with a residual compressive hoop stress near the inner surface by taking advantage of the mechanical anisotropy in hexagonal close-packed zirconium crystal. Since a crystal having its basal pole parallel to the tangential direction of the tubing is easier to exhibit plastic elongation under the hoop stress than that having its basal pole parallel to the radial direction, the plastic and elastic elongation can coexist under a certain set of temperature and hoop stress conditions. The mechanical anisotropy plays a role to extend the coexistent stress range. Thus, residual compressive hoop stress is formed at the inner surface where more plastic elongation occurs during the heat treatment. This process is referred to as expanded heat treatment. Since this is a fundamental crystallographic principle, it has various applications. The application to improve PCI/SCC (pellet cladding interaction/stress corrosion cracking) properties of water reactor fuel cladding is promising. Excellent results were obtained with laboratory-scale heat treatment and an out-reactor iodine SCC test. These results included an extension of the time to SCC failure. (author)

Changes in the design, fabrication and setting of guide tube support pins in alloy X750

International Nuclear Information System (INIS)

Benhamou, C.; Chambrin, J.L.; Todeschini, P.; Champredonde, J.; Lemaire, E.

2004-01-01

As a consequence of a problem of stress corrosion cracking (SCC) encountered on guide tube support pins (GTSP) of first generation (1982) and of second generation (1987), EDF and Framatome decided in mars 1988 to launch an important program involving a complete overhaul of the design, the material used, the fabrication and the setting in reactor of GTSP. This program has led to the implementation in 900 MWe and 1300 MWe PWR of a new tube guide support pin called NG89. This implementation began in 1989, now 15 years later, 40% of the operating GTSP in 900 MWe and 1300 MWe PWR are of NG89 type, the oldest ones cumulate 105000 hours in service without negative feedback experience. The main features of the NG89 is: - to be made from an alloy X-750 containing boron (from 25 to 45 ppm) - to have a SCC threshold set at 720 MPa - to be machined from metal bars completely treated, - to have a rolling of the fillets, and - to undergo a shot blasting on the zones of the surface the most acted upon. (A.C.)

Effect of heat treatment and composition on stress corrosion cracking of steam generation tubing materials

International Nuclear Information System (INIS)

Kim, H. P.; Hwang, S. S.; Kuk, I. H.; Kim, J. S.; Oh, C. Y.

1998-01-01

Effects of heat treatment and alloy composition on stress corrosion cracking (SCC) of steam generator tubing materials have been studied in 40% NaOH at 315.deg.C at potential of +200mV above corrosion potential using C-ring specimen and reverse U bend specimen. The tubing materials used were commercial Alloy 600, Alloy 690 and laboratory alloys, Ni-χCr-10Fe. Commercial Alloy 600, Alloy 690 were mill annealed or thermally treated.Laboratory alloy Ni-χCr-10Fe, and some of Alloy 600 and Alloy 690 were solution annealed. Polarization curves were measured to find out any relationship between SCC susceptibility and electrochemical behaviour. The variation in thermal treatment of Alloy 600 and Alloy 690 had no effect on polarization behaviour probably due to small area fraction of carbide and Cr depletion zone near grain boundary. In anodic polarization curves, the first and second anodic peaks at about 170mV and about at 260mV, respectively, above corrosion potential were independent of Cr content, whereas the third peak at 750mV above corrosion potential and passive current density in-creased with Cr content. SCC susceptibility decreased with Cr content and thermal treatment producing semicontinuous grain boundary decoration. Examination of cross sectional area of C-ring specimen showed deep SCC cracks for the alloys with less than 17%Cr and many shallow attacks for alloy 690. The role of Cr content in steam generator tubing materials and grain boundary carbide on SCC were discussed

Fireside corrosion testing of candidate superheater tube alloys, coatings, and claddings -- Phase 2

Energy Technology Data Exchange (ETDEWEB)

Blough, J.L.; Seitz, W.W. [Foster Wheeler Development Corp., Livingston, NJ (United States)

1997-12-01

In Phase 1 a variety of developmental and commercial tubing alloys and claddings were exposed to laboratory fireside corrosion testing simulating a superheater or reheater in a coal-fired boiler. Phase 2 (in situ testing) has exposed samples of 347 RA-85H, HR3C, 253MA, Fe{sub 3}Al + 5Cr, 310 Ta modified, NF 709, 690 clad, and 671 clad for approximately 4,000, 12,000, and 16,000 hours to the actual operating conditions of a 250-MW coal-fired boiler. The samples were assembled on an air-cooled, retractable corrosion probe, the probe was installed in the reheater activity of the boiler and controlled to the operating metal temperatures of an existing and advanced-cycle coal-fired boiler. The results will be presented for the preliminary metallurgical examination of the corrosion probe samples after 16,000 hours of exposure. Continued metallurgical and interpretive analysis is still on going.

Preparation of metallic uranium tubes; Elaboration des tubes d'uranium metallique

Energy Technology Data Exchange (ETDEWEB)

Lerouge, G; Decours, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The production furnace is an induction heated vacuum furnace having a capacity at the moment of 250 kg. Previously the crucible was heated by the inductor, the mould being outside the inductor. The tubes thus produced contained cavities, the alloy structure was fine; this was cold-mould casting, At the moment the top of the moulds are pre-heated, this is the so called hot-mould casting. This method has the advantage of eliminating the cavities but leads to a less fine microstructure. The alloy used for the 18 x 40 mm and 23 x 43 mm tubes is U-Mo (1.1 per cent). Since the moulds are now heated at the top, the solidification of the metal is very slow in this zone leading to a pronounced {gamma} grain, whereas towards the base the faster cooling leads to a smaller {gamma} grain. The {gamma} structure depends essentially on the solidification rate and on the time spent in this zone. In order to obtain a fine and homogeneous grain along the whole length of the tube, a controlled cooling treatment is effected. It consists in heating the uranium tubes in the {gamma} place and then in cooling them at a rate of between 20 and 50 deg C/mm down to 400 deg C. The 77 x 95 mm and 54 x 70 mm annular elements are at the moment being produced for research purposes. Their preparation is similar to that of 18 x 40 mm and 23 x 43 mm elements. The 77 x 95 mm tubes are at the moment made from U-Cr alloy (0.1 per cent); because of their size, their preparation is carried out in 600 mm diameter furnaces. (authors) [French] Le four d'elaboration est un four sous vide chaufffe par induction, dont la capacite actuelle est de 250 kg. Anterieurement le creuset seul etait chauffe par l'inducteur, les moules etaient hors de l'inducteur. Les tubes obtenus presentaient des cavites, la structure de l'alliage etait fine, c'etait la coulee en moules froids. Actuellement on prechauffe le haut des moules, c'est la coulee dite en moules chauds. Cette facon de faire a l'avantage de supprimer les cavites

Photomultiplier tube artifacts on 67Ga-citrate imaging caused by loss of correction floods due to an off-peak status of one head of a dual-head γ-camera.

Science.gov (United States)

Glaser, Joseph E; Song, Na; Jaini, Sridivya; Lorenzo, Ruth; Love, Charito

2012-12-01

γ-cameras use flood-field corrections to ensure image uniformity during clinical imaging. A loss or corruption of the correction data of one head of a dual-head camera can result in an off-peak artifactual appearance. We present our experience with the occurrence of such an incident on a (67)Ga scan. A patient was referred for a whole-body (67)Ga scan to evaluate for causes of neutropenic fever. Whole-body planar and static images of the head, chest, abdomen, pelvis, and lower extremities in multiple projections were obtained. Whole-body images showed decreased image quality on the anterior view obtained with detector 1 and an unremarkable posterior image obtained with detector 2. A problem with detector 2 was suspected, and additional static images were obtained after rotation of the detector heads. The posterior images taken with detector 1 showed photomultiplier tube outlines. The anterior images taken with detector 2 showed improved count and image quality. It was later found that the uniformity map for detector 2 had been lost and that this software malfunction led to the resulting imaging problem. When artifacts with an off-peak appearance are seen on scintigraphic images, evaluation of possible causes should include not only isotope window settings but also an incorrect or corrupted uniformity map.

Automated tube voltage adaptation in head and neck computed tomography between 120 and 100 kV: effects on image quality and radiation dose

Energy Technology Data Exchange (ETDEWEB)

May, Matthias S.; Uder, Michael; Lell, Michael M. [University Hospital Erlangen, Department of Radiology, Erlangen (Germany); University Erlangen, Imaging Science Institute, Erlangen (Germany); Kramer, Manuel R.; Eller, Achim; Wuest, Wolfgang; Scharf, Michael; Brand, Michael; Saake, Marc [University Hospital Erlangen, Department of Radiology, Erlangen (Germany); Schmidt, Bernhard [Siemens Healthcare, Erlangen (Germany)

2014-09-15

Low tube voltage allows for computed tomography (CT) imaging with increased iodine contrast at reduced radiation dose. We sought to evaluate the image quality and potential dose reduction using a combination of attenuation based tube current modulation (TCM) and automated tube voltage adaptation (TVA) between 100 and 120 kV in CT of the head and neck. One hundred thirty consecutive patients with indication for head and neck CT were examined with a 128-slice system capable of TCM and TVA. Reference protocol was set at 120 kV. Tube voltage was reduced to 100 kV whenever proposed by automated analysis of the localizer. An additional small scan aligned to the jaw was performed at a fixed 120 kV setting. Image quality was assessed by two radiologists on a standardized Likert-scale and measurements of signal- (SNR) and contrast-to-noise ratio (CNR). Radiation dose was assessed as CTDI{sub vol}. Diagnostic image quality was excellent in both groups and did not differ significantly (p = 0.34). Image noise in the 100 kV data was increased and SNR decreased (17.8/9.6) in the jugular veins and the sternocleidomastoid muscle when compared to 120 kV (SNR 24.4/10.3), but not in fatty tissue and air. However, CNR did not differ statistically significant between 100 (23.5/14.4/9.4) and 120 kV data (24.2/15.3/8.6) while radiation dose was decreased by 7-8 %. TVA between 100 and 120 kV in combination with TCM led to a radiation dose reduction compared to TCM alone, while keeping CNR constant though maintaining diagnostic image quality. (orig.)

SCC testing of steam generator tubes repaired by welded sleeves

International Nuclear Information System (INIS)

Pierson, E.; Stubbe, J.

1993-01-01

One way to repair steam generator tubing is to introduce a sleeve inside the tube so that it spans the corroded area and to seal it at both ends. This technique has been studied at Laborelec with a particular attention paid to the occurrence of new SCC cracks at the upper joint. Tube segments coming from the same lot of mill annealed alloy 600 were sent to six manufacturers to be sleeved by their own procedure (including TIG, laser or kinetic welding, followed or not by a stress relief heat treatment), and then tested at Laborelecin 10% NaOH at 350 degrees C. The tests were performed with and without differential pressure i.e. in capsules (Δ = 9 and 19 MPa) and in autoclave (Δp = 0). Nearly all the not stress relieved mock-ups developed through cracks in several hundred hours in auto-clave. The cracks were circumferential and situated near the weld. At 9 and 19 MPa, the time to failure decreased and longitudinal cracks appeared near the weld and at the transition zone of expanded areas. Cracks were never observed in the alloy 690 sleeve, except in the weld bead. Reference capsules (roll expaned tubes) made of the same lot of alloy 600 were tested in the same environment

X-ray tube targets

International Nuclear Information System (INIS)

Hirsch, H.H.

1980-01-01

In rotary targets for X-ray tubes warping is a problem which causes X-ray deficiency. A rotary target is described in which warping is reduced by using alloys of molybdenum with 0.05 to 10% iron, silicon, cobalt, tantalum, niobium, hafnium, stable metal oxide or mixture thereof. Suitable mixtures are 0.5 to 10% of tantalum, niobium or hafnium with from 0.5 to 5% yttrium oxide, or 0.05 to 0.3% of cobalt or silicon. Optionally 0.1 to 5% by weight of additional material may be alloyed with the molybdenum, such as tantalum or hafnium carbides. (author)

Ferromagnetic material inspection for feedwater heater and condenser tubes

International Nuclear Information System (INIS)

Anon.

1987-01-01

In recent years, special ferritic stainless steels, such as AL29-4C/sup TM/, Sea-Cure/sup TM/, E-Brite/sup TM/, 439, and similar alloys have been introduced as tube material in condensers, feedwater heaters, moisture separator/reheaters, and other heat exchangers. In addition, carbon steel tubes are widely used in feedwater heaters and heat exchangers in chemical plants. The main problem with the in-service inspection of these ferritic alloys and carbon steel tubes lies in their highly ferromagnetic properties. These properties severely limit the application of the standard eddy current techniques. The effort was undertaken under EPRI sponsorship to develop a reliable technique for in-service inspection of ferromagnetic tubes. The new method combines the measurement of magnetic flux leakage generated around the defects with measurement of total flux in the tube wall. The heart of the inspection system is a special ID probe that magnetizes the tube and generates signals for any tube defect. A permanent record of inspection is provided with a strip-chart or magnetic tape recorder. The laboratory and field evaluation of this new system demonstrated its very good sensitivity to small defects, its reliability, and its ruggedness. Defects as small as 10% external wall loss in heavy wall carbon steel tube were detected. Tubes in the power plant were inspected at a rate of 300-500 tubes per eight-hour shift. The other advantages of this newly developed technique are its simplicity, low cost of instrumentation, easy data interpretation, and full portability

Analysis of hafnium in zirconium alloys

International Nuclear Information System (INIS)

Kondo, Isao; Sakai, Fumiaki; Ohuchi, Yoshifusa; Nakamura, Hisashi

1977-01-01

It is required to analyse alloying components and impurity elements in the acceptance analysis of zirconium alloys as the material for fuel cladding tubes and pressure tubes for advanced thermal reactors. Because of extreme similarity in chemical properties between zirconium and hafnium, about 100 ppm of hafnium is usually contained in zirconium alloys. Zircaloy-2 alloy and 2.5% Nb-zirconium with the addition of hafnium had been prepared as in-house standard samples for rapid analysis. Study was made on fluorescent X-ray analysis and emission spectral analysis to establish the analytical method. By using these in-house standard samples, acceptance analysis was successfully carried out for the fuel cladding tubes for advanced thermal reactors. Sulfuric acid solution was prepared from JAERI-Z 1, 2 and 3, the standard sample for zircaloy-2 prepared by the Analytical Committee on Nuclear Fuel and Reactor Materials, JAERI, and zirconium oxide (Hf 1 ppm/Zr). Standard Hf solution was added to the sulfuric acid solution step by step, to make up a series of the standard oxide samples by the precipitation process. By the use of these standard samples, the development of the analytical method and joint analysis were made by the three-member analytical technique research group including PNC. The analytical precision for the fluorescent X-ray analysis was improved by attaching a metallic yttrium filter to the window of an X-ray tube so as to suppress the effect due to zirconium matrix. The variation factor of the joint analysis was about 10% to show good agreement, and the indication value was determined. (Kobatake, H.)

Qualification of stainless steel for OTEC heat exchanger tubes

Energy Technology Data Exchange (ETDEWEB)

LaQue, F.L.

1979-01-01

The history of the AL-6X alloy is reviewed and its credentials as a candidate for use as tubing in Ocean Thermal Energy Conversion Heat Exchangers are examined. Qualification is based on results of accelerated tests using ferric chloride for resistance to crevice corrosion and pitting, long-time crevice corrosion and pitting tests in natural sea water and anticipated resistance to attack by ammonia and mixtures of ammonia and sea water. Since the alloy has no natural resistance to fouling by marine organisms, it must be able to accomodate action to prevent fouling by chlorination or to remove it by mechanical cleaning techniques or appropriate chemical cleaning methods. The satisfactory behavior indicated by the various accelerated and long-time corrosion tests has been confirmed by excellent performance of several million feet of tubing in condensers in coastal power plants. Early evaluation tests demonstrated the need for proper heat treatment to avoid the presence of a sigma phase, which promoted severe pitting of some, but not all, specimens in tests in natural sea water. The available data qualify the AL-6X alloy as being a satisfactory alternate to titanium for tubes in OTEC heat exchangers.

Preparation of metallic uranium tubes; Elaboration des tubes d'uranium metallique

Energy Technology Data Exchange (ETDEWEB)

Lerouge, G.; Decours, J. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1964-07-01

The production furnace is an induction heated vacuum furnace having a capacity at the moment of 250 kg. Previously the crucible was heated by the inductor, the mould being outside the inductor. The tubes thus produced contained cavities, the alloy structure was fine; this was cold-mould casting, At the moment the top of the moulds are pre-heated, this is the so called hot-mould casting. This method has the advantage of eliminating the cavities but leads to a less fine microstructure. The alloy used for the 18 x 40 mm and 23 x 43 mm tubes is U-Mo (1.1 per cent). Since the moulds are now heated at the top, the solidification of the metal is very slow in this zone leading to a pronounced {gamma} grain, whereas towards the base the faster cooling leads to a smaller {gamma} grain. The {gamma} structure depends essentially on the solidification rate and on the time spent in this zone. In order to obtain a fine and homogeneous grain along the whole length of the tube, a controlled cooling treatment is effected. It consists in heating the uranium tubes in the {gamma} place and then in cooling them at a rate of between 20 and 50 deg C/mm down to 400 deg C. The 77 x 95 mm and 54 x 70 mm annular elements are at the moment being produced for research purposes. Their preparation is similar to that of 18 x 40 mm and 23 x 43 mm elements. The 77 x 95 mm tubes are at the moment made from U-Cr alloy (0.1 per cent); because of their size, their preparation is carried out in 600 mm diameter furnaces. (authors) [French] Le four d'elaboration est un four sous vide chaufffe par induction, dont la capacite actuelle est de 250 kg. Anterieurement le creuset seul etait chauffe par l'inducteur, les moules etaient hors de l'inducteur. Les tubes obtenus presentaient des cavites, la structure de l'alliage etait fine, c'etait la coulee en moules froids. Actuellement on prechauffe le haut des moules, c'est la coulee dite en moules chauds. Cette facon de faire a l

Numerical and Experimental Analysis on the Cavity Formation and Shrinkage for Investment Cast Alloy 738 4 mm-Thick Rectangular Tube

International Nuclear Information System (INIS)

Park, Myeong-Il; Choi, Yoon Suk; Yoo, Jae-Hyun; Park, Sang-Hu; Kim, Kyeong-Min; Lee, Yeong-Chul; Lee, Jung-Seok; Lee, Jae-Hyun

2017-01-01

Investment casting for the thin (4 mm thick) rectangular tube (40 mm wide, 80 mm high and 200 mm long) was carried out numerically and experimentally for Alloy 738, which is a precipitation-hardened Ni-base superalloy. Two types of rectangular tubes, one with a regular array (10 mm by 10 mm square array) of protruded rods (3 mm in diameter and 3mm in height) embedded on the outer surface and the other with just smooth surface, were investment-cast at the same time through the side feeding mold design. The investment casting simulation predicted the presence of cavities, particularly in the area away from the gate for both types of rectangular tubes. In particular, for the rectangular tube with embedded protruded rods cavities were found mainly in the areas between the protruded rods. This simulation result was qualitatively consistent with the experimental observation from the X-ray analysis. Also, both prediction and experiment showed that the dimensional shrinkage (particularly in the longitudinal direction) of the investment-cast rectangular tube is reduced by having protruded rods embedded on the outer surface. Additional numerical attempts were made to check how the amount of cavities and dimensional shrinkage change by varying the preheating temperature and the thickness of the mold. The results predicted that the amount of cavities and the dimensional shrinkage are significantly reduced by increasing the preheating temperature of the mold by 200 ℃. However, an increase in mold thickness from 10 mm to 12 mm showed almost no difference in cavity population and a slight decrease in dimensional shrinkage.

Numerical and Experimental Analysis on the Cavity Formation and Shrinkage for Investment Cast Alloy 738 4 mm-Thick Rectangular Tube

Energy Technology Data Exchange (ETDEWEB)

Park, Myeong-Il; Choi, Yoon Suk; Yoo, Jae-Hyun; Park, Sang-Hu [Pusan National University, Busan (Korea, Republic of); Kim, Kyeong-Min; Lee, Yeong-Chul [Sung Il Turbine Co., Ltd., Busan (Korea, Republic of); Lee, Jung-Seok; Lee, Jae-Hyun [Changwon National University, Changwon (Korea, Republic of)

2017-02-15

Investment casting for the thin (4 mm thick) rectangular tube (40 mm wide, 80 mm high and 200 mm long) was carried out numerically and experimentally for Alloy 738, which is a precipitation-hardened Ni-base superalloy. Two types of rectangular tubes, one with a regular array (10 mm by 10 mm square array) of protruded rods (3 mm in diameter and 3mm in height) embedded on the outer surface and the other with just smooth surface, were investment-cast at the same time through the side feeding mold design. The investment casting simulation predicted the presence of cavities, particularly in the area away from the gate for both types of rectangular tubes. In particular, for the rectangular tube with embedded protruded rods cavities were found mainly in the areas between the protruded rods. This simulation result was qualitatively consistent with the experimental observation from the X-ray analysis. Also, both prediction and experiment showed that the dimensional shrinkage (particularly in the longitudinal direction) of the investment-cast rectangular tube is reduced by having protruded rods embedded on the outer surface. Additional numerical attempts were made to check how the amount of cavities and dimensional shrinkage change by varying the preheating temperature and the thickness of the mold. The results predicted that the amount of cavities and the dimensional shrinkage are significantly reduced by increasing the preheating temperature of the mold by 200 ℃. However, an increase in mold thickness from 10 mm to 12 mm showed almost no difference in cavity population and a slight decrease in dimensional shrinkage.

Microstructure analysis and wear behavior of titanium cermet femoral head with hard TiC layer.

Science.gov (United States)

Luo, Yong; Ge, Shirong; Liu, Hongtao; Jin, Zhongmin

2009-12-11

Titanium cermet was successfully synthesized and formed a thin gradient titanium carbide coating on the surface of Ti6Al4V alloy by using a novel sequential carburization under high temperature, while the titanium cermet femoral head was produced. The titanium cermet phase and surface topography were characterized with X-ray diffraction (XRD) and backscattered electron imaging (BSE). And then the wear behavior of titanium cermet femoral head was investigated by using CUMT II artificial joint hip simulator. The surface characterization indicates that carbon effectively diffused into the titanium alloys and formed a hard TiC layer on the Ti6Al4V alloys surface with a micro-porous structure. The artificial hip joint experimental results show that titanium cermet femoral head could not only improve the wear resistance of artificial femoral head, but also decrease the wear of UHMWPE joint cup. In addition, the carburized titanium alloy femoral head could effectively control the UHMWPE debris distribution, and increase the size of UHMWPE debris. All of the results suggest that titanium cermet is a prospective femoral head material in artificial joint.

SG tube identification

International Nuclear Information System (INIS)

Hoogstraten, P. van

1994-01-01

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

Horizontal beam tubes in FRM-II

International Nuclear Information System (INIS)

Coors, D.; Vanvor, D.

2001-01-01

The new research reactor in Garching FRM-II is equipped with 10 leak tight horizontal beam tubes (BT1 - BT10), each of them consisting of a beam tube structure taking an insert with neutron channels. The design of all beam tube structures is similar whereas the inserts are adapted to the special requirements of the using of each beam tube. Inside the reflector tank the beam tube structures are shaped by the inner cones which are made of Al-alloy with circular and rectangular cross sections. They are located in the region of maximum neutron flux (exception BT10), they are directly connected to the flanges of the reflector tank, their lengths are about 1.5 m (exception BT10) and their axes are directed tagentially to the core centre thus contributing to a low γ-noise at the experiments. (orig.)

Flow behaviour of autoclaved, 20% cold worked, Zr-2.5Nb alloy pressure tube material in the temperature range of room temperature to 800 deg. C

International Nuclear Information System (INIS)

Dureja, A.K.; Sinha, S.K.; Srivastava, Ankit; Sinha, R.K.; Chakravartty, J.K.; Seshu, P.; Pawaskar, D.N.

2011-01-01

Pressure tube material of Indian Heavy Water Reactors is 20% cold-worked and stress relieved Zr-2.5Nb alloy. Inherent variability in the process parameters during the fabrication stages of pressure tube and also along the length of component have their effect on micro-structural and texture properties of the material, which in turn affect its strength parameters (yield strength and ultimate tensile strength) and flow characteristics. Data of tensile tests carried out in the temperature range from room temperature to 800 deg. C using the samples taken out from a single pressure tube have been used to develop correlations for characterizing the strength parameters' variation as a function of axial location along length of the tube and the test temperature. Applicability of Ramberg-Osgood, Holloman and Voce's correlations for defining the post yield behaviour of the material has been investigated. Effect of strain rate change on the deformation behaviour has also been studied.

