Numerical Simulations on the Laser Spot Welding of Zirconium Alloy Endplate for Nuclear Fuel Bundle Assembly

Science.gov (United States)

Satyanarayana, G.; Narayana, K. L.; Boggarapu, Nageswara Rao

2018-03-01

In the nuclear industry, a critical welding process is joining of an end plate to a fuel rod to form a fuel bundle. Literature on zirconium welding in such a critical operation is limited. A CFD model is developed and performed for the three-dimensional non-linear thermo-fluid analysis incorporating buoyancy and Marnangoni stress and specifying temperature dependent properties to predict weld geometry and temperature field in and around the melt pool of laser spot during welding of a zirconium alloy E110 endplate with a fuel rod. Using this method, it is possible to estimate the weld pool dimensions for the specified laser power and laser-on-time. The temperature profiles will estimate the HAZ and microstructure. The adequacy of generic nature of the model is validated with existing experimental data.

Welding of zircalloy-2 and zircalloy-4 by CO2 laser and by TIG

International Nuclear Information System (INIS)

Ram, V.

1990-01-01

This study deals with the welding of zircaloy-2 and zircaloy-4 by means of two techniqes, namely tungsten inert gas welding and CO 2 laser welding. Suitable devices and jigs were developed and manufactured to allow the welding of flat specimens and cylindrical specimens. The optimal welding parameters for the two welding methods were determined. The quality of the welds was determined by tensile strength tests at room temperature and by determining the corrosion resistance to steam at temprature of 450 deg C, 550 deg C, and at 650 deg C. The influence of the weld on the microstructure of the material, on its composition and its crystallographic structure was investigated. Analysis of fracture surfaces of the tensile specimens was carried out with a scanning electron microscope. (author)

Zircaloy-4 corrosion in PWR's

International Nuclear Information System (INIS)

Fyfitch, S.; Smalley, W.R.; Roberts, E.

1985-01-01

Zircaloy-4 waterside corrosion has been studied extensively in the nuclear industry for a number of years. Following the early crud-related corrosion failures in the Saxton test reactor, Westinghouse undertook numerous programs to minimize crud deposition on fuel rods in power reactors through primary coolant chemistry control. Modern plants today are operating with improved coolant chemistry guidelines, and crud deposition levels are very low in proportion to earlier experience. Zircaloy-4 corrosion under a variety of coolant chemistry, heat flux and exposure conditions has been studied extensively. Experience to date, even in relatively high coolant temperature plants, has indicated that -for both fuel cladding and structural components- Zircaloy-4 waterside corrosion performance has been excellent. Recognizing future industry trends, however, which will result in Zircaloy-4 being subjected to ever increasing corrosion duties, Westinghouse will continue accumulating Zircaloy-4 corrosion experience in large power plants. 13 refs.

Characterization of Zircaloy-2 and Zircaloy-4 by X-Ray fluorescence

International Nuclear Information System (INIS)

Sato, I.M.; Imakuma, K.; Salvador, V.L.R.

1981-03-01

The analytical characterization of zircaloy-2 and zircaloy-4 is intimataly connected with the determination of Sn, Fe, Cr, Ni, O, N, H, and Hf. An analytical method developed in this laboratory is discribed for the determination of metallic elements like Sn, Fe, Cr and Ni using the technique of X-ray fluorescence. The samples are prepared in the form of double-layer pellets using boric acid as a binding agent. The zircaloy-4 is dissolved in hydrofluoric acid and the metallic elements are converted to fluorides. The standard samples used for calibration are prepared from synthetic materials. The elements are determined by measuring the characteristic first order K α lines. A Zircaloy-4 sample analysed yielded the following values: Sn=1.30+-0.03%, Fe=0.18+-0.01%, Cr=0.088+-0.004% and Ni=14+-3 ppm. The reproducibility, precision, as well as the theoretical limit of detection of the method are discussed. The determination of the elements O, N and H present as occluded gas in the zircaloy is nearing completion. These analyses are being carried out by a Mass Spectrometric technique where an aliquot of the released gas is analysed. (Author) [pt

Zircaloy-sheathed element rods fitted with thermo-couples; Barre combustible a thermocouple gainee de zircaloy

Energy Technology Data Exchange (ETDEWEB)

Bernardy de Sigoyer, B; Jacques, F; Thome, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1963-07-01