Dissimilar Joining of ODS and F/M Steel Tube by Friction Stir Welding

Energy Technology Data Exchange (ETDEWEB)

Kang, Suk Hoon; Noh, Sanghoon; Kim, Jun Hwan; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Oxide Dispersion strengthened (ODS) steels, it is well known that uniform nano-oxide dispersoids act as pinning points to obstruct dislocation and grain boundary motion, however, those advantages will be disappeared while the material is subjected to the high temperature of conventional fusion welding. Rotary friction welding, also referred to as friction stir welding (FSW), has shown great promise as a method for welding traditionally difficult to weld materials such as aluminum alloys. This relatively new technology has more recently been applied to higher melting temperature alloys such as steels, nickel-based and titanium alloys. Friction stir processing (FSP) is a method of changing the properties of a metal through intense, localized plastic deformation. FSW is the precursor of the FSP technique. When ideally implemented, this process mixes the material without changing the phase and creates a microstructure with fine, equiaxed grains. This homogeneous grain structure, separated by high-angle boundaries, allows some alloys to take on superplastic properties. In this study, FSW is used as a substitutive welding process between FMS tube and ODS parts. The dimension of tube is 7.0 OD, 0.5 T. During the FSW, dynamic-recrystallized grains are developed; the uniform oxides Dispersion is preserved in the metal matrix. The microstructure and microtexture of the material near the stir zone is found to be influenced by the rotational behavior of the tool. The additive effect from FSP on sample surface is considered. Since the mechanical alloying (MA) and FSP commonly apply extreme shear deformation on materials, the Dispersion of oxide particle in ODS steels is very active during both processes. Friction stir welding appears to be a very promising technique for the welding of FMS and ODS steels in the form of sheet and tube. FSW could successfully produce defect-free welds on FMS tubes and ODS ring assembly. FSW produces a fine grain structure consisting of ferrite and

Dissimilar Joining of ODS and F/M Steel Tube by Friction Stir Welding

International Nuclear Information System (INIS)

Kang, Suk Hoon; Noh, Sanghoon; Kim, Jun Hwan; Kim, Tae Kyu

2014-01-01

Oxide Dispersion strengthened (ODS) steels, it is well known that uniform nano-oxide dispersoids act as pinning points to obstruct dislocation and grain boundary motion, however, those advantages will be disappeared while the material is subjected to the high temperature of conventional fusion welding. Rotary friction welding, also referred to as friction stir welding (FSW), has shown great promise as a method for welding traditionally difficult to weld materials such as aluminum alloys. This relatively new technology has more recently been applied to higher melting temperature alloys such as steels, nickel-based and titanium alloys. Friction stir processing (FSP) is a method of changing the properties of a metal through intense, localized plastic deformation. FSW is the precursor of the FSP technique. When ideally implemented, this process mixes the material without changing the phase and creates a microstructure with fine, equiaxed grains. This homogeneous grain structure, separated by high-angle boundaries, allows some alloys to take on superplastic properties. In this study, FSW is used as a substitutive welding process between FMS tube and ODS parts. The dimension of tube is 7.0 OD, 0.5 T. During the FSW, dynamic-recrystallized grains are developed; the uniform oxides Dispersion is preserved in the metal matrix. The microstructure and microtexture of the material near the stir zone is found to be influenced by the rotational behavior of the tool. The additive effect from FSP on sample surface is considered. Since the mechanical alloying (MA) and FSP commonly apply extreme shear deformation on materials, the Dispersion of oxide particle in ODS steels is very active during both processes. Friction stir welding appears to be a very promising technique for the welding of FMS and ODS steels in the form of sheet and tube. FSW could successfully produce defect-free welds on FMS tubes and ODS ring assembly. FSW produces a fine grain structure consisting of ferrite and

Automated tube voltage adaptation in head and neck computed tomography between 120 and 100 kV: effects on image quality and radiation dose.

Science.gov (United States)

May, Matthias S; Kramer, Manuel R; Eller, Achim; Wuest, Wolfgang; Scharf, Michael; Brand, Michael; Saake, Marc; Schmidt, Bernhard; Uder, Michael; Lell, Michael M

2014-09-01

Low tube voltage allows for computed tomography (CT) imaging with increased iodine contrast at reduced radiation dose. We sought to evaluate the image quality and potential dose reduction using a combination of attenuation based tube current modulation (TCM) and automated tube voltage adaptation (TVA) between 100 and 120 kV in CT of the head and neck. One hundred thirty consecutive patients with indication for head and neck CT were examined with a 128-slice system capable of TCM and TVA. Reference protocol was set at 120 kV. Tube voltage was reduced to 100 kV whenever proposed by automated analysis of the localizer. An additional small scan aligned to the jaw was performed at a fixed 120 kV setting. Image quality was assessed by two radiologists on a standardized Likert-scale and measurements of signal- (SNR) and contrast-to-noise ratio (CNR). Radiation dose was assessed as CTDIvol. Diagnostic image quality was excellent in both groups and did not differ significantly (p = 0.34). Image noise in the 100 kV data was increased and SNR decreased (17.8/9.6) in the jugular veins and the sternocleidomastoid muscle when compared to 120 kV (SNR 24.4/10.3), but not in fatty tissue and air. However, CNR did not differ statistically significant between 100 (23.5/14.4/9.4) and 120 kV data (24.2/15.3/8.6) while radiation dose was decreased by 7-8%. TVA between 100 and 120 kV in combination with TCM led to a radiation dose reduction compared to TCM alone, while keeping CNR constant though maintaining diagnostic image quality.

Manufacturing of FeCrAl/Zr Dual Layer tube for its application to LWR Fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Park, Dong Jun; Lim, Do Wan; Jung, Yang Il; Kim, Hyun Gil; Park, Jeong Yong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-10-15

Many advanced materials such as MAX phases, Mo, SiC, and Fe-based alloys are being considered a possible candidate to substitute the Zr-based alloy cladding has been used in light water reactors. Among the proposed candidate materials, Fe-based alloy is one of the most promising candidates owing to its excellent formability, very good high strength, and corrosion resistance at high temperature. However, neutron cross section of FeCrAl alloy is much higher than that of existing Zr-based alloys. In this study, FeCrAl/Zr dual layer tube was manufactured by using a hot isostatic pressing (HIP) method. The thickness of outer FeCrAl layer was varied from 50 to 250 μm but all the FeCrAl/Zr dual layer tube samples maintained its total thickness of 570 μm. For a detailed microstructural characterization of FeCrAl/Zr dual layer, polarized optical microscopy and scanning electron microscopy (SEM) study carried out and its mechanical property was measured by ring compression test. FeCrAl/Zr dual layer tube sample was successfully manufactured with good adhesion between both layers. Inter layer showing gradual element variation was observed at interface. Result obtained from simulated LOCA test indicates that FeCrAl/Zr dual layer tube may maintain its integrity during LOCA and its accident tolerance had greatly improved compared to that of Zr-based alloy.

The use of titanium for condenser tube bundles

International Nuclear Information System (INIS)

Dobrovitch, N.

2002-01-01

In a power plant, the condenser is a strategic heat exchanger with regards to the efficiency of the steam turbine and its reliability guarantees the performance and continuous operation of the plant. Until the early 1980's, copper alloys were routinely used in condenser tubes, thanks to their high heat transfer rates. Yet numerous problems arose from the use of this material, such as stress cracking corrosion, ammoniacal corrosion, fouling, erosion, dezincification, abrasion, erosion-corrosion,... and lately the problem of inadequateness of copper with nuclear steam generators (in nuclear power plant the abrasion problem of the copper alloy tubes created a deposit problem in the steam generator conducting to the replacement of all the condensers). The trend was then to consider new tube materials, such stainless steel and titanium, firstly for particular operating conditions and now for most of the projects, with several objectives, such as: 1) improve the reliability (titanium in particular can bring major improvements such as higher water velocities promoting better heat coefficients, excellent resistance to abrasion, erosion and corrosion thereby improving resistance to fouling; 2) find more cost-effective solutions. The first investment is higher but money is saved on maintenance costs and on time reliability of the material. Titanium tube manufacturing has greatly evolved for the last 20 years. Tubes are mostly welded tubes from ASTM SB 338 grade 1 made on a continuous manufacturing line. All manufacturing operations (welding, annealing, non-destructive testing) are fully automated to produce high quality tubes in large quantities. The most common way to attach tubes to a tubesheet is to roller expand them. (A.C.)

Hydrogen charging, hydrogen content analysis and metallographic examination of hydride in zirconium alloys

International Nuclear Information System (INIS)

Singh, R.N.; Kishore, R.; Mukherjee, S.; Roychowdhury, S.; Srivastava, D.; Sinha, T.K.; De, P.K.; Banerjee, S.; Gopalan, B.; Kameswaran, R.; Sheelvantra, Smita S.

2003-12-01

Gaseous and electrolytic hydrogen charging techniques for introducing controlled amount of hydrogen in zirconium alloy is described. Zr-1wt%Nb fuel tube, zircaloy-2 pressure tube and Zr-2.5Nb pressure tube samples were charged with up to 1000 ppm of hydrogen by weight using one of the aforementioned methods. These hydrogen charged Zr-alloy samples were analyzed for estimating the total hydrogen content using inert gas fusion technique. Influence of sample surface preparation on the estimated hydrogen content is also discussed. In zirconium alloys, hydrogen in excess of the terminal solid solubility precipitates out as brittle hydride phase, which acquire platelet shaped morphology due to its accommodation in the matrix and can make the host matrix brittle. The F N number, which represents susceptibility of Zr-alloy tubes to hydride embrittlement was measured from the metallographs. The volume fraction of the hydride phase, platelet size, distribution, interplatelet spacing and orientation were examined metallographically using samples sliced along the radial-axial and radial-circumferential plane of the tubes. It was observed that hydride platelet length increases with increase in hydrogen content. Considering the metallographs generated by Materials Science Division as standard, metallographs prepared by the IAEA round robin participants for different hydrogen concentration was compared. It is felt that hydride micrographs can be used to estimate not only that approximate hydrogen concentration of the sample but also its size, distribution and orientation which significantly affect the susceptibility to hydride embrittlement of these alloys. (author)

Method for automatic filling of nuclear fuel rod cladding tubes

International Nuclear Information System (INIS)

Bezold, H.

1979-01-01

Prior to welding the zirconium alloy cladding tubes with end caps, they are automatically filled with nuclear fuel tablets and ceramic insulating tablets. The tablets are introduced into magazine drums and led through a drying oven to a discharging station. The empty cladding tubes are removed from this discharging station and filled with tablets. A filling stamp pushes out the columns of tablets in the magazine tubes of the magazine drum into the cladding tube. Weight and measurement of length determine the filled state of the cladding tube. The cladding tubes are then led to the welding station via a conveyor belt. (DG) [de

Studies on the permeation of hydrogen through steam generator tubes at high temperatures using an electrochemical method

International Nuclear Information System (INIS)

Giraudeau, F.; Yang, L.; Steward, F.R.; DeBouvier, O.

1998-01-01

The permeation of hydrogen through steam generator tubes at high temperatures (∼ 300 degrees C) has been studied using an electrochemical technique. With this technique, hydrogen is generated on one side of the tube and monitored on the other side. The time for the hydrogen to reach the other side is used to determine the diffusion coefficient of hydrogen in the tube. Boundary conditions at the entry and exit sides have been investigated separately. Preliminary studies were performed on Stainless Steel 316 and Nickel Alloy 800 to better understand the influence of the solution chemistry on the electrochemical evolution of hydrogen. The surface phenomena effect and the trapping effect are discussed to account for differences observed in the permeation response. The hydrogen permeation through oxides at the exit side has been studied. Two nickel alloys (Alloy 800 and Alloy 600), materials widely used for steam generator tubes, have been investigated. The tubes were prefilmed using two different treatments. The oxides were formed in dry air at high temperatures (300 degrees C to 600 degrees C), or in humid gas at 300 degrees C. The diffusion coefficients at 300 degrees C in Stainless Steel 316 and Alloy 800 were determined to be of the order of 10 -6 - 10 -7 cm 2 /s for the bare metal. This is in agreement with results obtained by gas phase permeation techniques in the literature. (author)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Evaluation of Fatigue Crack Initiation for Volumetric Flaw in Pressure Tube

International Nuclear Information System (INIS)

Choi, Sung Nam; Yoo, Hyun Joo

2005-01-01

CAN/CSA.N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA-N285.05-2005, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of ASME B and PV Sec. XI, 'Inservice Inspection of Nuclear Power Plant Components'. However, the evaluation methodology for a blunt volumetric flaw is described in CSA-N285.05-2005 code. The object of this paper is to address the fatigue crack initiation evaluation for the blunt volumetric flaw as it applies to the pressure tube at Wolsong NPP

Characterization of BOR-60 Irradiated 14YWT-NFA1 Tubes

Energy Technology Data Exchange (ETDEWEB)

Saleh, Tarik A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Maloy, Stuart Andrew [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Aydogan, Eda [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Quintana, Matthew Estevan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Romero, Tobias J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-02-15

Tubes of FCRD 14YWT-NFA1 Alloy were placed in the BOR-60 reactor and irradiated under a fast flux neutron environment to two conditions: 7 dpa at 360-370 °C and 6 dpa at 385-430 °C. Small sections of the tube were cut and sent to UC Berkeley for nanohardness testing and focused ion beam (FIB) milling of TEM specimens. FIB specimens were sent back to LANL for final FIB milling and TEM imaging. Hardness data and TEM images are presented in this report. This is the first fast reactor neutron irradiated information on the 14YWT-NFA1 alloy.

Irradiation creep of dispersion strengthened copper alloy

Energy Technology Data Exchange (ETDEWEB)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A. [and others

1997-04-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al{sub 2}O{sub 3}, is very similar to the GlidCop{trademark} alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10{sup 21} n/cm{sup 2} (E>0.1 MeV), which corresponds to {approx}3-5 dpa. The irradiation temperature ranged from 60-90{degrees}C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of {+-}0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as {approx}2 x 10{sup {minus}9}s{sup {minus}1}. These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys.

Irradiation creep of dispersion strengthened copper alloy

International Nuclear Information System (INIS)

Pokrovsky, A.S.; Barabash, V.R.; Fabritsiev, S.A.

1997-01-01

Dispersion strengthened copper alloys are under consideration as reference materials for the ITER plasma facing components. Irradiation creep is one of the parameters which must be assessed because of its importance for the lifetime prediction of these components. In this study the irradiation creep of a dispersion strengthened copper (DS) alloy has been investigated. The alloy selected for evaluation, MAGT-0.2, which contains 0.2 wt.% Al 2 O 3 , is very similar to the GlidCop trademark alloy referred to as Al20. Irradiation creep was investigated using HE pressurized tubes. The tubes were machined from rod stock, then stainless steel caps were brazed onto the end of each tube. The creep specimens were pressurized by use of ultra-pure He and the stainless steel caps subsequently sealed by laser welding. These specimens were irradiated in reactor water in the core position of the SM-2 reactors to a fluence level of 4.5-7.1 x 10 21 n/cm 2 (E>0.1 MeV), which corresponds to ∼3-5 dpa. The irradiation temperature ranged from 60-90 degrees C, which yielded calculated hoop stresses from 39-117 MPa. A mechanical micrometer system was used to measure the outer diameter of the specimens before and after irradiation, with an accuracy of ±0.001 mm. The irradiation creep was calculated based on the change in the diameter. Comparison of pre- and post-irradiation diameter measurements indicates that irradiation induced creep is indeed observed in this alloy at low temperatures, with a creep rate as high as ∼2 x 10 -9 s -1 . These results are compared with available data for irradiation creep for stainless steels, pure copper, and for thermal creep of copper alloys

Effects of microstructure and mechanical properties of alloys 600 an 690 on secondary side SCC

International Nuclear Information System (INIS)

Vaillant, F.; Buisine, D.; Prieux, B.; Fournel, J.C.; Gelpi, A.

1996-03-01

Modeling for secondary side cracking is needed to understand the behaviour of alloy 600 in plants. They require a comprehensive understanding of the various influences of the material properties on Stress Corrosion Cracking (SCC), based on field experience and laboratory data. In an attempt to predict the materials effects on SCC behaviour of new steam generators, laboratory corrosion data of alloy 690 were overviewed. French field experience with steam generators equipped with drilled tube support plates (TSPs) has demonstrated that the lower the yield stress (YS) and the carbon content, the higher the susceptibility t secondary side cracking of mill-annealed (MA) alloy 600. Also heat treated (700 deg. C x 16 h) tubing has been shown to have a much better resistance, but this excellent resistance could not be attributed only to the material properties. In laboratory environments, particularly in caustics, results have confirmed several of the above mentioned key findings on alloy 600: in caustic environments and under constant loading, tubes fabricated from MA alloy 600 with low YS have exhibited the worst resistance to initiation; YS was found to be the most accurate parameter to account for the behaviour of MA alloy 600. A heat treatment at 700 deg. C appeared to reduce the propagation rates of cracks in alloy 600. The best IGSCC resistance of alloy 690 was obtained for tubes with intergranular precipitation of carbides. TT (700 deg. C) significantly improved the propagation resistance of alloy 690; in acidic and neutral sulfate environments, IGSCC of alloy 600 was not strongly dependent on the microstructure in the MA condition, but sensitization was detrimental. When alloy 600 and particularly alloy 690 were thermally treated at 700 deg. C x 16 h, the resistance to IGSCC was significantly improved. Tests performed on alloy 690 have shown a better resistance to IGSCC initiation and propagation than alloy 600, in NaOH and acidic sulfate environments. (authors

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Penetrameter positioner for bore-side radiography of tubes

International Nuclear Information System (INIS)

Davis, E.V.; Foster, B.E.

1983-01-01

A positioner is provided for placing plaque or wire penetrameters, as used in radiographic inspection, in close proximity with the inner wall of tubing at any desired location along the tubing. The positioner head carrying the penetrameter is inflatable whereby it is positioned in the deflated condition, inflated to place the penetrameter against a weld to be inspected in the tubing wall, and then deflated during removal. If desired, the penetrameter holder may be used to center the radiographic source on the axis of the tube

High-temperature air oxidation of E110 and Zr-1%Nb alloys claddings with coatings

International Nuclear Information System (INIS)

Kuprin, A.S.; Belous, V.A.; Voyevodin, V.N.; Bryk, V.V.; Vasilenko, R.L.; Ovcharenko, V.D.; Tolmachova, G.N.; V'yugov, P.N.

2014-01-01

Results of experimental study of the influence of protective vacuum-arc claddings on the base of compounds zirconium-chromium and of its nitrides on air oxidation resistance at temperatures 660, 770, 900, 1020, 1100 deg C during 3600 s. of tubes produced of zirconium alloys E110 and Zr-1%Nb (calcium-thermal alloy of Ukrainian production) are presented. Change of hardness, the width of oxide layer and depth of oxygen penetration into alloys from the side of coating and without coating are investigated by the methods of nanoindentation and by scanning electron microscopy. It is shown that the thickness of oxide layer in zirconium alloys at temperatures 1020 and 1100 deg C from the side of the coating doesn't exceed 5 μm, and from the unprotected side reaches the value of ≥ 120 μm with porous and rough structure. Tubes with coatings save their shape completely independently of the type of alloy; tubes without coatings deform with the production of through cracks

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Ploc, R.A.; McRae, G.A.

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable β-Zr that is present as a thin layer encasing the predominant α-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded β-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the β-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Ploc, R.A.; McRae, G.A

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable {beta}-Zr that is present as a thin layer encasing the predominant {alpha}-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded {beta}-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the {beta}-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Singh, R.N., E-mail: rnsingh@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Viswanathan, U.K.; Kumar, Sunil; Satheesh, P.M.; Anantharaman, S. [Post Irradiation Examination Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Stahle, P. [Division of Solid Mechanics, Lund University/LTH, SE22100 Lund (Sweden)

2011-07-15

Highlights: > For the first time impact behaviour of Zr-2.5Nb pressure tube material used in Indian Pressurized Heavy Water Reactor (IPHWR) as a function of hydrogen content and temperature is being reported. > The critical hydrogen concentration to cause low energy fracture at 25 and 200 deg. C is suggested. > The impact behaviour is rationalized in terms of hydrogen content, test temperature, microstructural features and state of stress ahead of a crack. - Abstract: Influence of hydrogen content on the impact toughness of Zr-2.5% Nb alloy was examined by carrying out instrumented drop weight tests in the temperature range of 25-250 deg. C using curved Charpy specimens fabricated from unirradiated pressure tubes of Indian Pressurized Heavy Water Reactor (IPHWR). Hydrogen content of the samples was between 10 and 170 ppm by weight (wppm). Sharp ductile-to-brittle-transition behaviour was demonstrated by hydrided materials. The temperature for the onset of transition increased with the increase in the hydrogen content of the specimens. The fracture surfaces of unhydrided specimen exhibited ductile fracture caused by micro void coalescence and tear ridges at lower temperatures and by fibrous fracture at intermediate and at higher temperatures. Except for the samples tested at the upper shelf energy levels, the fracture surfaces of all hydrided samples were suggestive of hydride assisted failure. In most cases the transverse cracks observed in the fracture path matched well with the hydride precipitate distribution and orientation.

Use of a holder-vacuum tube device to save on-site hands in preparing urine samples for head-space gas-chromatography, and its application to determine the time allowance for sample sealing.

Science.gov (United States)

Kawai, Toshio; Sumino, Kimiaki; Ohashi, Fumiko; Ikeda, Masayuki

2011-01-01

To facilitate urine sample preparation prior to head-space gas-chromatographic (HS-GC) analysis. Urine samples containing one of the five solvents (acetone, methanol, methyl ethyl ketone, methyl isobutyl ketone and toluene) at the levels of biological exposure limits were aspirated into a vacuum tube via holder, a device commercially available for venous blood collection (the vacuum tube method). The urine sample, 5 ml, was quantitatively transferred to a 20-ml head-space vial prior to HS-GC analysis. The loaded tubes were stored at +4 ℃ in dark for up to 3 d. The vacuum tube method facilitated on-site procedures of urine sample preparation for HS-GC with no significant loss of solvents in the sample and no need of skilled hands, whereas on-site sample preparation time was significantly reduced. Furthermore, no loss of solvents was detected during the 3-d storage, irrespective of hydrophilic (acetone) or lipophilic solvent (toluene). In a pilot application, high performance of the vacuum tube method in sealing a sample in an air-tight space succeeded to confirm that no solvent will be lost when sealing is completed within 5 min after urine voiding, and that the allowance time is as long as 30 min in case of toluene in urine. The use of the holder-vacuum tube device not only saves hands for transfer of the sample to air-tight space, but facilitates sample storage prior to HS-GC analysis.

Topogram-based tube current modulation of head computed tomography for optimizing image quality while protecting the eye lens with shielding.

Science.gov (United States)

Lin, Ming-Fang; Chen, Chia-Yuen; Lee, Yuan-Hao; Li, Chia-Wei; Gerweck, Leo E; Wang, Hao; Chan, Wing P

2018-01-01

Background Multiple rounds of head computed tomography (CT) scans increase the risk of radiation-induced lens opacification. Purpose To investigate the effects of CT eye shielding and topogram-based tube current modulation (TCM) on the radiation dose received by the lens and the image quality of nasal and periorbital imaging. Material and Methods An anthropomorphic phantom was CT-scanned using either automatic tube current modulation or a fixed tube current. The lens radiation dose was estimated using cropped Gafchromic films irradiated with or without a shield over the orbit. Image quality, assessed using regions of interest drawn on the bilateral extraorbital areas and the nasal bone with a water-based marker, was evaluated using both a signal-to-noise ratio (SNR) and contrast-noise ratio (CNR). Two CT specialists independently assessed image artifacts using a three-point Likert scale. Results The estimated radiation dose received by the lens was significantly lower when barium sulfate or bismuth-antimony shields were used in conjunction with a fixed tube current (22.0% and 35.6% reduction, respectively). Topogram-based TCM mitigated the beam hardening-associated artifacts of bismuth-antimony and barium sulfate shields. This increased the SNR by 21.6% in the extraorbital region and the CNR by 7.2% between the nasal bones and extraorbital regions. The combination of topogram-based TCM and barium sulfate or bismuth-antimony shields reduced lens doses by 12.2% and 27.2%, respectively. Conclusion Image artifacts induced by the bismuth-antimony shield at a fixed tube current for lenticular radioprotection were significantly reduced by topogram-based TCM, which increased the SNR of the anthropomorphic nasal bones and periorbital tissues.

Melting and casting of FeAl-based cast alloy

Energy Technology Data Exchange (ETDEWEB)

Sikka, V.K. [Oak Ridge National Lab., TN (United States); Wilkening, D. [Columbia Falls Aluminum Co., Columbia Falls, MT (United States); Liebetrau, J.; Mackey, B. [AFFCO, L.L.C., Anaconda, MT (United States)

1998-11-01

The FeAl-based intermetallic alloys are of great interest because of their low density, low raw material cost, and excellent resistance to high-temperature oxidation, sulfidation, carburization, and molten salts. The applications based on these unique properties of FeAl require methods to melt and cast these alloys into complex-shaped castings and centrifugal cast tubes. This paper addresses the melting-related issues and the effect of chemistry on the microstructure and hardness of castings. It is concluded that the use of the Exo-Melt{trademark} process for melting and the proper selection of the aluminum melt stock can result in porosity-free castings. The FeAl alloys can be melted and cast from the virgin and revert stock. A large variation in carbon content of the alloys is possible before the precipitation of graphite flakes occurs. Titanium is a very potent addition to refine the grain size of castings. A range of complex sand castings and two different sizes of centrifugal cast tubes of the alloy have already been cast.

Properties of the chalcogenide–carbon nano tubes and graphene composite materials

International Nuclear Information System (INIS)

Singh, Abhay Kumar; Kim, JunHo; Park, Jong Tae; Sangunni, K.S.

2015-01-01

Highlights: • Chalcogenides. • Melt quenched. • Composite materials. • Multi walled carbon nano tubes. • Bilayer graphene. - Abstract: Composite can deliver more than the individual elemental property of the material. Specifically chalcogenide- multi walled carbon nano tubes and chalcogenide- bilayer graphene composite materials could be interesting for the investigation, which have been less covered by the investigators. We describe micro structural properties of Se 55 Te 25 Ge 20, Se 55 Te 25 Ge 20 + 0.025% multi walled carbon nano tubes and Se 55 Te 25 Ge 20 + 0.025% bilayer graphene materials. This gives realization of the alloying constituents inclusion/or diffusion inside the multi walled carbon nano tubes and bilayer graphene under the homogeneous parent alloy configuration. Raman spectroscopy, X-ray photoelectron spectroscopy, UV/Visible spectroscopy and Fourier transmission infrared spectroscopy have also been carried out under the discussion. A considerable core energy levels peak shifts have been noticed for the composite materials by the X-ray photoelectron spectroscopy. The optical energy band gaps are measured to be varied in between 1.2 and 1.3 eV. In comparison to parent (Se 55 Te 25 Ge 20 ) alloy a higher infrared transmission has been observed for the composite materials. Subsequently, variation in physical properties has been explained on the basis of bond formation in solids

Development of low temperature solid state joining technology of dissimilar for nuclear heat exchanger tube components

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-08-15

By conventional fusion welding process (TIG), a realization of reliable and sound joints for the nuclear heat exchanger components is very difficult, especially for the parts comprising of the dissimilar metal couples (Ti-STS, Ti-Cu alloy etc.). This is mainly attributed to the formation of brittle intermetallics (Ti{sub x}Cu{sub y}, Ti{sub x}Fe{sub y}, Ti{sub x}Ni{sub y} etc.) and wide difference in physical properties. Moreover, it usually employs very high thermal input, so making it difficult to obtain sound joints due to generations of high residual stresses and degradation of the adjacent base metals, even for similar metal combinations. In this project, the low temperature solid-state joining technology was established by developing new alloy fillers, e.g. the multi-component eutectic based alloys or amorphous alloys, and thereby lowering the joining temperature down to {approx}800 .deg. C without affecting the structural properties of base metals. Based on a low temperature joining, the interlayer engineering technology was then developed to be able to eliminate the brittleness of the joints for strong Ti-STS dissimilar joints, and the diffusion brazing technology of Ti-Ti with a superior joining strength and corrosion-resistance comparable to those of base metal were developed. By using those developed technologies, the joining procedures feasible for the heat exchanger components were finally established for the dissimilar metal joints including Ti tube sheet to super STS tube, Ti tube sheet to super STS tube sheet, and the joints of the Ti tube to Ti tube sheet

A novel Fe–Cr–Nb matrix composite containing the TiB_2 neutron absorber synthesized by mechanical alloying and final hot isostatic pressing (HIP) in the Ti-tubing

International Nuclear Information System (INIS)

Litwa, Przemysław; Perkowski, Krzysztof; Zasada, Dariusz; Kobus, Izabela; Konopka, Gustaw; Czujko, Tomasz; Varin, Robert A.

2016-01-01

The Fe–Cr–Ti-Nb elemental powders were mechanically alloyed/ball milled with TiB_2 and a small quantity of Y_2O_3 ceramic to synthesize a novel Fe-based alloy-ceramic powder composite that could be processed by hot isostatic pressing (HIP) for a perceived potential application as a neutron absorber in nuclear reactors. After ball milling for the 30–80 h duration relatively uniform powders with micrometric sizes were produced. With increasing milling time a fraction of TiB_2 particles became covered with the much softer Fe-based alloy which resulted in the formation of a characteristic “core-mantel” structure. For the final HIP-ing process the mechanically alloyed powders were initially uniaxially pressed into rod-shaped compacts and then cold isostatically pressed (CIP-ed). Subsequently, the rod-shaped compacts were placed in the Ti-tubing and subjected to hot isostatic pressing (HIP) at 1150 °C/200 MPa pressure. The HIP-ing process resulted in the formation of the near-Ti and intermediate diffusional layers in the microstructure of HIP-ed samples which formed in accord with the Fe-Ti binary phase diagram. Those layers contain the phases such as α-Ti (HCP), the FeTi intermetallic and their hypo-eutectoid mixtures. In addition, needle-like particles were formed in both layers in accord with the Ti-B binary phase diagram. Nanohardness testing, using a Berkovich type diamond tip, shows that the nanohardness in the intermediate layer areas, corresponding to the composition of the hypo-eutectoid mixture of Ti-FeTi, equals 980.0 (±27.1) HV and correspondingly 1176.9 (±47.6) HV for the FeTi phase. The nanohardness in the sample's center in the areas with the fine mixture of Fe-based alloy and small TiB_2 particles equals 1048.3 (±201.8) HV. The average microhardness of samples HIP-ed from powders milled for 30 and 80 h is 588 HV and 733 HV, respectively. - Highlights: • A Fe–Cr–Nb-based composite with TiB_2 neutron absorbing ceramic was mechanically

Predicting tube repair at French nuclear steam generators using statistical modeling

Energy Technology Data Exchange (ETDEWEB)

Mathon, C., E-mail: cedric.mathon@edf.fr [EDF Generation, Basic Design Department (SEPTEN), 69628 Villeurbanne (France); Chaudhary, A. [EDF Generation, Basic Design Department (SEPTEN), 69628 Villeurbanne (France); Gay, N.; Pitner, P. [EDF Generation, Nuclear Operation Division (UNIE), Saint-Denis (France)

2014-04-01

Electricité de France (EDF) currently operates a total of 58 Nuclear Pressurized Water Reactors (PWR) which are composed of 34 units of 900 MWe, 20 units of 1300 MWe and 4 units of 1450 MWe. This report provides an overall status of SG tube bundles on the 1300 MWe units. These units are 4 loop reactors using the AREVA 68/19 type SG model which are equipped either with Alloy 600 thermally treated (TT) tubes or Alloy 690 TT tubes. As of 2011, the effective full power years of operation (EFPY) ranges from 13 to 20 and during this time, the main degradation mechanisms observed on SG tubes are primary water stress corrosion cracking (PWSCC) and wear at anti-vibration bars (AVB) level. Statistical models have been developed for each type of degradation in order to predict the growth rate and number of affected tubes. Additional plugging is also performed to prevent other degradations such as tube wear due to foreign objects or high-cycle flow-induced fatigue. The contribution of these degradation mechanisms on the rate of tube plugging is described. The results from the statistical models are then used in predicting the long-term life of the steam generators and therefore providing a useful tool toward their effective life management and possible replacement.

Effect of texture on creep deformation behavior of Zr-2.5Nb alloy

International Nuclear Information System (INIS)

Guguloth, Krishna; Swaminathan, J.; Mitra, Rahul; Ghosh, R.N.; Singh, R.N.; Chakravartty, J.K.

2016-01-01

Zr-2.5%Nb alloys are extensively used as high temperature pressure tubes in nuclear reactor. Therefore creep behavior of this alloy is of considerable importance. The paper presents creep strain-time plots on two sets of specimens made from two as received pressure tubes having different diameters. These tubes were reported to have undergone different processing routes; both tubes were autoclaved at the same temperature in the steam atmosphere. A comparison of the creep strain-time plots of the two sets of specimen under identical test conditions showed a marked difference. The chemical composition and the microstructure of the two sets of samples were also found to be similar. Therefore X-ray diffraction patterns were taken from the two tubes. The ratio of intensity of two prominent reflections from 0002 and 1120 planes of alpha Zr in the case of 90mm tube was found to be 1.79; whereas that from the 110mm tube was 0.25. This suggests that in the case of 110mm tube most of the basal planes were less favorably oriented with respect to the loading axis. This is the reason why creep strength of 110mm tube was found to be higher. The paper also describes how the effect of texture can be incorporated in evaluating the creep behavior of Zr-Nb alloy. This suggests that a relatively larger volume of creep test data on Zr-2.5Nb pressure tube is necessary to account for the effect texture so that a reliable estimate of its creep life could be obtained. (author)

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

Directory of Open Access Journals (Sweden)

HANSUB CHUNG

2013-08-01

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

Bromine-quenched high temperature G-M tube with passivated cathode

International Nuclear Information System (INIS)

Mitrofanov, N.

1975-01-01

A bromine doped self-quenching Geiger-Mueller tube having an operational life expectancy in excess of 1,200 hours at a temperature of 315 0 C is described. The tube comprises a passivated metal coated cathode which is conditioned or aged for operation at room temperature, thus obviating the necessity of thermally cycling the tube at progressively elevated temperatures. Useful metal coatings for the cathode include chromium, platinum, and nickel-copper alloys deposited in a layer less than about 1 mil thick. A method for passivating the metal coated cathode and subsequently conditioning the tube and its contents is disclosed. (auth)

Control rod guide tube of nuclear reactor

International Nuclear Information System (INIS)

Suda, Yoshitaka; Ito, Kenji; Matsumoto, Kunio.

1994-01-01

Zr having a residual tensile stress of 3 to 10kg/mm 2 in a circumferential direction is used for the main ingredient of a control guide tube of a nuclear reactor. For this purpose, an appropriate correction method such as a roll-correction, tension-correction and press-correction method is applied to an existent Zr-base alloy tube with no substantial residual stress. If the residual tensile stress in the circumferential direction is smaller than 3kg/mm 2 , an effect sufficient to suppress irradiation growth is not obtainable, if it exceeds 10kg/mm 2 , dimensional changes, cracks or the like occurs locally since the wall thickness of the control rod guide tube is small and, accordingly, this often results in failed products as the control guide tube. (N.H.)

Reactor fuel cladding tube with excellent corrosion resistance and method of manufacturing the same

International Nuclear Information System (INIS)

Okuda, Takanari; Kanehara, Mitsuo; Abe, Katsuhiro; Nishimura, Takashi.

1995-01-01

The present invention provides a fuel cladding tube having an excellent corrosion resistance and thus a long life, and a suitable manufacturing method therefor. Namely, in the fuel cladding tube, the outer circumference of an inner layer made of a zirconium base alloy is coated with an outer layer made of a metal more corrosion resistant than the zirconium base alloy. Ti or a titanium alloy is suitable for the corrosion resistant metal. In addition, the outer layer can be coated by a method such as vapor deposition or plating, not limited to joining of the inner layer material and the outer layer material. Specifically, a composite material having an inner layer made of a zirconium alloy coated by the outer material made of a titanium alloy is applied with hot fabrication at a temperature within a range of from 500 to 850degC and at a fabrication rate of not less than 5%. The fabrication method includes any of extrusion, rolling, drawing, and casting. As the titanium-base alloy, a Ti-Al alloy or a Ti-Nb alloy containing Al of not more than 20wt%, or Nb of not more than 20wt% is preferred. (I.S.)

Method of and apparatus for use in joining tubular components and tube assemblies made thereby

International Nuclear Information System (INIS)

Percival, S.R.

1979-01-01

A method of joining difficult to weld materials involves the forming of a rolled joint. A particular application is the joining of zirconium alloy calandria tubes to stainless steel tube-plates in a SGHWR. (UK)

Feasibility of Shape-Memory Ni/Ti Alloy Wire Containing Tube Elevators for Transcrestal Detaching Maxillary Sinus Mucosa: Ex Vivo Study

Directory of Open Access Journals (Sweden)

Yanfeng Li

2016-12-01

Full Text Available Background: Osteotome sinus floor elevation is a less invasive approach to augment an insufficient alveolar bone at the posterior maxilla for dental implantation. However, this approach has some limitations due to the lack of sinus lift tools available for clinical use and the small transcrestal access to the maxillary sinus floor. We recently invented shape-memory Ni/Ti alloy wire containing tube elevators for transcrestal detaching maxillary sinus mucosa, and developed goat ex vivo models for direct visualizing the effectiveness of detaching sinus mucosa in real time during transcrestal maxillary sinus floor elevation. Methods: We evaluated our invented elevators, namely elevator 012 and elevator 014, for their effectiveness for transcrestal detaching maxillary sinus mucosa using the goat ex vivo models. We measured the length of sinus mucosa detached in mesial and distal directions or buccal and palatal directions, and the space volume created by detaching maxillary sinus mucosa in mesial, distal, buccal and palatal directions using the invented elevators. Results: Elevator 012 had a shape-memory Ni/Ti alloy wire with a diameter of 0.012 inch, while elevator 014 had its shape-memory Ni/Ti alloy wire with a diameter of 0.014 inch. Elevator 012 could detach the goat maxillary sinus mucosa in the mesial or distal direction for 12.1±4.3 mm, while in the buccal or palatal direction for 12.5±6.7 mm. The elevator 014 could detach the goat maxillary sinus mucosa for 23.0±4.9 mm in the mesial or distal direction, and for 19.0±8.1 mm in the buccal or palatal direction. An average space volume of 1.7936±0.2079 ml was created after detaching the goat maxillay sinus mucosa in both mesial/distal direction and buccal/palatal direction using elevator 012; while the average space volume created using elevator 014 was 1.8764±0.2366 ml. Conclusion: Both two newly invented tube elevators could effectively detach the maxillary sinus mucosa on the goat ex

Feasibility of Shape-Memory Ni/Ti Alloy Wire Containing Tube Elevators for Transcrestal Detaching Maxillary Sinus Mucosa: Ex Vivo Study.

Science.gov (United States)

Li, Yanfeng; Wang, Fuli; Hu, Pin; Fan, Jiadong; Han, Yishi; Liu, Bin; Liu, Tao; Yang, Chunhao; Gu, Xiangmin

2016-01-01

Osteotome sinus floor elevation is a less invasive approach to augment an insufficient alveolar bone at the posterior maxilla for dental implantation. However, this approach has some limitations due to the lack of sinus lift tools available for clinical use and the small transcrestal access to the maxillary sinus floor. We recently invented shape-memory Ni/Ti alloy wire containing tube elevators for transcrestal detaching maxillary sinus mucosa, and developed goat ex vivo models for direct visualizing the effectiveness of detaching sinus mucosa in real time during transcrestal maxillary sinus floor elevation. We evaluated our invented elevators, namely elevator 012 and elevator 014, for their effectiveness for transcrestal detaching maxillary sinus mucosa using the goat ex vivo models. We measured the length of sinus mucosa detached in mesial and distal directions or buccal and palatal directions, and the space volume created by detaching maxillary sinus mucosa in mesial, distal, buccal and palatal directions using the invented elevators. Elevator 012 had a shape-memory Ni/Ti alloy wire with a diameter of 0.012 inch, while elevator 014 had its shape-memory Ni/Ti alloy wire with a diameter of 0.014 inch. Elevator 012 could detach the goat maxillary sinus mucosa in the mesial or distal direction for 12.1±4.3 mm, while in the buccal or palatal direction for 12.5±6.7 mm. The elevator 014 could detach the goat maxillary sinus mucosa for 23.0±4.9 mm in the mesial or distal direction, and for 19.0±8.1 mm in the buccal or palatal direction. An average space volume of 1.7936±0.2079 ml was created after detaching the goat maxillay sinus mucosa in both mesial/distal direction and buccal/palatal direction using elevator 012; while the average space volume created using elevator 014 was 1.8764±0.2366 ml. Both two newly invented tube elevators could effectively detach the maxillary sinus mucosa on the goat ex vivo sinus models. Moreover, elevator 014 has advantages over

Thin-walled large-diameter zirconium alloy tubes in CANDU reactors

International Nuclear Information System (INIS)

Price, E.G.; Richinson, P.J.

1978-08-01

The requirements of the thin-walled large-diameter Zircaloy-2 tubing used in CANDU reactors are reviewed. Strength, residual stress patterns, texture and prior deformation contribute to the stability of these tubes. The extent to which the present manufacturing route meets these requirements is discussed. (author)

Progressive degradation of alloy 690 and the development of a significant improvement in alloy 800CR

International Nuclear Information System (INIS)

Staehle, Roger W.; Arioka, Koji; Tapping, Robert

2015-01-01

The present most widely used alloys for tubing in steam generators and structural materials in water cooled reactors are Alloy 690 and Alloy 800. However, both alloys, while improved over Alloy 600 may not meet the needs of longer range applications in the range of 80-100 years. Alloy 690 sustains damage resulting from the formation of cavities at grain boundaries which eventually cover about 50% of the area of the grain boundaries with the remainder covering being covered with carbides. The cavities seem to nucleate on the carbides leaving the grain boundaries a structure of cavities and carbides. Such a structure will lead the Alloy 690 to fail completely. Normal Alloy 800 does not produce such cavities and probably retains a large amount of its corrosion resistance but does sustain progressive SCC at low rate. A new alloy, 800CR, has been developed in a collaboration among Arioka, Tapping, and Staehle. This alloy is based on a Cr composition of 23.5-27% with the remainder retaining the previous Alloy 800 composition. 800CR sustains a crack velocity about 100 times less than Alloy 690 and a negligible rate of initiation. The 800CR, alloy is now seeking a patent. (authors)

Effect of lead and silicon on localized corrosion of Alloy 800 in steam generator crevice environments

International Nuclear Information System (INIS)

Lu, Y.C.; Wright, M.D.; Cleland, R.D.

2001-09-01

The Alloy 800 tubes used in CANDU 6 steam generators have not experienced significant corrosion damage to date, which may be attributed to successful water chemistry control strategies. However, it is known that Alloy 800, like other steam generator (SG) tubing materials, is not immune to corrosion, especially pitting, under some plausible but off-specification operating scenarios. Electrochemical measurements provide information on corrosion susceptibility and rate, which are known to be a function of water chemistry. Using laboratory data in combination with chemistry monitoring and diagnostic software it is possible to assess the impact of plant operating conditions on SG tube corrosion for plant life management (PLIM). In this context, this paper discusses the results of electrochemical measurements made to elucidate the corrosion behaviour of Alloy 800 SG tubes under conditions simulating those plausible in SG crevices. In addition to crevice pH, the influence of PbO, acting alone or in combination with SiO 2 , on localized corrosion such as pitting or stress corrosion-cracking (SCC) was determined. Possible transient chemistry regimes that could significantly shorten expected tube lifetimes have been identified from the data analysis. Of equal significance, the results also support the position that under normal, near neutral pH and low dissolved oxygen conditions, pitting and cracking of Alloy 800 steam generator tubing will not be initiated. (author)

Development of Alloy 718 tubular product for nuclear-reactor internals

International Nuclear Information System (INIS)

1981-01-01

The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides mixing and flow direction of the core outlet flow. Alloy 718 tubes are the major components used in the UIS to provide this flow direction. The UIS is located directly above the reactor core and is exposed to a severe environment. This environment consists of high temperature sodium, alternating temperatures induced by mixing high temperature core assembly outlet flow with cooler core assembly outlet flow and rapid changes in temperature of the core outlet flow. The paper presents the UIS configuration, functions and environmental conditions that led to the selection of Alloy 718 as the material used to protect the basic UIS structure and to provide flow direction. The use of Alloy 718 is derived from the technology produced from the Department of Energy sponsored development programs. Alloy 718 tubes are made by a roll-extrusion process. The paper describes the tube fabrication process, the development of a finish sanding procedure and the results of high temperature thermal cycle testing. The high temperature thermal cycle testing generates peak strains on the metal surfaces, where the surface effects have a maximum influence on the fatigue strength. 9 figs., 2 tabs

Electron microscopy of nuclear zirconium alloys

International Nuclear Information System (INIS)

Versaci, R.A.; Ipohorski, Miguel

1986-01-01

Transmission electron microscopy observations of the microstructure of zirconium alloys used in fuel sheaths of nuclear power reactors are reported. Specimens were observed after different thermal and mechanical treatment, similar to those actually used during fabrication of the sheaths. Electron micrographs and electron diffraction patterns of second phase particles present in zircaloy-2 and zircaloy-4 were also obtained, as well as some characteristic parameters. Images of oxides and hydrides most commonly present in zirconium alloys are also shown. Finally, the structure of a Zr-2,5Nb alloy used in CANDU reactors pressure tubes, is observed by electron microscopy. (Author) [es

Stress corrosion cracking susceptibility of steam generator tube materials in AVT (all volatile treatment) chemistry contaminated with lead

International Nuclear Information System (INIS)

Gomez Briceno, D.; Castano, M.L.; Garcia, M.S.

1996-01-01

Alloy 600 steam generator tubing has shown a high susceptibility to stress corrosion degradation at the operation conditions of pressurized water reactors. Several contaminants, such as lead, have been postulated as being responsible for producing the secondary side stress corrosion cracking that has occurred mainly at the location where these contaminants can concentrate. An extensive experimental work has been carried out in order to better understand the effects of lead on the stress corrosion cracking susceptibility of steam generator tube materials, namely Alloys 600, 690 and 800. This paper presents the experimental work conducted with a view to determining the influence of lead oxide concentration in AVT (all volatile treatment) conditions on the stress corrosion resistance of nickel alloys used in the fabrication of steam generator tubing. (orig.)

PWSCC Potential on the Drain Tubes of WEC Model 51F Steam Generators in Domestic Plants

International Nuclear Information System (INIS)

Chang, B. I.; Song, M. S.; Yoon, K. S.; Kim, I. Y.

2009-01-01

The drain couplings of Westinghouse Model 51F steam generators that had been operated in the domestic Nuclear Power Plants were modified and repaired in accordance with the corrective action of replacing the existing Alloy 600 weld build-up with Alloy 690(Alloy 52/152) weld build-up in order to increase its resistance to primary water stress corrosion cracking (PWSCC). The drain tube made of alloy 600, however, was not replaced and left susceptible to the PWSCC. Among the environmental, metallurgical and mechanical factors controlling a susceptibility to the PWSCC, it is believed that tensile stresses play an important role. The objective of this study is to conservatively estimate stress state of the drain tube during fabrication and when exposed to normal plant operations, and to investigate its potential for the PWSCC

An experimental study of ECT for fin-type copper alloy tubes

International Nuclear Information System (INIS)

Lee, Hyung Joon; Lee, Jeong Soon; Sung, Je Joong; Park, Cheon Woong; Suh, Dong Man; Yu, Taek In

2002-01-01

Eddy current detecting probes with inner and encircling coils were designed for the fin-type tubes that have uneven outer and inner surface to enhance the efficiency of heat emission. As the uneven surface of them, it is difficult to detect flaws in the tubes by eddy current test. In this paper, standard and artificial specimens with flaws for the different types of the tubes were manufactured. Eddy current test was performed with the designed probes, which have inner and encircling coils, for the prepared specimens. From the signals of the eddy current detecting probes, the phase and amplitude variation were analyzed and the best conditions of the flaw detection for the tubes were found.

SAR analysis of a needle type applicator made from a shape memory alloy using 3-D anatomical human head model

International Nuclear Information System (INIS)

Kubo, Mitsunori; Mimoto, Naoki; Hirashima, Taku; Morita, Emi; Shindo, Yasuhiro; Kato, Kazuo; Takahashi, Hideaki; Uzuka, Takeo; Fujii, Yukihiko

2009-01-01

This paper describes the possibility of a new heating method with a needle applicator made of a shape memory alloy (SMA) to expand the heating area for interstitial brain tumor hyperthermia treatments. The purpose of the study described here is to show the capability of the method to expand a defined heating region with the developed three-dimensional (3-D) anatomical human head model using the finite element method (FEM). One major disadvantage of radiofrequency (RF) interstitial hyperthermia treatment is that this heating method has a small heating area. To overcome this problem, a new type of needle made of a SMA was developed. The specific absorption rate (SAR) distributions of this proposed method, when applied to the 3-D anatomical human head model reconstructed from two-dimensional (2-D) MRI and X-ray CT images, were calculated with computer simulations. The calculated SAR distributions showed no unexpected hot spots within the model. The heated area was localized around the tumor. These results suggest that the proposed heating method using the SMA needle applicator and the developed method for reconstructing a 3-D anatomical human head model are capable of being used for invasive brain tumor hyperthermia treatments. (author)

Method of processing spent fuel cladding tubes

International Nuclear Information System (INIS)

Nakatsuka, Masafumi; Ouchi, Atsuhiro; Imahashi, Hiromichi.

1986-01-01

Purpose: To decrease the residual activity of spent fuel cladding tubes in a short period of time and enable safety storage with simple storage equipments. Constitution: Spent fuel cladding tubes made of zirconium alloys discharged from a nuclear fuel reprocessing step are exposed to a grain boundary embrittling atmosphere to cause grain boundary destruction. This causes grain boundary fractures to the zirconium crystal grains as the matrix of nuclear fuels and then precipitation products precipitated to the grain boundary fractures are removed. The zirconium constituting the nuclear fuel cladding tube and other ingredient elements contained in the precipitation products are separated in this removing step and they are separately stored respectively. As a result, zirconium constituting most part of the composition of the spent nuclear fuel cladding tubes can be stored safely at a low activity level. (Takahashi, M.)

Secondary side IGA/IGSCC of SG alloys 600, 690 and 800 : R and D program in EDF Laboratories

International Nuclear Information System (INIS)

Vaillant, F.; DeBouvier, O.; Bouchacourt, M.; Stutzmann, A.; Lemaire, P.