In order to carry out thermal conductivity measurements on UO{sub 2} in conditions similar to those under which fuel rods are used, it was necessary to measure the temperature at the interior of a fuel element sheathed in zircaloy. The temperatures are taken with Thermocoax type thermocouples, that is to say fitted with a very thin sheath of stainless steel or Inconel. It is known also that fusion welding of zircaloy onto stainless steel is impossible and that high temperature welded joints are very difficult because of their aggressiveness. The technique used consists in brazing the thermocouples to relatively large stainless steel parts and then joining these plugs by electron bombardment welding to diffused stainless steel-zircaloy couplings. The properties of these diffused couplings and of the brazed joints were studied; the various stages in the fabrication of the containers are also described. (authors) [French] Pour des mesures de conductivite thermique de l'UO{sub 2} dans des conditions voisines du fonctionnement des barres combustibles, il s'agissait de mesurer la temperature a l'interieur d'un element combustible gaine de zircaloy. Les prises de temperature sont faites par thermocouples du type Thermocoax, c'est-a-dire pourvu d'une gaine tres mince en inox ou inconel. Par ailleurs on sait que le soudage par fusion du zircaloy sur l'inox est impossible et que les brasures a haute temperature sont difficiles car tres agressives. La technique utilisee consiste a braser les thermocouples sur des pieces en inox relativement massives et de rapporter par soudage au bombardement electronique ces bouchons sur des raccords diffuses zircaloy-inox. Les proprietes de ces raccords diffuses et celles de joints brases ont ete etudiees; on expose egalement les diverses etapes de fabrication des containers. (auteurs)

Zircaloy-sheathed element rods fitted with thermo-couples

International Nuclear Information System (INIS)

Bernardy de Sigoyer, B.; Jacques, F.; Thome, P.

1963-01-01

In order to carry out thermal conductivity measurements on UO 2 in conditions similar to those under which fuel rods are used, it was necessary to measure the temperature at the interior of a fuel element sheathed in zircaloy. The temperatures are taken with Thermocoax type thermocouples, that is to say fitted with a very thin sheath of stainless steel or Inconel. It is known also that fusion welding of zircaloy onto stainless steel is impossible and that high temperature welded joints are very difficult because of their aggressiveness. The technique used consists in brazing the thermocouples to relatively large stainless steel parts and then joining these plugs by electron bombardment welding to diffused stainless steel-zircaloy couplings. The properties of these diffused couplings and of the brazed joints were studied; the various stages in the fabrication of the containers are also described. (authors) [fr

Reaction of tellurium with Zircaloy-4

International Nuclear Information System (INIS)

Boer, R. de; Cordfunke, E.H.P.

1994-09-01

Interaction of tellurium vapour with Zircaloy during the initial stage of an accident will lead to retention of tellurium in the core. For reliable estimation of the release behaviour of tellurium, it is necessary to know which zirconium tellurides are formed during this interaction. In this work the reaction of tellurium with Zircaloy-4 has been studied, using various reaction temperatures and tellurium vapour pressures. The compound ZrTe 2-x is formed on the surface of the Zircaloy in a broad range of reaction temperatures and vapour pressures. It is found that the formation of the more zirconium-rich compound Zr 5 Te 4 is favoured at high reaction temperatures is combination with low tellurium vapour pressures. (orig.)

Investigation on Nd:YAG laser weldability of zircaloy-4 end cap closure for nuclear fuel elements

International Nuclear Information System (INIS)

Kim, Soo Sung; Lee, Chul Yung; Yang, Myung Seung

2001-01-01

Various welding processes are now available for end cap closure of nuclear fuel element such as TIG(Tungsten Inert Gas) welding, magnetic resistance welding and laser welding. Even though the resistance and TIG welding processes are widely used for manufacturing commercial fuel elements, they can not be recommended for the remote seal welding of a fuel element at a hot cell facility due to the complexity of electrode alignment, difficulty in the replacement of parts in the remote manner and a large heat input for a thin sheath. Therefore, the Nd:YAG laser system using optical fiber transmission was selected for Zircaloy-4 end cap welding inside hot cell. The laser welding apparatus was developed using a pulsed Nd:YAG laser of 500 watt average power with optical fiber transmission. The weldability of laser welding was satisfactory with respect to the microstructures and mechanical properties comparing with TIG and resistance welding. The optimum operation processes of laser welding and the optical fiber transmission system for hot cell operation in a remote manner have been developed. The effects of irradiation on the properties of the laser apparatus were also being studied

Development of laser welded appendages to Zircaloy-4 fuel tubing (sheath/cladding)

Energy Technology Data Exchange (ETDEWEB)