1998-01-01

Many steam generators (SGs) equipped with 'mill-annealed' (MA) Alloy 600 tubings suffer significant secondary side corrosion. Until now, no degradation has been observed with either Alloy 600 TT or Alloy 690 for new SGs. The understanding of IGA/SCC of Alloy 600 MA in plants and the development of predictive models have become an important challenge to assess the life span and to reduce the maintenance costs of SGs. As degradation occurs in crevice environments which are varied and little known, EDF has undertaken an important program to improve the knowledge of crevice environments which lead to cracking. Corrosion tests are performed on Alloys 600 MA (also on 600 TT) in various environments in order to reproduce the deposits and the cracking observed on pulled tubes in laboratory conditions. Other corrosion tests are conducted in environments containing some pollutants identified by analyses of secondary water after hideout-return (sulfates) or oxidizing compounds : the influences of pH and potential are evaluated on Alloy 600 (MA or TT) and also on Alloys 690 and 800. A comprehensive model is proposed using IGA/SCC results of Alloy 600 in caustic environments. The thermomechanical parameters of the tubes and the field environmental conditions, introduced in the model, confirm some important features of SGs tubings. The model will be improved to include other detrimental environments. It will provide a useful tool to predict the life span (then steam generator replacements) and to optimize the maintenance policy of SGs still equipped with Alloys 600 MA and particularly with 600 TT (frequency and best locations of inspections). Margins will also be assessed for new SGs equipped with Alloy 690, and a comparison will be performed with Alloy 800. (author)

Characterization of Friction Stir Welded Tubes by Means of Tube Bulge Test

International Nuclear Information System (INIS)

D'Urso, G.; Longo, M.; Giardini, C.

2011-01-01

Mechanical properties of friction stir welded joints are generally evaluated by means of conventional tensile test. This testing method might provide insufficient information because maximum strain obtained in tensile test before necking is small; moreover, the application of tensile test is limited when the joint path is not linear or even when the welds are executed on curved surfaces. Therefore, in some cases, it would be preferable to obtain the joints properties from other testing methods. Tube bulge test can be a valid solution for testing circumferential or longitudinal welds executed on tubular workpieces. The present work investigates the mechanical properties and the formability of friction stir welded tubes by means of tube bulge tests. The experimental campaign was performed on tubular specimens having a thickness of 3 mm and an external diameter of 40 mm, obtained starting from two semi-tubes longitudinally friction stir welded. The first step, regarding the fabrication of tubes, was performed combining a conventional forming process and friction stir welding. Sheets in Al-Mg-Si-Cu alloy AA6060 T6 were adopted for this purpose. Plates having a dimension of 225x60 mm were bent (with a bending axis parallel to the main dimension) in order to obtain semi-tubes. A particular care was devoted to the fabrication of forming devices (punch and die) in order to minimize the springback effects. Semi-tubes were then friction stir welded by means of a CNC machine tool. Some preliminary tests were carried out by varying the welding parameters, namely feed rate and rotational speed. A very simple tool having flat shoulder and cylindrical pin was used. The second step of the research was based on testing the welded tubes by means of tube bulge test. A specific equipment having axial actuators with a conical shape was adopted for this study. Some analyses were carried out on the tubes bulged up to a certain pressure level. In particular, the burst pressure and the

Gastrostomy versus nasogastric tube feeding for chemoradiation patients with head and neck cancer: the TUBE pilot RCT.

Science.gov (United States)

Paleri, Vinidh; Patterson, Joanne; Rousseau, Nikki; Moloney, Eoin; Craig, Dawn; Tzelis, Dimitrios; Wilkinson, Nina; Franks, Jeremy; Hynes, Ann Marie; Heaven, Ben; Hamilton, David; Guerrero-Urbano, Teresa; Donnelly, Rachael; Barclay, Stewart; Rapley, Tim; Stocken, Deborah

2018-04-01

Approximately 9000 new cases of head and neck squamous cell cancers (HNSCCs) are treated by the NHS each year. Chemoradiation therapy (CRT) is a commonly used treatment for advanced HNSCC. Approximately 90% of patients undergoing CRT require nutritional support via gastrostomy or nasogastric tube feeding. Long-term dysphagia following CRT is a primary concern for patients. The effect of enteral feeding routes on swallowing function is not well understood, and the two feeding methods have, to date (at the time of writing), not been compared. The aim of this pilot randomised controlled trial (RCT) was to compare these two options. This was a mixed-methods multicentre study to establish the feasibility of a RCT comparing oral feeding plus pre-treatment gastrostomy with oral feeding plus as-required nasogastric tube feeding in patients with HNSCC. Patients were recruited from four tertiary centres treating cancer and randomised to the two arms of the study (using a 1 : 1 ratio). The eligibility criteria were patients with advanced-staged HNSCC who were suitable for primary CRT with curative intent and who presented with no swallowing problems. The primary outcome was the willingness to be randomised. A qualitative process evaluation was conducted alongside an economic modelling exercise. The criteria for progression to a Phase III trial were based on a hypothesised recruitment rate of at least 50%, collection of outcome measures in at least 80% of those recruited and an economic value-of-information analysis for cost-effectiveness. Of the 75 patients approached about the trial, only 17 consented to be randomised [0.23, 95% confidence interval (CI) 0.13 to 0.32]. Among those who were randomised, the compliance rate was high (0.94, 95% CI 0.83 to 1.05). Retention rates were high at completion of treatment (0.94, 95% CI 0.83 to 1.05), at the 3-month follow-up (0.88, 95% CI 0.73 to 1.04) and at the 6-month follow-up (0.88, 95% CI 0.73 to 1.04). No serious adverse

A simplified technique for nasoendotracheal tube immobilization.

OpenAIRE

Berardo, N.; Leban, S. G.; Williams, F. A.

1989-01-01

A simplified technique for immobilization of a nasoendotracheal tube is described in which a wide strap of open cell, hypoallergenic, foam-backed fabric is secured to the patient's head with a Velcro fastener.

Titanium oxide dispersion-strengthened ferritic alloys

International Nuclear Information System (INIS)

Hendrix, W.; Vandermeulen, W.

1980-04-01

The available data on the DT02 and DT3911 ferritic dispersion strengthened alloys, developed at SCK/CEN, Mol, Belgium, are presented. Both alloys consist of Fe - 13% Cr - 1.5% Mo to which 2% TiO 2 and about 3.5% Ti are added (wt.%). Their main use is for the fabrication of fast breeder reactor cladding tubes but their application as turbine blade material is also envisaged for cases where high damping is important. (auth.)

Characterization of mechanical properties of hydroxyapatite-silicon-multi walled carbon nano tubes composite coatings synthesized by EPD on NiTi alloys for biomedical application.

Science.gov (United States)

Khalili, Vida; Khalil-Allafi, Jafar; Sengstock, Christina; Motemani, Yahya; Paulsen, Alexander; Frenzel, Jan; Eggeler, Gunther; Köller, Manfred

2016-06-01

Release of Ni(1+) ions from NiTi alloy into tissue environment, biological response on the surface of NiTi and the allergic reaction of atopic people towards Ni are challengeable issues for biomedical application. In this study, composite coatings of hydroxyapatite-silicon multi walled carbon nano-tubes with 20wt% Silicon and 1wt% multi walled carbon nano-tubes of HA were deposited on a NiTi substrate using electrophoretic methods. The SEM images of coated samples exhibit a continuous and compact morphology for hydroxyapatite-silicon and hydroxyapatite-silicon-multi walled carbon nano-tubes coatings. Nano-indentation analysis on different locations of coatings represents the highest elastic modulus (45.8GPa) for HA-Si-MWCNTs which is between the elastic modulus of NiTi substrate (66.5GPa) and bone tissue (≈30GPa). This results in decrease of stress gradient on coating-substrate-bone interfaces during performance. The results of nano-scratch analysis show the highest critical distance of delamination (2.5mm) and normal load before failure (837mN) as well as highest critical contact pressure for hydroxyapatite-silicon-multi walled carbon nano-tubes coating. The cell culture results show that human mesenchymal stem cells are able to adhere and proliferate on the pure hydroxyapatite and composite coatings. The presence of both silicon and multi walled carbon nano-tubes (CS3) in the hydroxyapatite coating induce more adherence of viable human mesenchymal stem cells in contrast to the HA coated samples with only silicon (CS2). These results make hydroxyapatite-silicon-multi walled carbon nano-tubes a promising composite coating for future bone implant application. Copyright © 2016 Elsevier Ltd. All rights reserved.

Influence of compressive load conditions and thickness on the two-way shape memory behavior in tube-shaped NiTi alloy

International Nuclear Information System (INIS)

Yoo, Young Ik; Shin, Dong Kil; Lee, Jung Ju; Lee, Chang Ho

2012-01-01

The two-way shape memory behavior of Ni 55 Ti 45 was investigated to develop a tube-shaped NiTi actuator which could generate a large amount of force. The two-way shape memory effect (TWSME) was induced by thermal cycling under various amounts of constant compressive stress. Six specimens with the same outer diameter and different thickness were used to apply the TWSME to an actuator. A fast saturation tendency of the recovery strain was shown through training at each level of constant stress, after which the two-way shape memory strain was quantitatively measured during thermal cycling for each level of applied stress. From the results, the maximum two-way strain value was obtained after training at a constant level of stress and then decreased thereafter. In addition, the two-way strain was found to depend on the thickness of the tube-shaped specimen. All specimens could be divided into two groups depending on the rate of increase in the two-way strain. After two-way strain was obtained, the two-way recovery stress was measured to verify the performance of the sample as an actuator. The results showed that the two-way recovery stress behavior was similar to the two-way strain; if the optimal thickness of the specimen and the stress applied for training are used for the development of the TWSME, tube-shaped NiTi using the TWSME can replace one-way shape memory alloys. (paper)

Manufacturing process optimization of nuclear fuel guide tube using HANA alloys

International Nuclear Information System (INIS)

Jeong, Yong Hwan; Park, S. Y.; Choi, B. K.; Park, J. Y.; Kim, H. G.; Jeong, Y. I.; Park, D. J.; Lim, J. K.

2010-08-01

From this project, the advanced manufacturing parameters which were contained of heat-treatment, reduction rate, and new process (2 step) were considered to increase the guide tube performance of HANA material. It was obtained that the strength and corrosion resistance of HANA material were improved by applying the improve manufacturing parameters when compared to the commercial guide tube material. · Manufacturing parameter study to increase mechanical property -Tensile strength increase of 30% by manufacturing parameter setup when compared to the guide tube specification · Manufacturing parameter study to decrease irradiation growth -Theoretical study of the texture effect on sample specimens related to the irradiation growth · Manufacturing parameter study to increase corrosion resistance -Corrosion resistance increase of 30% by manufacturing parameter setup when compared to the commercial guide tube

Characterisation of Oxides Formed on the Internal Surface of Steam Generator Tubes in Alloy 690 Corroded in the Primary Environment of Pressurised Water Reactors

International Nuclear Information System (INIS)

Carrette, Florence; Leclercq, Stephanie; Legras, Laurent

2012-09-01

Since the end of the 1990s, EDF R and D has been studying the phenomenon of corrosion product release from Steam Generator tubes in order to minimize the Source Term of the contamination and radiation exposure during operation and maintenance of Pressurised Water Reactors. With the BOREAL loop, release tests in primary water at 325 deg. C were performed on various Steam Generator tubes made of alloy 690. The experimental conditions of these tests (chemistry, temperature and hydraulics) were the same for all the tests but the results showed various behaviours towards release. For some tubes, the release was weak whereas for others, it was higher; the release rate of the tubes decreased more or less quickly with time. In order to explain these results, the internal surface of the tubes was characterised before and after the tests. Before the tests, various parameters were studied; the main parameters were the roughness, the impurities, the grain size and the cold work. The results demonstrated that it was not easy to quantify the influence of each parameter on release and to differentiate the tubes. A new parameter was proposed to characterise the internal extreme surface of SG tubes: the surface nano-hardness by nano-indentation measurements. The tubes were also observed and analysed by SEM, (X)TEM. Data obtained by (X)TEM revealed differences of the surface state (layer of perturbed microstructure, density of dislocations, grain size, impurities, initial oxide,...). After the tests, the oxides formed on the internal surface and the underlying material of the samples were characterised by SEM, (X)TEM and SIMS. The examinations showed various types of oxides. For some tubes, a duplex oxide scale was identified, for the others, only one oxide scale was observed. For equivalent durations of corrosion, the thickness of the enriched - chromium oxide layer can vary from 5 nm to 100 nm and the chemical composition can be different. The examinations of the underlying

Porcine head response to blast.

Science.gov (United States)

Shridharani, Jay K; Wood, Garrett W; Panzer, Matthew B; Capehart, Bruce P; Nyein, Michelle K; Radovitzky, Raul A; Bass, Cameron R 'dale'

2012-01-01

Recent studies have shown an increase in the frequency of traumatic brain injuries related to blast exposure. However, the mechanisms that cause blast neurotrauma are unknown. Blast neurotrauma research using computational models has been one method to elucidate that response of the brain in blast, and to identify possible mechanical correlates of injury. However, model validation against experimental data is required to ensure that the model output is representative of in vivo biomechanical response. This study exposes porcine subjects to primary blast overpressures generated using a compressed-gas shock tube. Shock tube blasts were directed to the unprotected head of each animal while the lungs and thorax were protected using ballistic protective vests similar to those employed in theater. The test conditions ranged from 110 to 740 kPa peak incident overpressure with scaled durations from 1.3 to 6.9 ms and correspond approximately with a 50% injury risk for brain bleeding and apnea in a ferret model scaled to porcine exposure. Instrumentation was placed on the porcine head to measure bulk acceleration, pressure at the surface of the head, and pressure inside the cranial cavity. Immediately after the blast, 5 of the 20 animals tested were apneic. Three subjects recovered without intervention within 30 s and the remaining two recovered within 8 min following respiratory assistance and administration of the respiratory stimulant doxapram. Gross examination of the brain revealed no indication of bleeding. Intracranial pressures ranged from 80 to 390 kPa as a result of the blast and were notably lower than the shock tube reflected pressures of 300-2830 kPa, indicating pressure attenuation by the skull up to a factor of 8.4. Peak head accelerations were measured from 385 to 3845 G's and were well correlated with peak incident overpressure (R(2) = 0.90). One SD corridors for the surface pressure, intracranial pressure (ICP), and head acceleration are

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

Specific Adaptation of Gas Atomization Processing for Al-Based Alloy Powder for Additive Manufacturing

Energy Technology Data Exchange (ETDEWEB)

Anderson, Iver [Ames Lab., Ames, IA (United States); Siemon, John [Alcoa, Inc, Pittsburgh, PA (United States)

2017-06-30

The initial three atomization attempts resulted in “freeze-outs” within the pour tubes in the pilot-scale system and yielded no powder. Re-evaluation of the alloy liquidus temperatures and melting characteristics, in collaboration with Alcoa, showed further superheat to be necessary to allow the liquid metal to flow through the pour tube to the atomization nozzle. A subsequent smaller run on the experimental atomization system verified these parameters and was successful, as were all successive runs on the larger pilot scale system. One alloy composition froze-out part way through the atomization on both pilot scale runs. SEM images showed needle formation and phase segregations within the microstructure. Analysis of the pour tube freeze-out microstructures showed that large needles formed within the pour tube during the atomization experiment, which eventually blocked the melt stream. Alcoa verified the needle formation in this alloy using theoretical modeling of phase solidification. Sufficient powder of this composition was still generated to allow powder characterization and additive manufacturing trials at Alcoa.

Hydrogen concentration determination in pressure tube samples using differential scanning calorimetry (dsc)

International Nuclear Information System (INIS)

Marinescu, R.; Mincu, M.

2015-01-01

Zirconium alloys are widely used as a structural material in nuclear reactors. It is known that zirconium based cladding alloys absorb hydrogen as a result of service in a pressurized water reactor. Hydrogen absorbed (during operation of the reactor) in the zirconium alloy, out of which the pressure tube is made, is one of the major factors determining the life time of the pressure tube. For monitoring the hydrides, samples of the pressure tube are periodically taken and analyzed. At normal reactor operating temperature, hydrogen has limited solubility in the zirconium lattice and precipitates out of solid solution as zirconium hydride when the solid solubility is exceeded. As a consequences material characterization of Zr-2.5Nb CANDU pressure tubes is required after manufacturing but also during the operation to assess its structural integrity and to predict its behavior until the next in-service inspection. Hydrogen and deuterium concentration determination is one of the most important parameters to be evaluated during the experimental tests. Hydrogen present in zirconium alloys has a strong effect of weakening. Following the zirconium-hydrogen reaction, the resulting zirconium hydride precipitates in the mass of material. Weakening of the material, due to the presence of 10 ppm of precipitated hydrogen significantly affects some of its properties. The concentration of hydrogen in a sample can be determined by several methods, one of them being the differential scanning calorimetry (DSC). The principle of the method consists in measuring the difference between the amount of heat required to raise the temperature of a sample and a reference to a certain value. The experiments were made using a TA Instruments DSC Q2000 calorimeter. This paper contains experimental work for hydrogen concentration determination by Differential Scanning Calorimetry (DSC) method. Also, the reproducibility and accuracy of the method used at INR Pitesti are presented. (authors)

Creep deformation behaviour and microstructural changes in Zr-2.5% Nb alloy

International Nuclear Information System (INIS)

Chaudhuri, S.; Singh, R.; Ghosh, R.N.; Sinha, T.K.; Banerjee, S.

2002-01-01

Cold worked and stress relieved Zr-2.5% Nb alloy is a well-known material used as pressure tubes in Pressurised Heavy Water Reactors. The pressure tubes, made of a typical Zr-alloy, consisting of 2.54% Nb, 0.1175% oxygen and less than 100 ppm impurities, are expected to withstand 9.5 MPa to 12.5 MPa pressure at 250 degC to 310 degC under fast neutron fluxes of 3.5 x 10 17 nm -2 s -1 . These tubes are made by hot extrusion at 780 degC with an extrusion ratio 8.3:1 and 40% cold pilgering followed by annealing at 550 degC for 3 hours and subsequently by 20-30% cold pilgering and stress relieving at 400 degC for 24 hours. The microstructure of such cold worked and stress relieved alloy consists of Β-Zr precipitates in the matrix of elongated Α-Zr grains. Although various factors such as irradiation creep, thermal creep, irradiation growth etc are responsible for limiting the life of pressure tubes; the thermal creep contributes significantly in overall creep deformation. Keeping this in view as well as due to non-availability of adequate published information including creep database on this alloy, an extensive investigation on the thermal creep behaviour of indigenously produced Zr-2.5% Nb alloy was undertaken. The creep tests in air using Mayes' creep testing machines were carried out in the temperature range of 300 degC to 450 degC under stresses in the range of 50 to 550 MPa. Analysis of data revealed that the mechanism of creep deformation remains the same in this range

Development of laser cladding system to repair wall thinning of 1-inch heat exchanger tube

International Nuclear Information System (INIS)

Terada, Takaya

2013-01-01

We developed a laser cladding system to repair the inner wall wastage of heat exchanger tubes. Our system, which is designed to repair thinning tube walls within 100 mm from the edge of a heat exchanger tube, consists of a fiber laser, a composite-type optical fiberscope, a coupling device, a laser processing head, and a wire-feeding device. All of these components were reconfigured from the technologies of FBR maintenance. The laser processing head, which has a 15-mm outer diameter, was designed to be inserted into a 1-inch heat exchanger tube. We mounted a heatproof broadband mirror for laser cladding and fiberscope observation with visible light inside the laser processing head. The wire-feeding device continuously supplied 0.4-mm wire to the laser irradiation spot with variable feeding speeds from 0.5 to 20 mm/s. We are planning to apply our proposed system to the maintenance of aging industrial plants. (author)

CT detection of occult pneumothorax in head trauma

International Nuclear Information System (INIS)

Tocino, I.M.; Miller, M.H.; Frederick, P.R.; Bahr, A.L.; Thomas, F.

1984-01-01

A prospective evaluation for occult pneumothorax was performed in 25 consecutive patients with serious head trauma by combining a limited chest CT examination with the emergency head CT examination. Of 21 pneuomothoraces present in 15 patients, 11 (52%) were found only by chest CT and were not identified clinically or by supine chest radiograph. Because of pending therapeutic measures, chest tubes were placed in nine of the 11 occult pneumothoraces, regardless of the volume. Chest CT proved itself as the most sensitive method for detection of occult pneumothorax, permitting early chest tube placement to prevent transition to a tension pneumothorax during subsequent mechanical ventilation or emergency surgery under general anesthesia

CT detection of occult pneumothorax in head trauma

Energy Technology Data Exchange (ETDEWEB)

Tocino, I.M.; Miller, M.H.; Frederick, P.R.; Bahr, A.L.; Thomas, F.

1984-11-01

A prospective evaluation for occult pneumothorax was performed in 25 consecutive patients with serious head trauma by combining a limited chest CT examination with the emergency head CT examination. Of 21 pneuomothoraces present in 15 patients, 11 (52%) were found only by chest CT and were not identified clinically or by supine chest radiograph. Because of pending therapeutic measures, chest tubes were placed in nine of the 11 occult pneumothoraces, regardless of the volume. Chest CT proved itself as the most sensitive method for detection of occult pneumothorax, permitting early chest tube placement to prevent transition to a tension pneumothorax during subsequent mechanical ventilation or emergency surgery under general anesthesia.

A novel Fe–Cr–Nb matrix composite containing the TiB{sub 2} neutron absorber synthesized by mechanical alloying and final hot isostatic pressing (HIP) in the Ti-tubing

Energy Technology Data Exchange (ETDEWEB)

Litwa, Przemysław [Department of Advanced Materials and Technologies, Military University of Technology, Kaliskiego 2, 00-908 Warsaw (Poland); Perkowski, Krzysztof [Department of Nanotechnology, Institute of Ceramics and Building Materials, Postępu 9, 02-676 Warsaw (Poland); Zasada, Dariusz [Department of Advanced Materials and Technologies, Military University of Technology, Kaliskiego 2, 00-908 Warsaw (Poland); Kobus, Izabela; Konopka, Gustaw [Department of Nanotechnology, Institute of Ceramics and Building Materials, Postępu 9, 02-676 Warsaw (Poland); Czujko, Tomasz [Department of Advanced Materials and Technologies, Military University of Technology, Kaliskiego 2, 00-908 Warsaw (Poland); Varin, Robert A., E-mail: robert.varin@uwaterloo.ca [Department of Mechanical and Mechatronics Engineering, University of Waterloo, 200 University Ave., Waterloo, ON N2L 3G1 (Canada)

2016-07-25

The Fe–Cr–Ti-Nb elemental powders were mechanically alloyed/ball milled with TiB{sub 2} and a small quantity of Y{sub 2}O{sub 3} ceramic to synthesize a novel Fe-based alloy-ceramic powder composite that could be processed by hot isostatic pressing (HIP) for a perceived potential application as a neutron absorber in nuclear reactors. After ball milling for the 30–80 h duration relatively uniform powders with micrometric sizes were produced. With increasing milling time a fraction of TiB{sub 2} particles became covered with the much softer Fe-based alloy which resulted in the formation of a characteristic “core-mantel” structure. For the final HIP-ing process the mechanically alloyed powders were initially uniaxially pressed into rod-shaped compacts and then cold isostatically pressed (CIP-ed). Subsequently, the rod-shaped compacts were placed in the Ti-tubing and subjected to hot isostatic pressing (HIP) at 1150 °C/200 MPa pressure. The HIP-ing process resulted in the formation of the near-Ti and intermediate diffusional layers in the microstructure of HIP-ed samples which formed in accord with the Fe-Ti binary phase diagram. Those layers contain the phases such as α-Ti (HCP), the FeTi intermetallic and their hypo-eutectoid mixtures. In addition, needle-like particles were formed in both layers in accord with the Ti-B binary phase diagram. Nanohardness testing, using a Berkovich type diamond tip, shows that the nanohardness in the intermediate layer areas, corresponding to the composition of the hypo-eutectoid mixture of Ti-FeTi, equals 980.0 (±27.1) HV and correspondingly 1176.9 (±47.6) HV for the FeTi phase. The nanohardness in the sample's center in the areas with the fine mixture of Fe-based alloy and small TiB{sub 2} particles equals 1048.3 (±201.8) HV. The average microhardness of samples HIP-ed from powders milled for 30 and 80 h is 588 HV and 733 HV, respectively. - Highlights: • A Fe–Cr–Nb-based composite with TiB{sub 2} neutron

Transfer of metallic debris from the metal surface of an acetabular cup to artificial femoral heads by scraping: comparison between alumina and cobalt-chrome heads.

Science.gov (United States)

Chang, Chong Bum; Yoo, Jeong Joon; Song, Won Seok; Kim, Deug Joong; Koo, Kyung-Hoi; Kim, Hee Joong

2008-04-01

We aimed to investigate the transfer of metal to both ceramic (alumina) and metal (cobalt-chrome) heads that were scraped by a titanium alloy surface under different load conditions. The ceramic and metal heads for total hip arthroplasties were scraped by an acetabular metal shell under various loads using a creep tester. Microstructural changes in the scraped area were visualized with a scanning electron microscope, and chemical element changes were assessed using an energy dispersive X-ray spectrometry. Changes in the roughness of the scraped surface were evaluated by a three-dimensional surface profiling system. Metal transfer to the ceramic and metal heads began to be detectable at a 10 kg load, which could be exerted by one-handed force. The surface roughness values significantly increased with increasing test loads in both heads. When the contact force increased, scratching of the head surface occurred in addition to the transfer of metal. The results documented that metallic debris was transferred from the titanium alloy acetabular shell to both ceramic and metal heads by minor scraping. This study suggests that the greatest possible effort should be made to protect femoral heads, regardless of material, from contact with metallic surfaces during total hip arthroplasty.

Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations

Science.gov (United States)

Aydogan, E.; Weaver, J. S.; Maloy, S. A.; El-Atwani, O.; Wang, Y. Q.; Mara, N. A.

2018-05-01

FeCrAl ferritic alloys are excellent cladding candidates for accident tolerant fuel systems due to their high resistance to oxidation as a result of formation of a protective Al2O3 scale at high temperatures in steam. In this study, we report the irradiation response of the 10Cr and 13Cr FeCrAl cladding tubes under Fe2+ ion irradiation up to ∼16 dpa at 300 °C. Dislocation loop size, density and characteristics were determined using both two-beam bright field transmission electron microscopy and on-zone scanning transmission electron microscopy techniques. 10Cr (C06M2) tube has a lower dislocation density, larger grain size and a slightly weaker texture compared to the 13Cr (C36M3) tube before irradiation. After irradiation to 0.7 dpa and 16 dpa, the fraction of type sessile dislocations decreases with increasing Cr amount in the alloys. It has been found that there is neither void formation nor α‧ precipitation as a result of ion irradiations in either alloy. Therefore, dislocation loops were determined to be the only irradiation induced defects contributing to the hardening. Nanoindentation testing before the irradiation revealed that the average nanohardness of the C36M3 tube is higher than that of the C06M2 tube. The average nanohardness of irradiated tube samples saturated at 1.6-2.0 GPa hardening for both tubes between ∼3.4 dpa and ∼16 dpa. The hardening calculated based on transmission electron microscopy was found to be consistent with nanohardness measurements.

The corrosion Characteristics and Behaviors of the Ti-2.19Al-2.35Zr alloy

International Nuclear Information System (INIS)

Kim, Tae Hoon; Kang, Chang Sun; Baek, Jong Hyuk; Kim, Hyun Gil; Choi, Byoung Kwon; Jeong, Yong Hwan

2007-01-01

Ti-2.19Al-2.35Zr alloy is being considered as a steam generator tube material for the advanced pressurized water reactor (PWR) which is being developed by KAERI for the purpose of seawater desalination as well as a small scale electricity production. The main operational environment of SMART differs somewhat from that of a commercial PWR. That is, a heat-exchange tube is always exposed to a high temperature/pressure condition and an ammonia water chemistry is designed as a pH controlling agent without an addition of boric acid. The excellent mechanical and corrosion resistance properties are required for the steam generator tube material in SMART. Thus Ti-2.19Al-2.35Zr alloy was studied to investigate of the corrosion characteristics and behaviors of the Ti- 2.19Al-2.35Zr alloy in a simulated-SMART loop

Method for the protection of the cladding tubes of fuel rods

International Nuclear Information System (INIS)

Steinberg, E.

1978-01-01

To present stress crack corrosion and to protect the cladding tubes of the fuel rods made of a circonium alloy from attack by iodine, the inward surfaces are provided with protective coatings. Therefore the casting tubes already filled with fuel element pellets are put under over-pressure at a temperature range between 300 and 500 0 C, until almost yield-point is reached. A small amount of H 2 O or H 2 O 2 , filled in, reacts with the cladding tube material to form the Zr-O 2 protective coating. Afterwards comes a pressure relief, and the cladding tube reaches its original dimensions. (DG) [de

Multifrequency Eddy current testing of heat exchange tubes with a rotating probe

International Nuclear Information System (INIS)

Levy, R.

1982-01-01

Multi-frequency eddy current analyses have been used in France industrially since 1975. In light of the experienced gained during many steam generator inspections, this technique was applied to the examination of sheet and tube heat exchangers featuring tubes in very different materials such as copper, stainless steel and titanium. The principle of multi-frequency Eddy current inspection is first reviewed, using the example of a condenser with nickel alloy tubes (Inconel, Incoloy). This is followed by the description of a specific application of this technique to a condenser with titanium tubes, analyzed with a rotating local probe [fr

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

International Nuclear Information System (INIS)

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

2000-09-01

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

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

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

2000-09-01

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

PWSCC in the tube expansion zone - an overview

International Nuclear Information System (INIS)

Hernalsteen, P.

1993-01-01

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

IGA/SCC propagation rate measurements on alloy 600 steam generator tubing using a side stream model boiler

International Nuclear Information System (INIS)

Takamatsu, H.; Matsueda, K.; Matsunaga, T.; Kitera, T.; Arioka, K.; Tsuruta, T.; Okamoto, S.

1993-01-01

IGA/SCC crack propagation rate measurements using various types of IGA/SCC predefected ALloy 600 tubing were tested in model boilers, a side stream model boiler at Ohi Unit 1 and similar model boilers in the laboratory. Types of IGA/SCC predefects introduced from the outside of the tubing were as follows. (1) Actual IGA/SCC predefect introduced by high temperature caustic environments; (2) Longitudinal predefect by electrodischarge machining (EDM) method, and then crack tip fatigue was introduced to serve as the marker on the fractured surface (EDM slit + fatigue). IGA/SCC crack propagation rate was measured after the destructive examination by Cr concentration profile on fracture surface for (1), and observation of intergranular fractured surface propagated from the marked fatigue was employed for (2) and (3) after the model boiler tests. As for the water chemistry conditions, mainly AVT (high N 2 H 4 ) + boric acid (5-10ppm as B in SGs) treatment for both model boilers, and some of the tests for the model boiler in the laboratory employed AVT (high N 2 H 4 ) without boric acid. The results of IGA/SCC crack propagation rate measurements were compared with each other, and the three methods employed showed a good coincidence with the rate of ca. 1 x 10 -5 mm/Hr for AVT (high N 2 H 4 ) + boric acid treatment condition, in the case that crack tip boron intensity (B/O value by IMMA analysis) of more than 1 was observed

Effect of chemical composition on corrosion resistance of Zircaloy fuel cladding tube for BWR

International Nuclear Information System (INIS)

Inagaki, Masahisa; Akahori, Kimihiko; Kuniya, Jirou; Masaoka, Isao; Suwa, Masateru; Maru, Akira; Yasuda, Teturou; Maki, Hideo.

1990-01-01

Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510degC) steam and a high temperature (288degC) water. In addition, four 450 kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance. Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30 [Ni]+0.15[Fe]≥0.045 (w%) showed no susceptibility to nodular corrosion. An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25 w% and Ni≤0.1 w% did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530degC steam test. (author)

Thermal storage/discharge performances of Cu-Si alloy for solar thermochemical process

Science.gov (United States)

Gokon, Nobuyuki; Yamaguchi, Tomoya; Cho, Hyun-seok; Bellan, Selvan; Hatamachi, Tsuyoshi; Kodama, Tatsuya

2017-06-01

The present authors (Niigata University, Japan) have developed a tubular reactor system using novel "double-walled" reactor/receiver tubes with carbonate molten-salt thermal storage as a phase change material (PCM) for solar reforming of natural gas and with Al-Si alloy thermal storage as a PCM for solar air receiver to produce high-temperature air. For both of the cases, the high heat capacity and large latent heat (heat of solidification) of the PCM phase circumvents the rapid temperature change of the reactor/receiver tubes at high temperatures under variable and uncontinuous characteristics of solar radiation. In this study, we examined cyclic properties of thermal storage/discharge for Cu-Si alloy in air stream in order to evaluate a potentiality of Cu-Si alloy as a PCM thermal storage material. Temperature-increasing performances of Cu-Si alloy are measured during thermal storage (or heat-charge) mode and during cooling (or heat-discharge) mode. A oxidation state of the Cu-Si alloy after the cyclic reaction was evaluated by using electron probe micro analyzer (EPMA).

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

The evaluation of the angles of Eustachian tubes in the patients with ...

African Journals Online (AJOL)

2015-12-02

Dec 2, 2015 ... plane is considered to be the position at which the head in a neutral anatomical ..... Proctor B. Embryology and anatomy of the eustachian tube. Arch Otolaryngol. 1967 ... Otolaryngol Head Neck Surg. 1986;94:78‑81. 22.

Cytokines levels, Severity of acute mucositis and the need of PEG tube installation during chemo-radiation for head and neck cancer - a prospective pilot study

International Nuclear Information System (INIS)

Meirovitz, Amichay; Kuten, Michal; Billan, Salem; Abdah-Bortnyak, Roxolyana; Sharon, Anat; Peretz, Tamar; Sela, Mordechai; Schaffer, Moshe; Barak, Vivian

2010-01-01

The purpose of this pilot study was to detect a correlation between serum cytokine levels and severity of mucositis, necessitating installation of a percutaneous endoscopic gastrostomy tube (PEG) in head and neck (H&N) cancer patients receiving combined chemo-radiation therapy. Fifteen patients with H&N epithelial cancer were recruited to this study. All patients received radiotherapy to the H&N region, with doses ranging from 50-70 Gy. Chemotherapy with cisplatin, carboplatin, 5-fluorouracil and taxanes was given to high-risk patients, using standard chemotherapy protocols. Patients were evaluated for mucositis according to WHO common toxicity criteria, and blood samples were drawn for inflammatory (IL-1, IL-6, IL-8, TNF-α) and anti-inflammatory (IL-10) cytokine levels before and during treatment. A positive correlation was found between IL-6 serum levels and severity of mucositis and dysphagia; specifically, high IL-6 levels at week 2 were correlated with a need for PEG tube installation. A seemingly contradictory correlation was found between low IL-8 serum levels and a need for a PEG tube. These preliminary results, indicating a correlation between IL-6 and IL-8 serum levels and severity of mucositis and a need for a PEG tube installation, justify a large scale study

Compressive Creep Behaviour of Extruded Mg Alloys at 150 °C

Science.gov (United States)

Fletcher, M.; Bichler, L.; Sediako, D.; Klassen, R.

Wrought magnesium alloy bars, sections and tubes have been extensively used in the aerospace, electronics and automotive industries, where component weight is of concern. The operating temperature of these components is typically limited to below 100°C, since appreciable creep relaxation of the wrought alloys takes place above this temperature.

Manufacture of seamless stainless steel tubings and related equipment

International Nuclear Information System (INIS)

Wali, D.K.; Chaudhary, S.

1997-01-01

Production of seamless tubes for special application is one of the important production activities of Nuclear Fuel Complex (NFC), Hyderabad. NFC had set up facility of Hot Extrusion Press and Cold Pilger Mills with related finishing and inspection equipment for manufacturing quality seamless tubes of zirconium alloy for application in nuclear power reactors in early 70''s. Being aware that the demand for seamless tube in a developing economy gradually increases till it reaches around 30 to 40% of the total requirement of tubes and pipes and also of the fact that manufacturing technology developed for production of zircaloy seamless tubes for nuclear application, can easily be harnessed and spinned off for production of seamless tubes in materials generally difficult to hot roll (in other than extrusion process), NFC augmented its seamless tube manufacturing facility by adding, a vertical piercing press, series of induction furnaces and large size pilger mills to meet existing market demand of power sector, engineering, fertilisers and petro chemical industries and any other specialised applications

Objective and subjective image quality of primary and recurrent squamous cell carcinoma on head and neck low-tube-voltage 80-kVp computed tomography

Energy Technology Data Exchange (ETDEWEB)

Scholtz, Jan-Erik; Kaup, Moritz; Kraft, Johannes; Noeske, Eva-Maria; Schulz, Boris; Burck, Iris; Kerl, J.M.; Bauer, Ralf W.; Lehnert, Thomas; Vogl, Thomas J.; Wichmann, Julian L. [University Hospital Frankfurt, Department of Diagnostic and Interventional Radiology, Frankfurt (Germany); Scheerer, Friedrich [University Hospital Frankfurt, Department of Cranio-Maxillofacial and Plastic Facial Surgery, Frankfurt (Germany); Wagenblast, Jens [University Hospital Frankfurt, Department of Otolaryngology, Head and Neck Surgery, Frankfurt (Germany)

2015-03-26

To investigate low-tube-voltage 80-kVp computed tomography (CT) of head and neck primary and recurrent squamous cell carcinoma (SCC) regarding objective and subjective image quality. We retrospectively evaluated 65 patients (47 male, 18 female; mean age: 62.1 years) who underwent head and neck dual-energy CT (DECT) due to biopsy-proven primary (n = 50) or recurrent (n = 15) SCC. Eighty peak kilovoltage and standard blended 120-kVp images were compared. Attenuation and noise of malignancy and various soft tissue structures were measured. Tumor signal-to-noise ratio (SNR) and contrast-to-noise ratio (CNR) were calculated. Subjective image quality was rated by three reviewers using 5-point grading scales regarding overall image quality, lesion delineation, image sharpness, and image noise. Radiation dose was assessed as CT dose index volume (CTDI{sub vol}). Interobserver agreement was calculated using intraclass correlation coefficient (ICC). Mean tumor attenuation (153.8 Hounsfield unit (HU) vs. 97.1 HU), SNR (10.7 vs. 8.3), CNR (8.1 vs. 4.8), and subjective tumor delineation (score, 4.46 vs. 4.13) were significantly increased (all P < 0.001) with 80-kVp acquisition compared to standard blended 120-kVp images. Noise of all measured structures was increased in 80-kVp acquisition (P < 0.001). Overall interobserver agreement was good (ICC, 0.86; 95 % confidence intervals: 0.82-0.89). CTDI{sub vol} was reduced by 48.7 % with 80-kVp acquisition compared to standard DECT (4.85 ± 0.51 vs. 9.94 ± 0.81 mGy cm, P < 0.001). Head and neck CT with low-tube-voltage 80-kVp acquisition provides increased tumor delineation, SNR, and CNR for CT imaging of primary and recurrent SCC compared to standard 120-kVp acquisition with an accompanying significant reduction of radiation exposure. (orig.)

Recovery of aluminium, nickel-copper alloys and salts from spent fluorescent lamps.

Science.gov (United States)

Rabah, Mahmoud A

2004-01-01

This study explores a combined pyro-hydrometallurgical method to recover pure aluminium, nickel-copper alloy(s), and some valuable salts from spent fluorescent lamps (SFLs). It also examines the safe recycling of clean glass tubes for the fluorescent lamp industry. Spent lamps were decapped under water containing 35% acetone to achieve safe capture of mercury vapour. Cleaned glass tubes, if broken, were cut using a rotating diamond disc to a standard shorter length. Aluminium and copper-nickel alloys in the separated metallic parts were recovered using suitable flux to decrease metal losses going to slag. Operation variables affecting the quality of the products and the extent of recovery with the suggested method were investigated. Results revealed that total loss in the glass tube recycling operation was 2% of the SFLs. Pure aluminium meeting standard specification DIN 1712 was recovered by melting at 800 degrees C under sodium chloride/carbon flux for 20 min. Standard nickel-copper alloys with less than 0.1% tin were prepared by melting at 1250 degrees C using a sodium borate/carbon flux. De-tinning of the molten nickel-copper alloy was carried out using oxygen gas. Tin in the slag as oxide was recovered by reduction using carbon or hydrogen gas at 650-700 degrees C. Different valuable chloride salts were also obtained in good quality. Further research is recommended on the thermodynamics of nickel-copper recovery, yttrium and europium recovery, and process economics.

Recovery of aluminium, nickel-copper alloys and salts from spent fluorescent lamps

International Nuclear Information System (INIS)

Rabah, Mahmoud A.

2004-01-01

This study explores a combined pyro-hydrometallurgical method to recover pure aluminium, nickel-copper alloy(s), and some valuable salts from spent fluorescent lamps (SFLs). It also examines the safe recycling of clean glass tubes for the fluorescent lamp industry. Spent lamps were decapped under water containing 35% acetone to achieve safe capture of mercury vapour. Cleaned glass tubes, if broken, were cut using a rotating diamond disc to a standard shorter length. Aluminium and copper-nickel alloys in the separated metallic parts were recovered using suitable flux to decrease metal losses going to slag. Operation variables affecting the quality of the products and the extent of recovery with the suggested method were investigated. Results revealed that total loss in the glass tube recycling operation was 2% of the SFLs. Pure aluminium meeting standard specification DIN 1712 was recovered by melting at 800 deg. C under sodium chloride/carbon flux for 20 min. Standard nickel-copper alloys with less than 0.1% tin were prepared by melting at 1250 deg. C using a sodium borate/carbon flux. De-tinning of the molten nickel-copper alloy was carried out using oxygen gas. Tin in the slag as oxide was recovered by reduction using carbon or hydrogen gas at 650-700 deg. C. Different valuable chloride salts were also obtained in good quality. Further research is recommended on the thermodynamics of nickel-copper recovery, yttrium and europium recovery, and process economics

Postoperative Airway Emergency following Accidental Flexometallic Tube Transection

Directory of Open Access Journals (Sweden)

Karim Habib MR

2015-10-01

Full Text Available Endotracheal intubation using flexometallic tubes are often required in anaesthesia practice for a variety of reasons. It is preferred in the head and neck region surgeries due to its relative resistance to kinking forces. At times, these patients postoperatively may need to be shifted to ICU or HDU without extubation for further stabilization/management and extubation after adequate recovery. We present an unusual accident where a new flexometallic endotracheal tube was permanently tapered, transected and migrated proximally due to patient’s bite on tube leading to airway emergency in post-operative recovery period.

Development of delayed hydride cracking resistant-pressure tube

Energy Technology Data Exchange (ETDEWEB)

Kim, Young Suk; Kwon, Sang Chul; Kim, S. S.; Yim, K. S

2000-10-01

For the first time, we demonstrate that the pattern of nucleation and growth of a DHC crack is governed by the precipitation of hydrides so that the DHC velocity and K{sub IH} are determined by an angle of the cracking plane and the hydride habit plane 10.7. Since texture controls the distribution of the 10.7 habit plane in Zr-2.5Nb pressure tube, we draw a conclusion that a textural change in Zr-2.5Nb tube from a strong tangential texture to the radial texture shall increase the threshold stress intensity factor, K{sub IH}, and decrease the delayed hydride cracking velocity. This conclusion is also verified by a complimentary experiment showing a linear dependence of DHCV and K{sub IH} with an increase in the basal component in the cracking plane. On the basis of the study on the DHC mechanism and the effect of manufacturing processes on the properties of Zr-2.5Nb tube, we have established a manufacturing procedure to make pressure tubes with improved DHC resistance. The main features of the established manufacturing process consist in the two step-cold pilgering process and the intermediate heat treatment in the {alpha} + {beta} phase for Zr-2.5Nb alloy and in the {alpha} phase for Zr-1Nb-1.2Sn-0.4Fe alloy. The manufacturing of DHC resistant-pressure tubes of Zr-2.5Nb and Zr-1N-1.2Sn-0.4Fe was made in the ChMP zirconium plant in Russia under a joint research with Drs. Nikulina and Markelov in VNIINM (Russia). Zr-2.5Nb pressure tube made with the established manufacturing process has met all the specification requirements put by KAERI. Chracterization tests have been jointly conducted by VNIINM and KAERI. As expected, the Zr-2.5Nb tube made with the established procedure has improved DHC resistance compared to that of CANDU Zr-2.5Nb pressure tube used currently. The measured DHC velocity of the Zr-2.5Nb tube meets the target value (DHCV <5x10{sup -8} m/s) and its other properties also were equivalent to those of the CANDU Zr-2.5Nb tube used currently. The Zr-1Nb-1

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

International Nuclear Information System (INIS)

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

1988-01-01

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

Computational fluid dynamic analysis of a closure head penetration in a pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Forsyth, D.R.; Schwirian, R.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

1995-09-01

ALLOY 600 has been used typically for penetrations through the closure head in pressurized water reactors because of its thermal compatibility with carbon steel, superior resistance to chloride attack and higher strength than the austenitic stainless steels. Recent plant operating experience with this alloy has indicated that this material may be susceptible to degradation. One of the major parameters relating to degradation of the head penetrations are the operational temperatures and stress levels in the penetration.

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

Directory of Open Access Journals (Sweden)

Mohammed Al Hajri

2015-04-01

Full Text Available Dissimilar metal weld (DMW joint between alloyed steel (AS and stainless steel (SS failed at one of intermediate temperature superheater (ITSH tube in steam/power generation plant boiler. The premature failure was detected after a relatively short time of operation (8 years where the crack propagated circumferentially from AS side through the ITSH tube. Apart from physical examination, microstructural studies based on optical microscopy, SEM and EDX analysis were performed. The results of the investigation point out the limitation of Carbides precipitation at the alloyed steel/welding interface. This is synonym of creep stage I involvement in the failure of ITSH. Improper post-welding operation and bending moment are considered as root causes of the premature failure.

X-ray target with substrate of molybdenum alloy

International Nuclear Information System (INIS)

Hirsch, H.H.

1980-01-01

Rotary targets for x-ray tubes are provided comprising a molybdenum base body alloyed with a stabilizing proportion of iron, silicon, cobalt, tantalum, niobium, hafnium, stable metal oxide, or a mixture of the preceding

The French regulatory experience and views on nickel-base alloy PWSCC prevention and treatment

Energy Technology Data Exchange (ETDEWEB)

Turluer, G.; Cattiaux, G.; Monnot, B. [Institut de Radioprotection et de Surete Nucleaire, IRSN, 92 - Fontenay aux Roses (France); Emond, D.; Reuchet, J.; Chartier, Ph. [Direction Generale de la Surete Nucleaire et de la Radioprotection, 75 - Paris (France)

2003-10-01

This paper presents the experience feedback and views of the French Regulatory Authority (ASN) and of the technical support institute (IRSN) on PWSCC prevention since the initiation in 1989 of the 'Inconel Zones Review' requested by ASN to Electricite de France (EDF), the national operator of a fleet of 58 PWRs. This proactive requirement, launched before the discovery, in September 1991, of the only CRDM nozzle leak in France, on Bugey unit 3, was then triggered by the recurrence of many alloy 600 rapid degradations and leaks, world wide, and also in France in the late 1980's, particularly on steam generator tubes and on some pressurizer penetrations. Thus, the ASN requested that EDF, perform a comprehensive (generic) proactive assessment on all the nickel-base alloy components and parts of the main primary circuits, which of course included vessel head penetrations and bottom vessel head penetrations, and some other zones as a first priority. This proactive 'review' did, a minima, include the following tasks and actions: - Update and complete, by an extensive R and D program, the understanding and characterization of the Ni base alloys prone to PWSCC, - Analyze the various materials, metallurgical features, mechanical stresses, and physicochemical conditions of the parts exposed to primary water, in order to predict the occurrence of PWSCC initiation and propagation, - Provide a prioritization of the zones to be inspected, - Implement by improved NDE techniques a practical inspection program on the 58 PWRs, - Prepare and implement any needed mitigation actions as a result of the components conditions assessment. The present paper relates the main features of the French regulatory experience over more than 13 years and recalls the main principles of the assessment, which were applied by ASN. These principles, which are formalized in the current regulation rules revised in 1999, are briefly listed hereunder: - It is based on avoiding and

The French regulatory experience and views on nickel-base alloy PWSCC prevention and treatment

International Nuclear Information System (INIS)

Turluer, G.; Cattiaux, G.; Monnot, B.; Emond, D.; Reuchet, J.; Chartier, Ph.

2003-10-01

This paper presents the experience feedback and views of the French Regulatory Authority (ASN) and of the technical support institute (IRSN) on PWSCC prevention since the initiation in 1989 of the 'Inconel Zones Review' requested by ASN to Electricite de France (EDF), the national operator of a fleet of 58 PWRs. This proactive requirement, launched before the discovery, in September 1991, of the only CRDM nozzle leak in France, on Bugey unit 3, was then triggered by the recurrence of many alloy 600 rapid degradations and leaks, world wide, and also in France in the late 1980's, particularly on steam generator tubes and on some pressurizer penetrations. Thus, the ASN requested that EDF, perform a comprehensive (generic) proactive assessment on all the nickel-base alloy components and parts of the main primary circuits, which of course included vessel head penetrations and bottom vessel head penetrations, and some other zones as a first priority. This proactive 'review' did, a minima, include the following tasks and actions: - Update and complete, by an extensive R and D program, the understanding and characterization of the Ni base alloys prone to PWSCC, - Analyze the various materials, metallurgical features, mechanical stresses, and physicochemical conditions of the parts exposed to primary water, in order to predict the occurrence of PWSCC initiation and propagation, - Provide a prioritization of the zones to be inspected, - Implement by improved NDE techniques a practical inspection program on the 58 PWRs, - Prepare and implement any needed mitigation actions as a result of the components conditions assessment. The present paper relates the main features of the French regulatory experience over more than 13 years and recalls the main principles of the assessment, which were applied by ASN. These principles, which are formalized in the current regulation rules revised in 1999, are briefly listed hereunder: - It is based on avoiding and preventing any leaking on

Creep-Data Analysis of Alloy 617 for High Temperature Reactor Intermediate Heat Exchanger

International Nuclear Information System (INIS)

Kim, Woo Gon; Ryu, Woo Seog; Kim, Yong Wan; Yin, Song Nan

2006-01-01

The design of the metallic components such as hot gas ducts, intermediate heat exchanger (IHX) tube, and steam reformer tubes of very high temperature reactor (VHTR) is principally determined by the creep properties, because an integrity of the components should be preserved during a design life over 30 year life at the maximum operating temperature up to 1000 .deg. C. For designing the time dependent creep of the components, a material database is needed, and an allowable design stress at temperature should be determined by using the material database. Alloy 617, a nicked based superalloy with chromium, molybdenum and cobalt additions, is considered as a prospective candidate material for the IHX because it has the highest design temperature. The alloy 617 is approved to 982 .deg. C (1800 .deg. F) and other alloys approved to 898 .deg. C (1650 .deg. C), such as alloy 556, alloy 230, alloy HX, alloy 800. Also, the alloy 617 exhibits the highest level of creep strength at high temperatures. Therefore, it is needed to collect the creep data for the alloy 617 and the creep-rupture life at the given conditions of temperature and stress should be predicted for the IHX construction. In this paper, the creep data for the alloy 617 was collected through literature survey. Using the collected data, the creep life for the alloy 617 was predicted based on the Larson-Miller parameter. Creep master curves with standard deviations were presented for a safety design, and failure probability for the alloy 617 was obtained with a time coefficient

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)

Evaluation of two styles of slotted, flat-head screws

International Nuclear Information System (INIS)

Reeves, C.A. Jr.; Johnson, W.B.