Livingstone, S., E-mail: steve.livingstone@cnl.ca [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Xiao, L. [Canadian Nuclear Laboratories Limited, Chalk River, ON, Canada K0J 1J0 (Canada); Corcoran, E.C.; Ferrier, G.A.; Potter, K.N. [Department of Chemistry and Chemical Engineering, Royal Military College of Canada, Kingston, ON, Canada K7K 7B4 (Canada)

2015-04-01

Highlights: • Examines feasibility of laser welding appendages to Zr-4 tubing. • Laser welding minimizes the HAZ and removes toxic Be. • Mechanical properties of laser welds appear competitive with induction brazed joints. • Work appears promising and lays the foundation for further investigations. - Abstract: Laser welding is a potential alternative to the induction brazing process commonly used for appendage attachment in CANDU{sup ®} fuel fabrication that uses toxic Be as a filler metal, and creates multiple large heat affected zones in the sheath. For this work, several appendages were laser welded to tubing using different laser heat input settings and then examined with a variety of techniques: visual examination, metallography, shear strength testing, impact testing, and fracture surface analysis. Where possible, the examination results are contrasted against production induction brazed joints. The work to date looks promising for laser welded appendages. Further work on joint optimization, corrosion testing, irradiation testing, and post-irradiation examination will be performed in the future.

Determination of I-SCC crack propagation rate of zircaloy-4

International Nuclear Information System (INIS)

Woo-Seog, Ryu

2002-01-01

Threshold stress intensity (K ISCC ) and propagation rate of iodine-induced SCC in recrystallized and stress-relieved Zircaloy-4 were determined using a DCPD method. Dynamic system flowing Ar gas through iodine chamber at 60 deg C provided a constant iodine pressure of 1000 Pa during test. The SCC curves of crack velocity vs. stress intensity showed the typical SCC curves that are composed of stages I, II and III. The threshold K ISCC at 350 deg C was about 9 and 9.5 MPa √m for the stress- relieved Zircaloy-4 and the recrystallized Zircaloy-4, respectively. The plateau velocity in the stage II at 350 deg C was 4-8x 10 -4 mm/sec in the range of 20-40 MPa√m. In comparison with recrystallized Zircaloy-4, stress-relieved Zircaloy-4 had a lower threshold stress intensity factor and a little higher SCC velocity, indicating that SRA Zircaloy-4 was more sensitive to SCC in respect of velocity. The fracture mode in recrystallized Zircaloy was mostly a transgranular fracture with river pattern. An intergranular mode and the flutting were scarcely observed. (author)

Process for reliewing stresses in a zircaloy 2 or zircaloy 4 strip

International Nuclear Information System (INIS)

Charquet, D.; Dombre, M.

1986-01-01

Fabrication process of a zircaloy 2 or zircaloy 4 strip with an oxygen content between 900 and 1600 ppm with the following mechanical properties: E0.2≥250MPa at 315 deg C, parallel and perpendicular A% ≥4 at 20 deg C. The strip is rolled and stabilized by heat treatment between 490 and 580 deg C for 1 to 10 minutes and partially recrystallized for 0.5 to 5 vol.%. It is used for spacers of nuclear fuels [fr

Hydrogen pickup and redistribution in alpha-annealed Zircaloy-4

International Nuclear Information System (INIS)

Kammenzind, B.F.; Franklin, D.G.; Duffin, W.J.; Peters, H.R.

1996-01-01

Zircaloy-4, which is widely used as a core structural material in Pressurized-Water Reactors (PWR), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and hydrides precipitate after the Zircaloy-4 matrix becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4. To study hydrogen pickup and concentration, a postirradiation nondestructive radiographic technique for measuring hydrogen concentration was developed and qualified. Experiments on hydrogen pickup were conducted in the Advanced Test Reactor (ATR). Ex-reactor tests were conducted to determine the conditions for which hydrogen would dissolve, migrate, and precipitate. Finally, a phenomenological model for hydrogen diffusion was indexed to the data. This presentation describes the equipment and the model, presents the results of experiments, and compares the model predictions to experimental results

Creep damage in zircaloy-4 at LWR temperatures

International Nuclear Information System (INIS)

Keusseyan, R.L.; Hu, C.P.; Li, C.Y.

1978-08-01

The observation of creep damage in the form of grain boundary cavitation in Zircaloy-4 in the temperature range of interest to Light Water Reactor (LWR) applications is reported. The observed damage is shown to reduce the ductility of Zircaloy-4 in a tensile test at LWR temperatures

Hydrogen terminal solubility in Zircaloy-4

International Nuclear Information System (INIS)

Vizcaino, Pablo; Banchik, Abrahan D.