1979-01-01

A series of torque tests were performed to evaluate the relative merits of two different flat-head screws fabricated from a uranium--6% niobium alloy. The screws tested were machined with both normal, straight-through slots in the head and with slots having radiused bottoms. Test results indicate that both designs easily surpass the required 20-inch-pound-proof torque

PWSCC issues and material aging management for nuclear power plants in Korea

Energy Technology Data Exchange (ETDEWEB)

Hwang, Seong Sik; Lim, Yun Soo; Kim, Dong Jin; Kim, Sung Woo; Kim, Hong Pyo [Nuclear Materials Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-10-15

The primary water stress corrosion cracking (PWSCC) of alloy 600 in a PWR has been reported in the control rod drive mechanism (CRDM). Beginning in the mid-seventies, the pressurized water reactor (PWR) plants suffered from a sequence of SCC events mostly confined to S/G tubes, initially ODSCC, and then PWSCC. PWSCC was first reported in Bugey 3 vessel head penetration made of forged alloy 600 materials in September 1991. Other PWRs experienced cracking attributed to the PWSCC of the major primary side weld area made from alloy 182 at the end of the year 2000. Examples of dissimilar metal butt welds between the main austenitic stainless steel primary circuit piping and the outlet pressure vessel nozzles are the cracking of Ringhals 4, V. C. Summer and some J-groove welds of the CRDM of the RVH at Oconee 1. In addition to the Reactor Vessel Head (RVH), the PWSCC of alloy 182/82 has been reported at bottom mounted instrumentation (BMI) nozzle J-welds, steam generator(SG) J-weld drain nozzle, and SG tube sheet cladding. Two cases of boric acid precipitation were reported at the bottom head surface of a SG in Korea. Cracking was found in the cold leg drain nozzles made of alloy 600 in two units, hot side nozzles were fabricated with alloy 690 from the beginning. The cracking of steam generator tubings made of alloy 600 is another concern in Korea, because some plants still have alloy 600 HTMA tubings. The flow accelerated corrosion of secondary pipings is another type of corrosion problems, though it has not been treated as a severe problem in Korea. To properly manage the corrosion issues and seek out research items for maintaining the integrity of nuclear plants, the PRIMA-Net (Proactive Research and Innovative Material Aging Network) was organized in 2007. The research and development expert group consists of a National research laboratory (KAERI), regulatory body (KINS), utility (KHNP), engineering and design company (KEPCO EC), manufacturer (Doosan Heavy

An evaluation of corrosion resistant alloys by field corrosion test in Japanese refuse incineration plants

International Nuclear Information System (INIS)

Kawahara, Yuuzou; Nakamura, Masanori; Shibuya, Eiichi; Yukawa, Kenichi

1995-01-01

As the first step for development of the corrosion resistant superheater tube materials of 500 C, 100 ata used in high efficient waste-to-energy plants, field corrosion tests of six conventional alloys were carried out at metal temperatures of 450 C and 550 C for 700 and 3,000 hours in four typical Japanese waste incineration plants. The test results indicate that austenitic alloys containing approximately 80 wt% [Cr+Ni] show excellent corrosion resistance. When the corrosive environment is severe, intergranular corrosion of 40∼200 microm depth occurs in stainless steel and high alloyed materials. It is confirmed quantitatively that corrosion behavior is influenced by environmental corrosion factors such as Cl concentration and thickness of deposits on tube surface, metal temperature, and flue gas temperature. The excellent corrosion resistance of high [Cr+Ni+Mo] alloys such as Alloy 625 is explained by the stability of its protective oxide, such that the time dependence of corrosion nearly obeys the parabolic rate law

Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun-Gil, E-mail: hgkim@kaeri.re.kr; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jeong-Yong; Koo, Yang-Hyun

2015-10-15

A 3D laser coating technology using Cr powder was developed for Zr-based alloys considering parameters such as: the laser beam power, inert gas flow, cooling of Zr-based alloys, and Cr powder control. This technology was then applied to Zr cladding tube samples to study the effect of Cr coating on the high-temperature oxidation of Zr-based alloys in a steam environment of 1200 °C for 2000s. It was revealed that the oxide layer thickness formed on the Cr-coated tube surface was about 25-times lower than that formed on a Zircaloy-4 tube surface. In addition, both the ring compression and the tensile tests were performed to evaluate the adhesion properties of the Cr-coated sample. Although some cracks were formed on the Cr-coated layer, the Cr-coated layer had not peeled off after the two tests.

The "Air in the CT X-ray Tube Oil" Artifact-Examples of the Quality Control Images and the Evaluation of Four Potential Clinical Patients' Head Computed Tomography Cases.

Science.gov (United States)

Törmänen, Juhani; Rautiainen, Jari; Tahvonen, Pirita; Leinonen, Kimmo; Nieminen, Miika T; Tervonen, Osmo

We present a newly reposted scanner-based artifact-with 4 potential patients' head computed tomography (CT) cases-the "Air in the CT X-ray Tube Oil" artifact with a 64-slice multidetector CT. This artifact mimics diseases, which cause hypodense findings in CT images. It can be difficult to notice in the clinical patient imaging but can be also very difficult to verify in quality control tests.

Estimation of improved productivity based on materials substitution in high temperature applications. Use of alloy ASTM A-335 P91

International Nuclear Information System (INIS)

Serna, J A; Afanador, W

2001-01-01

In ECOPETROL-ICP was carried out an evaluation of the mechanical and micro structural properties of modified 9 Cr-1 Mo alloy, ASTM A-335 Gr. P91, finding higher strength mechanical properties, allowable stresses and creep rupture strength, than the conventional 9 Cr-1 Mo alloy, ASTM A-335 Gr. P9, recommending the alloy P91 as substitute tube material in the radiation zone of the Visbreaking heater of Cartagena's refinery (furnace in revamping process). The results obtained permit a thickness reduction of radiation tubes of material P91 close to 25% and increase the internal volume tube over up 8%, which is a parameter to consider in improving productivity and efficiency process. Also would be obtained a significant savings cost in the material among 5 and 10%. Additionally, expectations of both design and remaining useful life would be seen extensively favored with this change of alloy

Optimum Frequency for Eddy Current Testing Method of SMART SG tubes

International Nuclear Information System (INIS)

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

2009-01-01

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

Microstructural characterization and mechanical properties of Excel alloy pressure tube material

Science.gov (United States)

Sattari, Mohammad

Microstructural characterization and mechanical properties of Excel (Zr-3.5%Sn-0.8%Mo-0.8%Nb), a dual phase alphaZr -hcp and betaZr-bcc pressure tube material, is discussed in the current study which is presented in manuscript format. Chapter 3 discusses phase transformation temperatures using different techniques such as quantitative metallography, differential scanning calorimetry (DSC), and electrical resistivity. It was found that the alphaZr → alphaZr+beta Zr and alphaZr+betaZr → betaZr transformation temperatures are in the range of 600-690°C and 960-970°C respectively. Also it was observed that upon quenching from temperatures below ˜860°C the martensitic transformation of betaZr to alpha'--hcp is halted and instead the microstructure transforms into retained Zr with o hexagonal precipitates inside betaZr grains. Chapter 4 deals with aging response of Excel alloy. Precipitation hardening was observed in samples water-quenched from high in the alphaZr+beta Zr or betaZr regions followed by aging. The optimum aging conditions were found to be 450°C for 1 hour. Transmission electron microscopy (TEM) showed dispersion of fine precipitates (˜10nm) inside the martensitic phase. Energy dispersive X-ray spectroscopy (EDS) showed the chemical composition of precipitates to be Zr-30wt%Mo-25wt%Nb-2wt%Fe. Electron crystallography using whole pattern symmetry of the convergent beam electron diffraction (CBED) patterns together with selected area diffraction (SAD) polycrystalline ring patterns, suggests the -6m2 point group for the precipitates belonging to hexagonal crystal structure, with a= 2.936 A and c=4.481 A, i.e. c/a =1.526. Crystallographic texture and high temperature tensile properties as well as creep-rupture properties of different microstructures are discussed in Chapter 5. Texture analysis showed that solution treatment high in the alpha Zr+betaZr or betaZr regions followed by water quenching or air cooling results in a more random texture compared

Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

International Nuclear Information System (INIS)

1998-08-01

The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

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.

The performance of alloy 625 in the high temperature application of Heavy Water Plants

International Nuclear Information System (INIS)

Mitra, J.; Dey, G.K.; Sundararaman, M.; Dubey, J.S.; De, P.K.; Kumar, Niraj

2006-01-01

Wrought and centrifugally cast alloy 625 tubes are used in the cracker units of ammonia based Heavy Water Plants (HWP). During the service of about 100,000 h, the ammonia cracker tubes, predictably, have been exposed to temperatures below 600degC to above 765degC and have undergone several hundreds of start-shutdown cycles, producing several ordered phases in the alloy. To understand the effect of the ordered phases on the structure properties, Alloy 625 samples were aged at 540degC, 700degC and 850degC temperatures, for duration up to 1200 h. Results were compared with that of cast and wrought Alloy 625 samples, which aged during the service of 100,000 h and that failed during the service after about 24,000 h along with that of aged samples, which were resolutionised at 1170degC for 2h. (author)

Improvement of hydro-turbine draft tube efficiency using vortex generator

Directory of Open Access Journals (Sweden)

Xiaoqing Tian

2015-07-01

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

Study of low vibration 4 K pulse tube cryocoolers

Science.gov (United States)

Xu, Mingyao; Nakano, Kyosuke; Saito, Motokazu; Takayama, Hirokazu; Tsuchiya, Akihiro; Maruyama, Hiroki

2012-06-01

Sumitomo Heavy Industries, Ltd. (SHI) has been continuously improving the efficiency and reducing the vibration of a 4 K pulse tube cryocooler. One advantage of a pulse tube cryocooler over a GM cryocooler is low vibration. In order to reduce vibration, both the displacement and the acceleration have to be reduced. The vibration acceleration can be reduced by splitting the valve unit from the cold head. One simple way to reduce vibration displacement is to increase the wall thickness of the tubes on the cylinder. However, heat conduction loss increases while the wall thickness increases. To overcome this dilemma, a novel concept, a tube with non-uniform wall thickness, is proposed. Theoretical analysis of this concept, and the measured vibration results of an SHI lowvibration pulse tube cryocooler, will be introduced in this paper.

PWSCC Growth Assessment Model Considering Stress Triaxiality Factor for Primary Alloy 600 Components

Directory of Open Access Journals (Sweden)

Jong-Sung Kim

2016-08-01

Full Text Available We propose a primary water stress corrosion cracking (PWSCC initiation model of Alloy 600 that considers the stress triaxiality factor to apply to finite element analysis. We investigated the correlation between stress triaxiality effects and PWSCC growth behavior in cold-worked Alloy 600 stream generator tubes, and identified an additional stress triaxiality factor that can be added to Garud's PWSCC initiation model. By applying the proposed PWSCC initiation model considering the stress triaxiality factor, PWSCC growth simulations based on the macroscopic phenomenological damage mechanics approach were carried out on the PWSCC growth tests of various cold-worked Alloy 600 steam generator tubes and compact tension specimens. As a result, PWSCC growth behavior results from the finite element prediction are in good agreement with the experimental results.

Dysphagia after sequential chemoradiation therapy for advanced head and neck cancer.

Science.gov (United States)

Goguen, Laura A; Posner, Marshall R; Norris, Charles M; Tishler, Roy B; Wirth, Lori J; Annino, Donald J; Gagne, Adele; Sullivan, Christopher A; Sammartino, Daniel E; Haddad, Robert I

2006-06-01

Assess impact of sequential chemoradiation therapy (SCRT) for advanced head and neck cancer (HNCA) on swallowing, nutrition, and quality of life. Prospective cohort study of 59 patients undergoing SCRT for advanced head and neck cancer. Follow-up median was 47.5 months. Regional Cancer Center. Median time to gastrostomy tube removal was 21 weeks. Eighteen of 23 patients who underwent modified barium swallow demonstrated aspiration; none developed pneumonia. Six of 7 with pharyngoesophageal stricture underwent successful dilatation. Functional Assessment of Cancer Therapy-Head and Neck Scale questionnaires at median 6 months after treatment revealed "somewhat" satisfaction with swallowing. At the time of analysis, 97% have the gastronomy tube removed and take soft/regular diet. Early after treatment dysphagia adversely affected weight, modified barium swallow results, and quality of life. Diligent swallow therapy, and dilation as needed, allowed nearly all patients to have their gastronomy tubes removed and return to a soft/regular diet. Dysphagia is significant after SCRT but generally slowly recovers 6 to 12 months after SCRT. C-4.

Delayed hydrogen cracking test design for pressure tubes

International Nuclear Information System (INIS)

Haddad, Roberto; Loberse, Antonio N.; Yawny, Alejandro A.; Riquelme, Pablo

1999-01-01

CANDU nuclear power stations pressure tubes of alloy Zr-2,5 % Nb present a cracking phenomenon known as delayed hydrogen cracking (DHC). This is a brittle fracture of zirconium hydrides that are developed by hydrogen due to aqueous corrosion on the metal surface. This hydrogen diffuses to the crack tip where brittle zirconium hydrides develops and promotes the crack propagation. A direct current potential decay (DCPD) technique has been developed to measure crack propagation rates on compact test (CT) samples machined from a non irradiated pressure tube. Those test samples were hydrogen charged by cathodic polarization in an acid solution and then pre cracked in a fatigue machine. This technique proved to be useful to measure crack propagation rates with at least 1% accuracy for DHC in pressure tubes. (author)

Nickel electroplating as a remedy to steam generator tubing PWSCC

International Nuclear Information System (INIS)

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

1993-01-01

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

Chloride removal from plutonium alloy

International Nuclear Information System (INIS)

Holcomb, H.P.

1983-01-01

SRP is evaluating a program to recover plutonium from a metallic alloy that will contain chloride salt impurities. Removal of chloride to sufficiently low levels to prevent damaging corrosion to canyon equipment is feasible as a head-end step following dissolution. Silver nitrate and mercurous nitrate were each successfully used in laboratory tests to remove chloride from simulated alloy dissolver solution containing plutonium. Levels less than 10 ppM chloride were achieved in the supernates over the precipitated and centrifuged insoluble salts. Also, less than 0.05% loss of plutonium in the +3, +4, or +6 oxidation states was incurred via precipitate carrying. These results provide impetus for further study and development of a plant-scale process to recover plutonium from metal alloy at SRP

Galvanic Corrosion between Alloy 690 and Magnetite in Alkaline Aqueous Solutions

Directory of Open Access Journals (Sweden)

Soon-Hyeok Jeon

2015-12-01

Full Text Available The galvanic corrosion behavior of Alloy 690 coupled with magnetite has been investigated in an alkaline solution at 30 °C and 60 °C using a potentiodynamic polarization method and a zero resistance ammeter. The positive current values were recorded in the galvanic couple and the corrosion potential of Alloy 690 was relatively lower. These results indicate that Alloy 690 behaves as the anode of the pair. The galvanic coupling between Alloy 690 and magnetite increased the corrosion rate of Alloy 690. The temperature increase led to an increase in the extent of galvanic effect and a decrease in the stability of passive film. Galvanic effect between Alloy 690 and magnetite is proposed as an additional factor accelerating the corrosion rate of Alloy 690 steam generator tubing in secondary water.

Pulse tube coolers for Meteosat third generation

International Nuclear Information System (INIS)

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

2014-01-01

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

Increasing Weldability of Service-Aged Reformer Tubes by Partial Solution Annealing

Science.gov (United States)

Mostafaei, M.; Shamanian, M.; Purmohamad, H.; Amini, M.

2016-04-01

A dissimilar joint of 25Cr-35Ni/30Cr-48Ni (HP/HV) heat-resistant steels was evaluated. The investigations indicated that the as-cast HP alloy contained M7C3, M23C6, and NbC carbides and HV alloy with 5 wt.% tungsten, contained M23C6 and M6C carbides embedded in an austenitic matrix. After 8 years of ex-service aging at 1050 °C, the ductility of HP/HV reformer tubes was decreased dramatically, and thus, the repair welding of the aged HP/HV dissimilar joint was at a risk. In order to repair the aged reformer tubes and increase weldability properties, a new partial solution annealing treatment was designed. Mechanical testing results showed that partial solution annealing at 1200 °C for 6 h increased the elongation and toughness of the aged HP and HV alloys drastically. Also, a mechanism for constitutional liquation cracking in the heat-affected zones (HAZ) of the HP/HV dissimilar joint was proposed. In the HAZ of the aged HP/HV welded joint, the cracks around the locally melted carbides were initiated and propagated during carbides solidification at the cooling cycle of welding associated with the decrease in the ductility of the aged HP and HV alloys. In addition, Varestraint weldability test showed that the susceptibility to hot cracking was decreased with partial solution annealing.

Comparison of DHC behaviour of two zirconium alloys

International Nuclear Information System (INIS)

Ponzoni, Lucio; Mieza, Ignacio; Heras, Evangelina De Las; Domizzi, Gladys

2011-01-01

Delayed hydride cracking (DHC) is an important cracking mechanism that may occur in Zr alloys during service in water-cooled reactors. Two conditions must be attained to initiate DHC: the stress intensity factor must be higher than a threshold value called K IH and hydrogen concentration must exceed a critical value. Currently the pressure tubes for CANDU reactor are fabricated from Zr-2.5Nb, but another Zr-alloy, Excel was evaluated demonstrating similar values of K IH but higher DHC velocity. In this paper the critical hydrogen concentration of Excel tube was evaluated and compared with that of Zr-2.5Nb. Due to higher hydrogen solubility limits in Excel, its critical concentration for DHC initiation is 10-40 wppm over that of Zr-2.5Nb in the range of 150 to 300 deg C. (author)

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

International Nuclear Information System (INIS)

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

2007-08-01

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

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

International Nuclear Information System (INIS)

Delabrouille, F.

2004-11-01

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

Influence of thermal buoyancy on vertical tube bundle thermal density head predictions under transient conditions

International Nuclear Information System (INIS)

Lin, H.C.; Kasza, K.E.

1984-01-01

The thermal-hydraulic behavior of an LMFBR system under various types of plant transients is usually studied using one-dimensional (1-D) flow and energy transport models of the system components. Many of the transient events involve the change from a high to a low flow with an accompanying change in temperature of the fluid passing through the components which can be conductive to significant thermal bouyancy forces. Thermal bouyancy can exert its influence on system dynamic energy transport predictions through alterations of flow and thermal distributions which in turn can influence decay heat removal, system-response time constants, heat transport between primary and secondary systems, and thermal energy rejection at the reactor heat sink, i.e., the steam generator. In this paper the results from a comparison of a 1-D model prediction and experimental data for vertical tube bundle overall thermal density head and outlet temperature under transient conditions causing varying degrees of thermal bouyancy are presented. These comparisons are being used to generate insight into how, when, and to what degree thermal buoyancy can cause departures from 1-D model predictions

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

OpenAIRE

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

2015-01-01

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

Inconel type resistive alloys based on ultrahigh purity nickel

International Nuclear Information System (INIS)

Matsarin, K.A.; Matsarin, S.K.

2000-01-01

The new nickel high-ohm alloys (ρ = 1.2-1.4 μOhm · m), containing the W, Al, Mo alloying elements in the quantity, not exceeding their solubility in a solid solution, are developed on the basis of the Inconel-type standard alloy. The optical composition of the alloy was determined by the results of the alloy was determined by the results of the electric resistance measurement and technological effectiveness indices (relative to the pressure and workable metal yield). The following optimal component concentrations were established: 14-17 %Cr; 10-12 %Fe; 0.5-1.0 %Cu; 1.0-1.5 %Mn; 0.1-0.2 %C; 0.4-0.6 %Si; 0.5-3.0 %W; 5-16 %Mo; 0.5-2.0 %Al; the remainder - Ni. The new alloys are recommended as materials for resistive elements of direct-glow cathode nodes of low capacity electron tubes [ru

Leak behaviors of steam generator tube-to-tubesheet joints from room temperature to 320 °C

International Nuclear Information System (INIS)

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

2013-01-01

To address concerns about excessive leakage from throughwall cracks in nuclear reactor tube-to-tubesheet joints under accident conditions, leak rates were measured experimentally by using tube-to-collar joint specimens and nitrogen gas. Rates were dependent on differential pressure between the tube internal surface and the crevice (i.e., the tube-to-collar interface region) and on temperature. As specimen temperature was raised to 320 °C, leak rates decreased gradually due to changes in gas properties and to differential thermal expansion between the Alloy 600 tubes and the SA508 collars. The leak rates did not change even after repeated temperature excursions to 320 °C, suggesting that thermally induced creep and subsequent contact pressure relaxation is negligible below that temperature. When considering factors that could increase flow resistance, such as oxidation, or debris on top of the tubesheet, the measured leak rates in this work are considered to be conservative. The test results were further used to validate the contact pressure calculation and a leak rate model. Highlights: ► Leak rates were measured by using tube-to-collar joint specimens. ► Leak rates were dependent on differential pressure between tube internal and joint interface. ► Leak rates decreased gradually as specimen temperature was raided to 320 °C. ► Differential thermal expansion between Alloy 600 tube and SA508 collar plays a major role on the leak behavior.

Novel Zn-based alloys for biodegradable stent applications: Design, development and in vitro degradation.

Science.gov (United States)

Mostaed, E; Sikora-Jasinska, M; Mostaed, A; Loffredo, S; Demir, A G; Previtali, B; Mantovani, D; Beanland, R; Vedani, M

2016-07-01

The search for a degradable metal simultaneously showing mechanical properties equal or higher to that of stainless steel and uniform degradation is still an open challenge. Several magnesium-based alloys have been studied, but their degradation rate has proved to be too fast and rarely homogeneous. Fe-based alloys show appropriate mechanical properties but very low degradation rate. In the present work, four novel Zn-Mg and two Zn-Al binary alloys were investigated as potential biodegradable materials for stent applications. The alloys were developed by casting process and homogenized at 350°C for 48h followed by hot extrusion at 250°C. Tube extrusion was performed at 300°C to produce tubes with outer/inner diameter of 4/1.5mm as precursors for biodegradable stents. Corrosion tests were performed using Hanks׳ modified solution. Extruded alloys exhibited slightly superior corrosion resistance and slower degradation rate than those of their cast counterparts, but all had corrosion rates roughly half that of a standard purity Mg control. Hot extrusion of Zn-Mg alloys shifted the corrosion regime from localized pitting to more uniform erosion, mainly due to the refinement of second phase particles. Zn-0.5Mg is the most promising material for stent applications with a good combination of strength, ductility, strain hardening exponent and an appropriate rate of loss of mechanical integrity during degradation. An EBSD analysis in the vicinity of the laser cut Zn-0.5Mg tube found no grain coarsening or texture modification confirming that, after laser cutting, the grain size and texture orientation of the final stent remains unchanged. This work shows the potential for Zn alloys to be considered for stent applications. Copyright © 2016 Elsevier Ltd. All rights reserved.

Cardox braking unit and its application results; Cardox kirma unitesi ve uygulama sonuclari

Energy Technology Data Exchange (ETDEWEB)

Basaran, C.; Basaran, S.C. [Tire Linyit Sanayi ve Ticaret A.S., Izmir (Turkey)

1996-11-01

The Cardox system is based on a high grade alloy-steel tube, closed both ends with screw-in plugs. One of these plugs incorporates a filling head and electrical initiation system, while the other acts as the discharge part of the system. A chemical energising mixture, which is initiated by a low tension electric fuse head, is inserted into the firing head end of the tube and the head is then screwed into place. The discharge end of the tube is fitted with a mild steel disc of predetermined thickness. Finally the tube is filled with a charge of liquid CO{sub 2}. Passing an impulse from a low tension activator through the fuse head initiates a rapid reaction in the energiser causing the CO{sub 2} to change instantaneously from liquid to gas. Some of the advantages of Cardox are: no permits required for storage use or transportation, large clean lump coal, safe in gaseous mines, immediate return to the working face, cost effective. 7 refs., 5 figs.

Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

International Nuclear Information System (INIS)

An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho

2016-01-01

For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B 4 C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

Corium Configuration and Penetration Tube Failure for Fukushima Daiichi Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

An, Sang Mo; Lee, Jae Bong; Kim, Hwan Yeol; Song, Jin Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

For the LWRs (light water reactors), the penetration tubes at the reactor vessel lower head are regarded as the most vulnerable structures along with a global vessel failure during a severe accident because they can be seriously damaged by a corium melt or debris relocated into the lower plenum of the vessel. The research on the penetration tube failure is of higher importance in the BWRs, as it could lead to melt discharge into the containment and subsequent release of radioactive materials to the environment due to the containment failure. There are more than one hundred of penetration tubes in the Fukushima Daiichi NPPs (nuclear power plants), such as ICM-GTs (in-core monitoring guide tubes), CRGTs (control rod guide tubes) and drain tubes. The ICM-GTs include SRMs (source range monitors), IRMs (intermediate range monitors), LPRMs (local power range monitors) and TIPs (traversing in-core probes), which are much thinner than other tubes. The experimental researches to investigate the corium configuration and the penetration tube failure for the Fukushima Daiichi NPPs were introduced and some meaningful results were summarized. It was shown that the corium ingot was separated into two layers, of which the upper layer was metal-rich while the lower one was oxide-rich. It seemed that B{sub 4}C would contribute to reducing the density of the metallic melt. The two-layered configuration will provide useful information to understand the core melt progression and post-recovery actions for the Fukushima Daiichi NPPs. In addition, we performed a large scale penetration tube failure experiment for the SRM/IRM guide tube, and showed high possibilities of large amount of corium discharge out of the reactor vessel lower head, which followed by the tube melting in a very short time. We are planning to perform the penetration tube failure experiments for another dry tube of ICM-GT (LPRM guide tube), and later for the wet tube (CRGT)

Enhanced Hot Tensile Ductility of Mg-3Al-1Zn Alloy Thin-Walled Tubes Processed Via a Combined Severe Plastic Deformation

Science.gov (United States)

Fata, A.; Eftekhari, M.; Faraji, G.; Mosavi Mashhadi, M.

2018-05-01

In the current study, combined parallel tubular channel angular pressing (PTCAP) and tube backward extrusion (TBE), as a recently developed severe plastic deformation (SPD) method, were applied at 300 °C on a commercial Mg-3Al-1Zn alloy tubes to achieve an ultrafine grained structure. Then, the microstructure, hardness, tensile properties, and fractography evaluations were done at room temperature on the SPD-processed samples. Also, to study the hot tensile ductility of the SPD-processed samples, tensile testing was performed at an elevated temperature of 400 °C, and then, the fractured surface of the tensile samples was studied. It was observed that a bimodal microstructure, with large gains surrounded by many tiny ones, was created in the sample processed by PTCAP followed by TBE. This microstructure led to reach higher hardness and higher strength at room temperature and also led to reach very high elongation to failure ( 181%) at 400 °C. Also, the value of elongation to failure for this sample was 14.1% at room temperature. The fractographic SEM images showed the occurrence of predominately ductile fracture in the samples pulled at 400 °C. This was mostly due to the nucleation of microvoids and their subsequent growth and coalescence with each other.

Macroscopic and microscopic determinations of residual stresses in thin oxide dispersion strengthened steel tubes

International Nuclear Information System (INIS)

Bechade, J.L.; Toualbi, L.; Bosonnet, S.; Carlan, Y. de; Castelnau, O.

2014-01-01

To improve the efficiency of components operating at high temperatures, many efforts are deployed to develop new materials. Oxide Dispersion Strengthened (ODS) materials could be used for heat exchangers or cladding tubes for the new GENIV nuclear reactors. This type of materials are composed with a metallic matrix (usually iron base alloy for nuclear applications or nickel base alloy for heat exchangers) reinforced by a distribution of nano-oxides. They are obtained by powder metallurgy and mechanical alloying. The creep resistance of these materials is excellent, and they usually exhibit a high tensile strength at room temperature. Depending on the cold working and/or the heat treatments, several types of microstructure can be obtained: recrystallised, stress relieved. One of the key challenges is to transform ODS materials into thin tubes (up to 500 microns thick) within a robust fabrication route while keeping the excellent mechanical properties. To prevent cracking during the process or to obtain a final product with low residual stresses, it is important to quantify the effect of the heat treatments on the release of internal stresses. The aim of this study is to show how residual stresses can be determined on different thin tubes using two complementary approaches: (i) macroscopic stresses determination in the tube using beam theory (small cuts along the longitudinal and circumferential directions and measurements of the deflection), (ii) stress determination from x-ray diffraction analyses (surface analyses, using 'sin"2ψ' method with different hypothesis). Depending on the material and the heat treatment, residual stresses vary dramatically and can reach 800 MPa which is not far from the yield stress; comparisons between both methods are performed and suggestions are given in order to optimize the thermo-mechanical treatment of thin ODS tubes. (authors)

A comparison of wakeboard-, water skiing-, and tubing-related injuries in the United States, 2000-2007.

Science.gov (United States)

Baker, John I; Griffin, Russell; Brauneis, Paul F; Rue, Loring W; McGwin, Gerald

2010-01-01

The purpose of the study was to compare tubing-related injuries to wakeboarding- and water skiing-related injuries. Data was collected from the 2000-2007 National Electronic Injury Surveillance Survey for 1,761 individuals seeking care at an emergency department due to a tubing-, wakeboarding, or water skiing-related injury. Data included patient age and sex, as well as injury characteristics including body region injured (i.e., head and neck, trunk, shoulder and upper extremity, and hip and lower extremity) and diagnosis of injury (e.g., contusion, laceration, or fracture). Case narratives were reviewed to ensure that a tubing-, wakeboarding-, or water skiing-related injury occurred while the individual was being towed behind a boat. Severe injury (defined as an injury resulting in the individual being hospitalized, transferred, held for observation) was compared among the groups using logistic regression. Wakeboard- and tubing-related injuries more commonly involved the head and neck, while water skiing- related injuries were likely to involve the hip and lower extremity. Tubing-related injuries, compared to water skiing-related injuries, were more likely to be severe (OR 2.31, 95% CI 1.23-4. 33). Like wakeboarding and water skiing, tubing has inherent risks that must be understood by the participant. While tubing is generally considered a safer alternative to wakeboarding and water skiing, the results of the current study suggest otherwise. Both the number and severity of tubing- related injuries could be prevented through means such as advocating the use of protective wear such as helmets while riding a tube or having recommended safe towing speeds prominently placed on inner tubes. Key pointsIncrease annual injury rate trend in wakeboard injuries.Wakeboard- and tubing-related injuries more often to head and neck, waterskiing-related injuries more often to hip and lower extremity.Tubing-related injuries over 2-times as likely to be severe compared to

Ballooning of CANDU pressure tube in local thermal transients

International Nuclear Information System (INIS)

Mihalache, Maria; Ionescu, Viorel

2008-01-01

In certain LOCA scenarios for the CANDU fuel channel, the ballooning of the pressure tube and contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator takes place through the contact area. If the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. In INR-Pitesti the DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after occurrence of the contact between the two tubes. The code contains few models: thermal creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This paper gives a DELOCA code description and the fuel channel behaviour analysis, in transient temperature conditions of the pressure tube, using the materials properties, time and temperature dependencies of these properties as obtained in the different laboratories of the world and in the INR - Pitesti in the last years. DELOCA computer code simulated the fuel channel response to the constant heating rates of inside pressure tube surface. The paper presents contact temperature and time dependencies on the heating rate, and the appropriate fitting functions. (authors)

Process Technology Development of Ni Electroplating in Steam Generator Tube

International Nuclear Information System (INIS)

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

2009-11-01

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

Production technique of vermicular graphite iron cylinder head of vehicle diesel engine

Directory of Open Access Journals (Sweden)

Zhou Gen

2008-11-01

Full Text Available The 25 years’production and application have proved that vermicular graphite iron cylinder heads with vermicularity ≥50% satisfy the machinability and performance demand of diesel engine. The method, in which using cupola-induction furnace duplex melting and pour-over process with rare earth-ferrosilicon or rare earthsilicon compound as vermicularizing alloy plus rare earth-magnesium-ferrosilicon as stirring alloy, is an optimal vermicularizing process for obtaining satisfi ed vermicularity. Using top kiss risers, enlarging kissing areas and expanding covering width and making ingates to freeze earlier are the effective measures to eliminate shrinkage, blowhole and oxide inclusions in the vermicular graphite iron cylinder heads.

PROFIL-360 high resolution steam generator tube profilometry system

International Nuclear Information System (INIS)

Glass, S.W.

1985-01-01

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

Profil-360 high resolution steam generator tube profilometry system

International Nuclear Information System (INIS)

Glass, S.W.

1985-01-01

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

Evaluating the Role of Prophylactic Gastrostomy Tube Placement Prior to Definitive Chemoradiotherapy for Head and Neck Cancer

International Nuclear Information System (INIS)

Chen, Allen M.; Li Baoqing; Lau, Derick H.; Farwell, D. Gregory; Luu, Quang; Stuart, Kerri; Newman, Kathleen; Purdy, James A.; Vijayakumar, Srinivasan M.D.

2010-01-01

Purpose: To determine the effect of prophylactic gastrostomy tube (GT) placement on acute and long-term outcome for patients treated with definitive chemoradiotherapy for locally advanced head and neck cancer. Methods and Materials: One hundred twenty consecutive patients were treated with chemoradiotherapy for Stage III/IV head and neck cancer to a median dose of 70 Gy (range, 64-74 Gy). The most common primary site was the oropharynx (66 patients). Sixty-seven patients (56%) were treated using intensity-modulated radiotherapy (IMRT). Seventy patients (58%) received prophylactic GT placement at the discretion of the physician before initiation of chemoradiotherapy. Results: Prophylactic GT placement significantly reduced weight loss during radiation therapy from 43 pounds (range, 0 to 76 pounds) to 19 pounds (range, 0 to 51 pounds), which corresponded to a net change of -14% (range, 0% to -30%) and -8% (range, +1% to -22%) from baseline, respectively (p < 0.001). However, the proportion of patients who were GT-dependent at 6- and 12-months after treatment was 41% and 21%, respectively, compared with 8% and 0%, respectively, for those with and without prophylactic GT (p < 0.001). Additionally, prophylactic GT was associated with a significantly higher incidence of late esophageal stricture compared with those who did not have prophylactic GT (30% vs. 6%, p < 0.001). Conclusions: Although prophylactic GT placement was effective at preventing acute weight loss and the need for intravenous hydration, it was also associated with significantly higher rates of late esophageal toxicity. The benefits of this strategy must be balanced with the risks.

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

International Nuclear Information System (INIS)

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

2012-09-01

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

Accounting for the residual stress effects on the creep deformation of channel tubes

International Nuclear Information System (INIS)

Knizhnikov, Yu.N.; Platonov, P.A.; Ul'yanov, A.I.

1985-01-01

The effect of the first kind residual stresses arising in the walls of the zirconium base alloy fules in the process of fabrication on the RBMK type reactor channel tube creep is investigated. Models for calculation of the reactor component creep with account for the relaxation of residual stresses distributed by the wall thickness as well as the radiation and temperature fields are developed. On the basis of the analysis of the data obtained it is concluded that the effect of the residual stresses on the RBMK channel tube deformation for a long-term operation is negligible. But for the short-term fests the results can be noticeably distorted by this factor. The role of internal stresses can also manifest when determining the deformation of radiation elongation of the zirconium base alloy samples

The Feasibility and Outcome of Oro-esophageal Tube Feeding in Patients with Various Etiologies.

Science.gov (United States)

Kim, Juyong; Seo, Han Gil; Lee, Goo Joo; Han, Tai Ryoon; Oh, Byung-Mo

2015-12-01

The oro-esophageal tube (OE tube) is widely used in dysphagia patients although its success rate for transition to oral feeding is reported only in stroke patients. The aim of this study was to evaluate the feasibility and outcome of OE tube feeding for patients with dysphagia resulting from various etiologies. The authors reviewed the medical records of 1995 dysphagic patients that had undergone videofluoroscopic swallowing study (VFSS) in a tertiary hospital from April 2002 through December 2009. Of these, 97 patients were recommended to use OE tube feeding based on the VFSS findings. Follow-up VFSS were performed on 54 patients. The mean duration of tube use at the time of follow-up VFSS was 274 days. We evaluated clinical information including age, sex, diet, etiology of dysphagia, location of lesions, duration of intervention, and complications of OE tube feeding. Initially, all 54 patients were fed using the OE tube. After their last follow-up evaluation, 19 patients (35.2 %) resumed full oral feeding without the OE tube, 12 patients (22.2 %) used partial OE tube feeding, and 23 patients (42.6 %) continued OE tube feeding only. Full oral feeding was achieved again most often in brain tumor, stroke, and head and neck cancer patients (54.5, 27.3, and 20.0 %, respectively). Mild adverse events, such as blood-tinged sputum, nausea, dyspepsia, and regurgitation of food, were reported in 4 patients. OE tube feeding is a feasible feeding method also in conditions other than stroke such as brain tumors, and head and neck cancers.

Tube pancreatico-duodenostomy for management of a severe penetrating pancreaticoduodenal injury.

Science.gov (United States)

Hatzigeorgiadis, Anestis; Boulas, Konstantinos A; Barettas, Nikolaos; Papageorgiou, Irene; Blouhos, Konstantinos

2014-05-27

Optimal management of penetrating pancreaticoduodenal injuries and better outcomes are associated with simple, fast damage control surgery and shorter operative time. The performance of pyloric exclusion and tube duodenostomy has markedly decreased. However, there is still a trend toward their performance in cases of delay duodenal repair or severe pancreaticoduodenal injury. The present report describes a case of a hemodynamically stable patient with a single penetrating gunshot trauma causing an AAST-OIS grade III pancreatic head injury and grade IV injury of the second portion of the duodenum. The patient was treated in our Level IV rural trauma center and submitted to primary closure of the posterolateral duodenal wall (the laceration of the contralateral inner medial duodenal wall could not be repaired), external duodenal and pancreatic drainage, and duodenal decompression by tube pancreatico-duodenostomy (insertion of a 18 Fr Foley catheter through the laceration of the pancreatic head toward the duodenal lumen), tube cholangiostomy, and pyloric exclusion accompanied with a feeding jejunostomy. Tube pancreatico-duodenostomy, which is described for the first time in the literature, turned out to be effective and can be considered as an option in pancreaticoduodenal trauma when the inner medial duodenal wall cannot be repaired.

Modification of structural phase state in superficial layers of fuel tubes made of Zirconium alloys

International Nuclear Information System (INIS)

Volkov, N.; Kalin, B.; Pimenov, Y.; Timoshin, S.

2011-01-01

The paper presents the results obtained in developing the method for introduction of the required changes into states and properties of outer surface on fuel rod cladding made of zirconium alloys E110 and E635 through irradiation by radial Ar + ion beam with a broad energy spectrum. In particular, the paper demonstrates that ion beam treatment of the claddings surface, at the final stage of their fabrication, can upgrade substantially quality of outer tubular surface after mechanical polishing (the cleaner surface, the lower roughness, removal of technological transversal scratches). In addition, the ion beam irradiation results in higher micro-hardness of the modified layer and in better tribological parameters. Kinetic effects in growth of oxide films were studied for the tubular samples of zirconium alloys after ion-beam treatment (cleaning and polishing by radial Ar + ion beam). Also, corrosion tests of the tubular samples were carried out in water (at 350 0 C) and steam (at 350, 375 and 400 0 C) with duration up to 3000 hours. It was revealed that oxide layer consisting mainly of zirconium dioxide in monoclinic modification was formed on tubular surface after oxidation at 3500 0 C in water or steam. The oxidizing process in the pressurized steam created thicker oxide layer on tubular surface than that in the pressurized water. Experimental data were used to determine optimal conditions for ion-beam treatment of outer fuel tube surface. The tubular samples with the following geometrical parameters were investigated: length - up to 500 mm, diameter - 9,15 mm. Optimal regimes for ion-beam cleaning and polishing of the tubular samples were studied up to the process rate of 1 meter per minute. Within the frames of linear approximation, analytical relationships were derived for time dependent growth of oxide films and used to evaluate thickness of oxide film under test conditions (duration . up to 10000 hours). Thickness of oxide films can cover the range from 6 �

A statistical approach to the prediction of pressure tube fracture toughness

International Nuclear Information System (INIS)

Pandey, M.D.; Radford, D.D.

2008-01-01

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in a nuclear reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The paper presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower-bound approach

The transformation behaviour of the beta phase in Zr-2.5 wt% Nb pressure tubes

International Nuclear Information System (INIS)

Griffiths, M.; Winegar, J.E.

1994-12-01

A temperature-time-transformation (TTT) diagram has been developed for the β-phase in Zr-2.5 wt% Nb pressure tubes. The results show that the morphology and/or physical state of the β-phase in pressure tubes has a significant effect on the transformation behaviour compared with a bulk Zr-19 wt%Nb alloy. (author). 14 refs., 1 tab., 15 figs

Fatigue of thin walled tubes in copper alloy CuNi10

DEFF Research Database (Denmark)

Lambertsen, Søren Heide; Damkilde, Lars; Jepsen, Michael S.

2016-01-01

The current work concerns the investigation of the fatigue resistance of CuNi10 tubes, which are frequently used in heat exchangers of large ship engines. The lifetime performances of the exchanger tubes are greatly affected by the environmental conditions, where especially the temperature...... by means of the ASTM E739 guideline and one-sided tolerance limits factor method. The tests show good fatigue resistance and the risk for a failure is low in aspect to the case of a ship heat exchanger....

A computational approach for thermomechanical fatigue life prediction of dissimilarly welded superheater tubes

Energy Technology Data Exchange (ETDEWEB)

Krishnasamy, Ram-Kumar; Seifert, Thomas; Siegele, Dieter [Fraunhofer-Institut fuer Werkstoffmechanik (IWM), Freiburg im Breisgau (Germany)

2010-07-01

In this paper a computational approach for fatigue life prediction of dissimilarly welded superheater tubes is presented and applied to a dissimilar weld between tubes made of the nickel base alloy Alloy617 tube and the 12% chromium steel VM12. The approach comprises the calculation of the residual stresses in the welded tubes with a multi-pass dissimilar welding simulation, the relaxation of the residual stresses in a post weld heat treatment (PWHT) simulation and the fatigue life prediction using the remaining residual stresses as initial condition. A cyclic fiscoplasticity model is used to calculate the transient stresses and strains under thermocyclic service loadings. The fatigue life is predicted with a damage parameter which is based on fracture mechanics. The adjustable parameters of the model are determined based on LCF and TMF experiments. The simulations show, that the residual stresses that remain after PWHT further relax in the first loading cycles. The predicted fatigue lives depend on the residual stresses and, thus, on the choice of the loading cycle in which the damage parameter is evaluated. It the first loading cycle, where residual stresses are still present, is considered, lower fatigue lives are predicted compared to predictions considering loading cycles with relaxed residual stresses. (orig.)

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-07-01

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

A tube seepage meter for in situ measurement of seepage rate and groundwater sampling

Science.gov (United States)

Solder, John; Gilmore, Troy E.; Genereux, David P.; Solomon, D. Kip

2016-01-01

We designed and evaluated a “tube seepage meter” for point measurements of vertical seepage rates (q), collecting groundwater samples, and estimating vertical hydraulic conductivity (K) in streambeds. Laboratory testing in artificial streambeds show that seepage rates from the tube seepage meter agreed well with expected values. Results of field testing of the tube seepage meter in a sandy-bottom stream with a mean seepage rate of about 0.5 m/day agreed well with Darcian estimates (vertical hydraulic conductivity times head gradient) when averaged over multiple measurements. The uncertainties in q and K were evaluated with a Monte Carlo method and are typically 20% and 60%, respectively, for field data, and depend on the magnitude of the hydraulic gradient and the uncertainty in head measurements. The primary advantages of the tube seepage meter are its small footprint, concurrent and colocated assessments of q and K, and that it can also be configured as a self-purging groundwater-sampling device.

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

Leak on a steam generator tube: in-depth analysis

International Nuclear Information System (INIS)

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

2015-01-01

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

The development of zirconium alloy and its manufacturing

International Nuclear Information System (INIS)

Yuan Gaihuan; Yue Qiang

2015-01-01

Nuclear power which acts as one of low-carbon energy resources is the most realistic in large-scale application. It is also the preferred choice for many countries to develop energy resources and optimize its structure. Zirconium alloy is a key structural material for nuclear power plant fuel assemblies and cladding tubes of zirconium alloy are often referred as the first safeguard to nuclear power safety. With the development of nuclear power, three kinds of zirconium alloys Zr-Sn, Zr-Nb, Zr-Sn-Nb and with the representative products of Zr-4, M5, Zirlo respectively are developed and widely applied. Because of its severe operating environment and influence to nuclear safety, the requirements to zirconium alloys for physical and chemical properties, nuclear capability, tolerance and surface quality are very strict. The in-depth research and its manufacture capability become one of the main barriers for many countries who are developing the nuclear energy. In recent years, a stated-owned company, State Nuclear Bao Ti Zirconium Industry Company ('SNZ' for short) as well as National R and D Center for Nuclear Grade Zirconium material, is founded to meet the requirement of the rapid development of China's nuclear power industry. SNZ is dedicated for the fabrication and the research of nuclear grade zirconium products. After the successful completion of technology transfer of manufacturing for production chain and fully grasped of the manufacturing technology for the nuclear grade zirconium sponge through zirconium alloy tube, rod and strip products. National R and D Center for Nuclear Grade Zirconium material is cooperating with universities, nuclear energy research and design institutes and the owners of nuclear power plant to develop new zirconium alloy of self-owned brand. Through the selection of components, in-process testing and product inspection, four kinds of new zirconium alloys owns better performance than currently commercialized M5, Zirlo etc

Safety of pull-type and introducer percutaneous endoscopic gastrostomy tubes in oncology patients: a retrospective analysis

Directory of Open Access Journals (Sweden)

Pelckmans Paul A

2011-03-01

Full Text Available Abstract Background Percutaneous endoscopic gastrostomy (PEG allows long-term tube feeding. Safety of pull-type and introducer PEG placement in oncology patients with head/neck or oesophageal malignancies is unknown. Methods Retrospective analysis of 299 patients undergoing PEG tube placement between January 2006 and December 2008 revealed 57 oncology patients. All patients with head/neck or oesophageal malignancy were treated with chemo- and radiotherapy. In case of high-grade stenosis introducer Freka® Pexact PEG tube was placed (n = 24 and in all other patients (n = 33 conventional pull-type PEG tube. Short-term complications and mortality rates were compared. Results Patients' characteristics and clinical status were comparable in both groups. Short-term complications were encountered in 11/24 (48% introducer PEG patients as compared to only 4/33 (12% pull-type PEG patients (P vs. 0/33 (0%, P vs. 3/33 (9%, NS. Finally, 3/24 gastrointestinal perforations (12% resulted from a difficult placement procedure vs. 1/33 (3%, leading to urgent surgical intervention and admission to ICU. Two introducer PEG patients died at ICU, resulting in an overall mortality rate of 8% vs. 0% (P = 0.091. Conclusion The introducer Freka® Pexact PEG procedure for long-term tube feeding may lead to significantly higher complication and mortality rates in patients with head/neck or oesophageal malignancies treated with chemo- and radiotherapy. It is suggested to use the conventional pull-type PEG tube placement in this group of patients, if possible.

Control rod guide tube cleaning device

International Nuclear Information System (INIS)

Tsuji, Tadashi; Shiota, Yoshiaki.

1990-01-01

Since there was no exclusive device for cleaning control rods, no effective cleaning could not be conducted and there was a possibility that obstacles may not be recovered. Then, there are disposed a first pump for supplying pressurized water, a spray nozzle for forming a swirling flow in a control rod guide tube, a second pump for pressurizing water introduced by a sucking pipeline and a collecting device for recovering obstacles intruding to water from the second pump. The pressurized water supplied from the first pump is introduced to a head passing through a blowing pipe and jetted from the spray nozzle to the control rod guide tube. In this case, a swirling stream occurs and obstacles in the control guide tube are mixed into water. The water containing the obstacles passes from the sucking port through a pipeline, introduced to the second pump and recovered to the collecting device. Since there is no water staying portion upon cleaning operation, the obstacles accumulating over the entire region of the bottom of the guide tube can be recovered reliably and efficiently. (N.H.)

Experimental residual stress evaluation of hydraulic expansion transitions in Alloy 690 steam generator tubing

International Nuclear Information System (INIS)

McGregor, R.; Doherty, P.; Hornbach, D.; Abdelsalam, U.