1999-01-01

Terminal solubility temperature of hydrogen in zirconium and its alloys is an important parameter because hydrides precipitation embrittled these materials making them susceptible to the phenomenon known as retarded hydrogen cracking. This work continues the study presented in the 25 AATN Meeting. Within this framework, a study focused on determining these curves in recrystallized Zircaloy-4, using scanning differential calorimetric technique. Terminal solubility curves for Zircaloy-4 were constructed within a concentration range from 40 to 640 ppm in hydrogen weight and comparisons with results obtained by other authors were made. (author)

Zircaloy sheathed thermocouples for PWR fuel rod temperature measurements

International Nuclear Information System (INIS)

Anderson, J.V.; Wesley, R.D.; Wilkins, S.C.

1979-01-01

Small diameter zircaloy sheathed thermocouples have been developed by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Surface mounted thermocouples were developed to measure the temperature of zircaloy clad fuel rods used in the Thermal Fuels Behavior Program (TFBP), and embedded thermocouples were developed for use by the Loss-of-Fluid Test (LOFT) Program for support tests using zircaloy clad electrically heated nuclear fuel rod simulators. The first objective of this developmental effort was to produce zircaloy sheathed thermocouples to replace titanium sheathed thermocouples and thereby eliminate the long-term corrosion of the titanium-to-zircaloy attachment weld. The second objective was to reduce the sheath diameter to obtain faster thermal response and minimize cladding temperature disturbance due to thermocouple attachment

Investigations of the interaction between ballooning Zircaloy cladding and emergency core cooling

International Nuclear Information System (INIS)

Wiehr, K.; Barth, S.; Erbacher, F.; Hame, W.; Harten, U.; Just, W.; Megerle, A.; Mueller, S.; Neitzel, H.J.; Reimann; Schaeffner, P.; Schmidt, H.

1975-01-01

The development of fabrication methods for the production of fuel rod simulators has been largely terminated. For welding of Zircaloy-4 and inconel 600 explosive welding has proved to be promissory in preliminary tests. A prototype fuel rod simulator was tested at full power. Its performance was faultless and the fuel rod and ring pellets could be easily dismantled and reused after the experiment. Planning of the test rig and electricity supply were terminated. Most of the assembly work has been finished. For electric heating of the fuel rod simulators a special device was built and tested which allows to program the power control. The radiographic system recording ballooning of the Zircaloy clad was erected outside the test space and put into operation. First trial pictures yielded good results. (orig.) [de

Oxide thickness measurement technique for duplex-layer Zircaloy-4 cladding

International Nuclear Information System (INIS)

McClelland, R.G.; O'Leary, P.M.

1992-01-01

Siemens Nuclear Power Corporation (SNP) is investigating the use of duplex-layer Zircaloy-4 tubing to improve the waterside corrosion resistance of cladding for high-burnup pressurized water reactor (PWR) fuel designs. Standard SNP PWR cladding is typically 0.762-mm (0.030-in.)-thick Zircaloy-4. The SNP duplex cladding is nominally 0.660-mm (0.026-in.)-thick Zircalloy-4 with an ∼0.102-mm (0.004-in.) outer layer of another, more corrosion-resistant, zirconium-based alloy. It is common industry practice to monitor the in-reactor corrosion behavior of Zircaloy cladding by using an eddy-current 'lift-off' technique to measure the oxide thickness on the outer surface of the fuel cladding. The test program evaluated three different cladding samples, all with the same outer diameter and wall thickness: Zircaloy-4 and duplex clad types D2 and D4

Hydriding and neutron irradiation in zircaloy-4

International Nuclear Information System (INIS)

Ramos, Ruben Fortunato; Martin, Juan Ezequiel; Orellano, Pablo; Dorao, Carlos; Analia Soldati; Ghilarducci, Ada Albertina; Corso, Hugo Luis; Peretti, Hernan Americo; Bolcich, Juan Carlos

2003-01-01

The composition of Zircaloy-4 for nuclear applications is specified by the ASTM B350 Standard, that fixes the amount of alloying elements (Sn, Fe, Cr) and impurities (Ni, Hf, O, N, C, among others) to optimize good corrosion and mechanical behavior.The recycling of zircaloy-4 scrap and chips resulting from cladding tube fabrication is an interesting issue.However, changes in the final composition of the recycled material may occur due to contamination with tool pieces, stainless steel chips, turnings, etc. while scrap is stored and handled. Since the main components of the possible contaminants are Fe, Cr and Ni, it arises the interest in studying up to what limit the Fe, Ni and Cr contents could be exceeded beyond the standard specification without affecting significantly the alloy properties.Zircaloy-4 alloys elaborated with Fe, Cr and Ni additions and others of standard composition in use in nuclear plants are studied by tensile tests, SEM observations and EDS microanalysis.Some samples are tested in the initial condition and others after hydriding treatments and neutron irradiation in the RA6

Study on kinetic of strain-aging in zircaloy-4

International Nuclear Information System (INIS)

Gomes, P.A.