1995-01-01

Nuclear Steam Generator (SG) service reliability and longevity have been seriously affected worldwide by corrosion at the tube-to-tubesheet joint expansion. Current SG designs for new facilities and replacement projects enhance corrosion resistance through the use of advanced tubing materials and improved joint design and fabrication techniques. Here, transition zones of hydraulic expansions have undergone detailed experimental evaluation to define residual stress and cold-work distribution on and below the secondary-side surface. Using X-ray diffraction techniques, with supporting finite element analysis, variations are compared in tubing metallurgical condition, tube/pitch geometry, expansion pressure, and tube-to-hole clearance. Initial measurements to characterize the unexpanded tube reveal compressive stresses associated with a thin work-hardened layer on the outer surface of the tube. The gradient of cold-work was measured as 3% to 0% within .001 inch of the surface. The levels and character of residual stresses following hydraulic expansion are primarily dependent on this work-hardened surface layer and initial stress state that is unique to each tube fabrication process. Tensile stresses following expansion are less than 25% of the local yield stress and are found on the transition in a narrow circumferential band at the immediate tube surface (< .0002 inch/0.005 mm depth). The measurements otherwise indicate a predominance of compressive stresses on and below the secondary-side surface of the transition zone. Excellent resistance to SWSCC initiation is offered by the low levels of tensile stress and cold-work. Propagation of any possible cracking would be deterred by the compressive stress field that surrounds this small volume of tensile material

Corrosion resistant properties and weldabilities of ASTM Grade 12 titanium alloy

International Nuclear Information System (INIS)

Tsumori, Yoshikatsu; Itoh, Hideo

1988-01-01

Plates, sheets, bars, wires and thinner seam-welded tubings were manufactured from large-scaled ingot of ASTM Grade 12 alloy (Ti-0.8Ni-0.3Mo). The processability of G-12 alloy has proved almost similar to that of conventional commercially pure titanium grades. It has been clarified that the G-12 alloy showed several advantageous features: Chlorides-Crevice corrosion resistance of the alloy was almost equals to G-7 and Pd0/TiO 2 coated titanium, and the maximum allowable stress was able to be designed higher than that of commercially pure titanium. This alloy has been in applications also offers where such environments as seawater, brines and moist chlorine, various oil refinery and chemical industries, and others. (author)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

CFD based draft tube hydraulic design optimization

International Nuclear Information System (INIS)

McNabb, J; Murry, N; Mullins, B F; Devals, C; Kyriacou, S A

2014-01-01

The draft tube design of a hydraulic turbine, particularly in low to medium head applications, plays an important role in determining the efficiency and power characteristics of the overall machine, since an important proportion of the available energy, being in kinetic form leaving the runner, needs to be recovered by the draft tube into static head. For large units, these efficiency and power characteristics can equate to large sums of money when considering the anticipated selling price of the energy produced over the machine's life-cycle. This same draft tube design is also a key factor in determining the overall civil costs of the powerhouse, primarily in excavation and concreting, which can amount to similar orders of magnitude as the price of the energy produced. Therefore, there is a need to find the optimum compromise between these two conflicting requirements. In this paper, an elaborate approach is described for dealing with this optimization problem. First, the draft tube's detailed geometry is defined as a function of a comprehensive set of design parameters (about 20 of which a subset is allowed to vary during the optimization process) and are then used in a non-uniform rational B-spline based geometric modeller to fully define the wetted surfaces geometry. Since the performance of the draft tube is largely governed by 3D viscous effects, such as boundary layer separation from the walls and swirling flow characteristics, which in turn governs the portion of the available kinetic energy which will be converted into pressure, a full 3D meshing and Navier-Stokes analysis is performed for each design. What makes this even more challenging is the fact that the inlet velocity distribution to the draft tube is governed by the runner at each of the various operating conditions that are of interest for the exploitation of the powerhouse. In order to determine these inlet conditions, a combined steady-state runner and an initial draft tube analysis

CFD based draft tube hydraulic design optimization

Science.gov (United States)

McNabb, J.; Devals, C.; Kyriacou, S. A.; Murry, N.; Mullins, B. F.

2014-03-01

The draft tube design of a hydraulic turbine, particularly in low to medium head applications, plays an important role in determining the efficiency and power characteristics of the overall machine, since an important proportion of the available energy, being in kinetic form leaving the runner, needs to be recovered by the draft tube into static head. For large units, these efficiency and power characteristics can equate to large sums of money when considering the anticipated selling price of the energy produced over the machine's life-cycle. This same draft tube design is also a key factor in determining the overall civil costs of the powerhouse, primarily in excavation and concreting, which can amount to similar orders of magnitude as the price of the energy produced. Therefore, there is a need to find the optimum compromise between these two conflicting requirements. In this paper, an elaborate approach is described for dealing with this optimization problem. First, the draft tube's detailed geometry is defined as a function of a comprehensive set of design parameters (about 20 of which a subset is allowed to vary during the optimization process) and are then used in a non-uniform rational B-spline based geometric modeller to fully define the wetted surfaces geometry. Since the performance of the draft tube is largely governed by 3D viscous effects, such as boundary layer separation from the walls and swirling flow characteristics, which in turn governs the portion of the available kinetic energy which will be converted into pressure, a full 3D meshing and Navier-Stokes analysis is performed for each design. What makes this even more challenging is the fact that the inlet velocity distribution to the draft tube is governed by the runner at each of the various operating conditions that are of interest for the exploitation of the powerhouse. In order to determine these inlet conditions, a combined steady-state runner and an initial draft tube analysis, using a

Proposal of 99.99%-aluminum/7N01-Aluminum clad beam tube for high energy booster of Superconducting Super Collider

International Nuclear Information System (INIS)

Ishimaru, Hajime

1994-01-01

Proposal of 99.99% pure aluminum/7N01 aluminum alloy clad beam tube for high energy booster in Superconducting Super Collider is described. This aluminum clad beam tube has many good performances, but a eddy current effect is large in superconducting magnet quench collapse. The quench test result for aluminum clad beam tube is basically no problem against magnet quench collapse. (author)

Influence of impurities and ion surface alloying on the corrosion resistance of E110 alloy

International Nuclear Information System (INIS)

Kalin, B. A.; Volkov, N. V.; Valikov, R. A.; Novikov, V. V.; Markelov, V. A.; Pimenov, Yu. V.

2013-01-01

The corrosion resistance of zirconium alloys depends on their structural-phase state, the type of core coolant and operating factors. The formation of a protective oxide film on the zirconium alloys is sensitive to the content of impurity atoms present in the charge base of alloys and accumulating in them in the manufacture of products. The impurity composition of the initial zirconium is determined by the method of its manufacture and generally remains unchanged in the products, deter-mining their properties, including their corrosion resistance. An increased content of impurities (C, N, Al, Mo, Fe) both individually and in their combination negatively affects the corrosion resistance of zirconium and its alloys. One of the potentially effective methods to increase the protective properties of oxide films on zirconium alloys is a surface alloying using the regime of mixing the atoms of a film, preliminarily coated on the surface, and the atoms of a target. This method makes it possible to form a given structural-phase state in the thin surface layer with unique physicochemical properties and thus to in-crease the corrosion resistance and wear resistance of fuel claddings. In this context, the object of investigation was samples of cladding tubes from alloy E110 with various content of impurity elements (nitrogen, aluminum, and carbon) with the aim to reduce the negative influence of impurities on the corrosion resistance by changing the structural-phase state of the surface layer of fuel claddings and fuel assembly components with alloying in the regime of ion mixing of atoms

Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

Energy Technology Data Exchange (ETDEWEB)

Gomez Briceno, D.; Serrano, M.

2005-07-01

Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

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)

Corrosion phenomena on alloy 625 in aqueous solutions containing hydrochloric acid and oxygen under subcritical and supercritical conditions

International Nuclear Information System (INIS)

Boukis, N.; Kritzer, P.

1997-01-01

Supercritical Water Oxidation (SCWO) is a very effective process to destroy hazardous aqueous wastes containing organic contaminants. The main target applications in the USA are the destruction of DOD and DOE wastes such as rocket fuels and explosives, warfare agents and organics present in low level radioactive liquid wastes. Alloy 625 is frequently used as reactor material for Supercritical Water Oxidation (SCWO) applications. This is due to the favorable combination of mechanical properties, corrosion resistance, price and availability. Nevertheless, the corrosion of alloy 625 like the corrosion of other Ni-base alloys during oxidation of hazardous organic waste containing chloride proceeds too fast and is a major problem in SCWO applications. In these experiments high pressure, high-temperature resistant tube reactors made of alloy 625 were used as specimens. They were exposed to SCWO conditions, without organics, at temperatures up to 500 C and pressures up to 37 MPa for up to 150 h. Simultaneously, coupons also made from alloy 625 are exposed inside the test tubes. The most important corrosion problem for alloy 625 is pitting and intercrystalline corrosion at temperatures near the critical temperature, i.e. in the preheater and cooling sections of the test tubes. Under certain conditions, stress corrosion cracking appears and leads to premature failure of the test reactors. The corrosion products were insoluble in supercritical water and formed thick layers in the supercritical part of the reactor. Under these layers only minor corrosion occurred. 33 refs

Irradiation of Wrought FeCrAl Tubes in the High Flux Isotope Reactor

Energy Technology Data Exchange (ETDEWEB)

Linton, Kory D. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Field, Kevin G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Petrie, Christian M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-09-01

The Advanced Fuels Campaign within the Nuclear Technology Research and Development program of the Department of Energy Office of Nuclear Energy is seeking to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are one of the leading candidate materials for fuel cladding to replace traditional zirconium alloys because of the superior oxidation resistance of FeCrAl. However, there are still some unresolved questions regarding irradiation effects on the microstructure and mechanical properties of FeCrAl at end-of-life dose levels. In particular, there are concerns related to irradiation-induced embrittlement of FeCrAl alloys due to secondary phase formation. To address this issue, Oak Ridge National Laboratory has developed a new experimental design to irradiate shortened cladding tube specimens with representative 17×17 array pressurized water reactor diameter and thickness in the High Flux Isotope Reactor (HFIR) under relevant temperatures (300–350°C). Post-irradiation examination will include studies of dimensional change, microstructural changes, and mechanical performance. This report briefly summarizes the capsule design concept and the irradiation test matrix for six rabbit capsules. Each rabbit contains two FeCrAl alloy tube specimens. The specimens include Generation I and Generation II FeCrAl alloys with varying processing conditions, Cr concentrations, and minor alloying elements. The rabbits were successfully assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. Pictures of the rabbit assembly process and detailed dimensional inspection of select specimens are included in this report. The rabbits were inserted into HFIR starting in cycle 472 (May 2017).

Percutaneous endoscopic gastrostomy versus nasogastric tube feeding for patients with head and neck cancer. A systematic review

International Nuclear Information System (INIS)

Wang Jinfeng; Liu Minjie; Ye Yun; Liu Chao; Huang Guanhong

2014-01-01

There are two main enteral feeding strategies—namely nasogastric (NG) tube feeding and percutaneous gastrostomy—used to improve the nutritional status of patients with head and neck cancer (HNC). But up till now there has been no consistent evidence about which method of enteral feeding is the optimal method for this patient group. To compare the effectiveness of percutaneous gastrostomy and NGT feeding in patients with HNC, relevant literature was identified through Medline, Embase, Pubmed, Cochrane, Wiley and manual searches. We included randomized controlled trials (RCTs) and non-experimental studies comparing percutaneous gastrostomy—including percutaneous endoscopic gastrostomy (PEG) and percutaneous fluoroscopic gastrostomy (PFG)—with NG for HNC patients. Data extraction recorded characteristics of intervention, type of study and factors that contributed to the methodological quality of the individual studies. Data were then compared with respect to nutritional status, duration of feeding, complications, radiotherapy delays, disease-free survival and overall survival. Methodological quality of RCTs and non-experimental studies were assessed with separate standard grading scales. It became apparent from our studies that both feeding strategies have advantages and disadvantages. (author)

Evaluation of radiation dose in pediatric head CT examination: a phantom study

Science.gov (United States)

Norhasrina Nik Din, Nik; Zainon, Rafidah; Rahman, Ahmad Taufek Abdul

2018-01-01

The aim of this study was to evaluate the radiation dose in pediatric head Computed Tomography examination. It was reported that decreasing tube voltage in CT examination can reduce the dose to patients significantly. A head phantom was scanned with dual-energy CT at 80 kV and 120 kV. The tube current was set using automatic exposure control mode and manual setting. The pitch was adjusted to 1.4, 1.45 and 1.5 while the slice thickness was set at 5 mm. The dose was measured based on CT Dose Index (CTDI). Results from this study have shown that the image noise increases substantially with low tube voltage. The average dose was 2.60 mGy at CT imaging parameters of 80 kV and 10 - 30 mAs. The dose increases up to 17.19 mGy when the CT tube voltage increases to 120 kV. With the reduction of tube voltage from 120 kV to 80 kV, the radiation dose can be reduced by 12.1% to 15.1% without degradation of contrast-to-noise ratio.

Zirconium alloy fuel cladding resistant to PCI crack propagation

International Nuclear Information System (INIS)

Boyle, R.F.; Foster, J.P.

1987-01-01

A nuclear fuel element is described cladding tube comprising: concentric tubular layers of zirconium base alloys; the concentric tubular layers including an inner layer and outer layer; the outer layer metallurgically bonded to the inner layer; the outer layer composed of a first zirconium base alloy characterized by excellent resistance to corrosion caused by exposure to high temperature and pressure aqueous environments; the inner layer composed of a second zirconium base alloy consisting of: about 0.2 to 0.6 wt.% tin, about 0.03 to 0.11 wt.% iron, less than about 0.02 wt.% chromium, up to about 350 ppm oxygen and the remainder being zirconium and incidental impurities, and the inner layer characterized by improved resistance to crack propagation under reactor operating conditions compared to the first zirconium alloy

Corrosion aspects of Ni-Cr-Fe based and Ni-Cu based steam generator tube materials

International Nuclear Information System (INIS)

Dutta, R.S.

2009-01-01

This paper reviews corrosion related issues of Ni-Cr-Fe based (in a general sense) and Ni-Cu based steam generator tube materials for nuclear power plants those have been dealt with for last more than four decades along with some updated information on corrosion research. The materials include austenitic stainless steels (SSs), Alloy 600, Monel 400, Alloy 800 and Alloy 690. Compatibility related issues of these alloys are briefly discussed along with the alloy chemistry and microstructure. For austenitic SSs, stress corrosion cracking (SCC) behaviour in high temperature aqueous environments is discussed. For Alloy 600, intergranular cracking in high temperature water including hydrogen-induced intergranular cracking is highlighted along with the interactions of material in various environments. In case of Monel 400, intergranular corrosion and pitting corrosion at ambient temperature and SCC behaviour at elevated temperature are briefly described. For Alloy 800, the discussion covers SCC behaviour, surface characterization and microstructural aspects of pitting, whereas hydrogen-related issues are also highlighted for Alloy 690.

Variation of microstructures and mechanical properties of hot heading process of super heat resisting alloy Inconel 718

International Nuclear Information System (INIS)

Choi, Hong Seok; Ko, Dae Chul; Kim, Byung Min

2007-01-01

Metal forming is the process changing shapes and mechanical properties of the workpiece without initial material reduction through plastic deformation. Above all, because of hot working carried out above recrystallization temperature can be generated large deformation with one blow, it can produce with forging complicated parts or heat resisting super alloy such as Inconel 718 has the worst forgeability. In this paper, we established optimal variation of hot heading process of the Inconel 718 used in heat resisting component and evaluated mechanical properties hot worked product. Die material is SKD61 and initial temperature is 300 .deg. C. Initial billet temperature and punch velocity changed, relatively. Friction coefficient is 0.3 as lubricated condition of hot working. CAE is carried out using DEFORM software before marking the tryout part, and it is manufactured 150 ton screw press with optimal condition. It is know that forming load was decreased according to decreasing punch velocity

Delayed hydride cracking in irradiated Zr-2.5 % Nb pressure tubes

International Nuclear Information System (INIS)

Cirimello, Pablo; Coronel, Pascual; Haddad, Roberto; Lafont, Claudio; Mizrahi, Rafael

2003-01-01

Pressure tubes in CANDU nuclear power plants are made of Zr-2.5 % Nb alloy, which is susceptible to a cracking process called Delayed Hydride Cracking (DHC). Measurement of DHC velocity on irradiated pressure tubes is essential to assure the validity of the Leak Before Break criterion. This work was performed on samples from two pressure tubes taken out of the Embalse NPP in 1995, belonging to fuel channels A-14 and L-12. DHC velocity in the axial direction was measured at 211 C degrees for samples taken from different axial positions, which allowed to study its dependence on fast neutron fluency and irradiation temperature. Non-irradiated material was also tested. It was found that DHC velocity results for the tested material were similar to those obtained for a great number of tubes irradiated in other CANDU plants. (author)

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

International Nuclear Information System (INIS)

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

1985-12-01

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

The Prediction of Microstructure Evolution of 6005A Aluminum Alloy in a P-ECAP Extrusion Study

Science.gov (United States)

Lei, Shi; Jiu-Ba, Wen; Chang, Ren

2018-04-01

Finite element modeling (FEM) was applied for predicting the recrystallized structure in extruded 6005 aluminum alloy, and simulated results were experimentally validated. First, microstructure evolution of 6005 aluminum alloy during deformation was studied by means of isothermal compression test, where the processing parameters were chosen to reproduce the typical industrial conditions. Second, microstructure evolution was analyzed, and the obtained information was used to fit a dynamic recrystallization model implementing inside the DEFORM-3D FEM code environment. FEM of deformation of 6005 aluminum has been established and validated by microstructure comparison. Finally, the obtained dynamic recrystallization model was applied to tube extrusion by using a portholes-equal channel angular pressing die. The finite element analysis results showed that coarse DRX grains occur in the extruded tube at higher temperature and in the extruded tube at the faster speed of the stem. The test results showed material from the front end of the extruded tube has coarse grains (60 μm) and other extruded tube has finer grains (20 μm).

The Prediction of Microstructure Evolution of 6005A Aluminum Alloy in a P-ECAP Extrusion Study

Science.gov (United States)

Lei, Shi; Jiu-Ba, Wen; Chang, Ren

2018-05-01

Finite element modeling (FEM) was applied for predicting the recrystallized structure in extruded 6005 aluminum alloy, and simulated results were experimentally validated. First, microstructure evolution of 6005 aluminum alloy during deformation was studied by means of isothermal compression test, where the processing parameters were chosen to reproduce the typical industrial conditions. Second, microstructure evolution was analyzed, and the obtained information was used to fit a dynamic recrystallization model implementing inside the DEFORM-3D FEM code environment. FEM of deformation of 6005 aluminum has been established and validated by microstructure comparison. Finally, the obtained dynamic recrystallization model was applied to tube extrusion by using a portholes-equal channel angular pressing die. The finite element analysis results showed that coarse DRX grains occur in the extruded tube at higher temperature and in the extruded tube at the faster speed of the stem. The test results showed material from the front end of the extruded tube has coarse grains (60 μm) and other extruded tube has finer grains (20 μm).

Pyrochemical head-end treatment for spent nuclear fuels

International Nuclear Information System (INIS)

Bowersox, D.F.

1977-01-01

A program based upon thermodynamic values and scouting experiments at Argonne National Laboratory is proposed for development of a pyrochemical head-end treatment of spent nuclear fuels to replace the proposed chopping and leaching operation in the Purex process. The treatment consists of separation of the cladding from the oxide fuel by dissolution into liquid zinc; oxide reduction of uranium and plutonium and dissolution into a zinc--magnesium alloy; separation of alkali, alkaline earth, and rare earth fission products into a molten salt; and, finally, separation and recovery of the plutonium and uranium in the alloy. Uranium and plutonium would be separated from the fuel cladding and selected fission products in a form readily dissolvable in nitric acid. The head-end process could be developed eventually into an optimum method for recovering uranium, plutonium, and selected fission products and for minimizing wastes as compact, stable solids. Developmental expenses are not known clearly, but the potential advantages of the process are impressive

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-02-15

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

Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; McRae, G.A.; Coleman, C.E.; Nitheanandan, T.; Sanderson, D.B.

1999-10-01

In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

Ultrasonic inspection of inpile tubes

International Nuclear Information System (INIS)

Boyd, D.M.; Bossi, H.

1985-01-01

The in-service inspection (ISI) of inpile tubes can be performed accurately and safely with a semiautomatic ultrasonic inspection system. The ultrasonic technique uses a set of multiple transducers to detect and size cracks, voids, and laminations radially and circumferentially. Welds are also inspected for defects. The system is designed to inspect stainless steel and Inconel tubes ranging from 53.8 mm (2.12 in.) to 101.6 mm (4 in.) inner diameter with wall thickness on the order of 5 mm. The inspection head contains seven transducers mounted in a surface-following device. Six angle-beam transducers generate shear waves in the tubes. Two of the six are oriented to detect circumferential cracks, and two detect axial cracks. Although each of these four transducers is used in the pulse-echo mode, they are oriented in aligned sets so pitch-catch operation is possible if desired. The remaining angle-beam transducers are angulated to detect flaws that are off axial or circumferential orientation. The seventh transducer is used for longitudinal inspection and detects and sizes laminar-type defects

Low gravity containerless processing of immiscible gold rhodium alloys

Science.gov (United States)

Andrews, J. Barry

1986-01-01

Under normal one-g conditions immiscible alloys segregate extensively during solidification due to sedementation of the more dense of the immiscible liquid phases. However, under low-g conditions it should be possible to form a dispersion of the two immiscible liquids and maintain this dispersed structure during solidification. Immiscible (hypermonotectic) gold-rhodium alloys were processed in the Marshall Space Flight Center 105 meter drop tube in order to investigate the influence of low gravity, containerless solidification on their microstructure. Hypermonotectic alloys composed of 65 atomic % rhodium exhibited a tendency for the gold rich liquid to wet the outer surface of the containerless processed samples. This tendency led to extensive segregation in several cases. However, well dispersed microstructures consisting of 2 to 3 micron diameter rhodium-rich spheres in a gold-rich matrix were produced in 23.4 atomic % rhodium alloys. This is one of the best dispersions obtained in research on immiscible alloy-systems to data.

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

Pseudobinary eutectics in Cu–Ag–Ge alloy droplets under containerless condition

International Nuclear Information System (INIS)

Ruan, Y.; Wang, X.J.; Lu, X.Y.

2013-01-01

Highlights: ► Two pseudobinary eutectics form in Cu–Ag–Ge alloy. ► It is influenced by thermodynamic and kinetic factors of the alloy in the drop tube. ► As droplet size reduces, anomalous → lamellar → anomalous transition happens in (Ag + ζ). ► (Ag + ε 2 ) is a product of both peri-eutectic and pseudobinary eutectic transitions. -- Abstract: Pseudobinary eutectic generated by pseudobinary eutectic transition or peri-eutectic transition is a crucial structure in ternary alloy systems. Its formation mechanism strongly influences mechanical properties of these metallic materials. However, it was customarily neglected. In this paper, two pseudobinary eutectics, i.e. (Ag + ζ) and (Ag + ε 2 ), were investigated during the rapid solidification of Cu–Ag–Ge ternary alloy in a 3 m-drop tube. The sharp temperature variations and dramatic kinetic activities of the falling alloy droplets before solidification cause special microstructural characteristics. (Ag) dendrite is the heterogeneous nucleus for anomalous (Ag + ζ) pseudobinary eutectic in large droplets. Lamellar (Ag + ζ) pseudobinary eutectic grain forms independently on condition that primary (Ag) dendrite cannot form and its eutectic morphology becomes anomalous with the decrease of droplet size. Nanoscaled (Ag + ε 2 ) pseudobinary eutectic generating at the last stage of solidification is the product of both peri-eutectic and pseudobinary eutectic transitions. It distributes in the gaps of (Ag + ζ) pseudobinary eutectic grains and its morphology remains lamellar regardless of droplet size

Effect of the collector tube profile on Pitot pump performances

Science.gov (United States)

Komaki, K.; Kanemoto, T.; Sagara, K.; Umekage, T.

2013-12-01

The pitot pump is composed of the rotating casing with the impeller channel and the pitot tube type collector as the discharge line. The radial impeller feeds water to the rotating casing. The water rotating together with the casing is caught by the stationary pitot tube type collector, and then discharges to the outside. This type pump, as the extra high head pump, is provided mainly for boiler feed systems, and has been designed by trial and error. To optimize the pump profiles, it is desirable to investigate not only performances but also internal flow conditions. This paper discusses experimentally and numerically the relation between the pump performances and the flow conditions in the rotating casing. The moderately larger dimensions of the collector make the pump head and the discharge high with the higher hydraulic efficiency. The flow in the casing is almost the forced vortex type whose velocity is in proportion to the radius but the core velocity is affected with the drag force of the stationary collector. Based upon the above results, the profile of the pitot tube type collector was optimized with the numerical simulation.

Effect of the collector tube profile on Pitot pump performances

International Nuclear Information System (INIS)

Komaki, K; Sagara, K; Kanemoto, T; Umekage, T

2013-01-01

The pitot pump is composed of the rotating casing with the impeller channel and the pitot tube type collector as the discharge line. The radial impeller feeds water to the rotating casing. The water rotating together with the casing is caught by the stationary pitot tube type collector, and then discharges to the outside. This type pump, as the extra high head pump, is provided mainly for boiler feed systems, and has been designed by trial and error. To optimize the pump profiles, it is desirable to investigate not only performances but also internal flow conditions. This paper discusses experimentally and numerically the relation between the pump performances and the flow conditions in the rotating casing. The moderately larger dimensions of the collector make the pump head and the discharge high with the higher hydraulic efficiency. The flow in the casing is almost the forced vortex type whose velocity is in proportion to the radius but the core velocity is affected with the drag force of the stationary collector. Based upon the above results, the profile of the pitot tube type collector was optimized with the numerical simulation

Influence of microstructure on grain boundary sliding of alloys 600 and 690