1977-01-01

The strain-aging in zircaloy-4 has been investigated in this work and a study of the general problems involving this phenomenon has been realized in Zirconium and its alloys. It has been verified that a yield point appears in the Zircaloy-4, when it is submitted to strain-aging treatment between the temperatures 200 0 C and 400 0 C. (author)

Influence of hydrogen on the oxygen solubility in Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Guilbert-Banti, Séverine, E-mail: severine.guilbert@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France); Lacote, Pauline; Taraud, Gaëlle [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France); Berger, Pascal [NIMBE, CEA, CNRS, Université Paris-Saclay, 91191 Gif-sur-Yvette (France); Desquines, Jean; Duriez, Christian [Institut de Radioprotection et de Sûreté Nucléaire, SEREX/LE2M, Bâtiment 327, BP3, 13115 Saint Paul lez Durance (France)

2016-02-15

Despite the influence of hydrogen on the behavior of zirconium fuel cladding in many nuclear safety issues, the pseudo-binary Zircaloy-4 – oxygen phase diagram still lacks of data, especially above 1000 °C. The aim of this study was to provide experimental data to better assess the influence of hydrogen on the oxygen solubility in Zircaloy-4. Homogenized two-phase Zircaloy-4 samples were elaborated from 300 to 1000 wppm pre-hydrided samples. Local distributions were characterized thoroughly using Electron Probe Micro-Analysis (EPMA) for oxygen and Elastic Recoil Detection Analysis (ERDA) for hydrogen. The data obtained in this work were included in the pseudo-binary Zircaloy-4 – oxygen phase diagram and have shown that hydrogen has limited influence on the α + β → β transus. Regarding the α → α + β transus, no influence of hydrogen concentration in the α phase below 400 wppm was evidenced.

Surface analytical investigations of the thermal behaviour of passivated Zircaloy-4 surfaces and of the reaction behaviour of iodine with Zircaloy-4 surfaces

International Nuclear Information System (INIS)

Kaufmann, R.

1988-07-01

In the first part of the present work the thermal behaviour of atmospherically oxidized Zircaloy-4 samples was investigated at various temperatures. In a next step the amount of iodine adsorbed at the metallic surface was determined as well at room temperature with varying iodine exposures as for constant exposure but varying temperatures. Furthermore, the zirconium iodide species resulting from the interaction of iodine with the Zircaloy-4 and desorbed at higher temperatures were identified by means of residual gas analysis. During these studies it was found that the oxidic overlayer of the passivated Zircaloy-4 samples is decomposed at temperatures above 200 0 C. The iodine uptake at metallic surfaces (cleaned by Ar-ion sputtering) at room temperature slows markedly down after formation of a closed zirconium-iodide overlayer and consequently the further reaction proceeds diffusion-controlled. At 200 0 C ZrI 4 is formed being the thermodynamically most stable Zr-iodide. During desorption experiments using iodine exposed Zircaloy-4 samples the release of ZrI 4 was proved. The results obtained from the various experiments are finally discussed with respect to the iodine-induced stress corrosion cracking process and the underlying basic mechanisms and a transport mechanism for the SCC in nuclear fuel rods is proposed. (orig./RB) [de

Formation of Lamellar Structured Oxide Dispersion Strengthening Layers in Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Jung, Yang-Il; Park, Jung-Hwan; Park, Dong-Jun; Kim, Hyun-Gil; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lim, Yoon-Soo [Hanbat National University, Daejeon (Korea, Republic of)

2016-10-15

Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures. According to our previous investigations, the tensile strength of Zircaloy-4 was increased by up to 20% with the formation of a thin dispersed oxide layer with a thickness less than 10% of that of the Zircaloy-4 substrate. However, the tensile elongation of the samples decreased drastically. The brittle fracture was a major concern in development of the ODS Zircaloy-4. In this study, a lamellar structure of ODS layer was formed to increase ductility of the ODS Zircaloy-4. The mechanical properties were varied depending on the structure of ODS layer. For example, the partial formation of ODS layer with the thickness of 10% to the substrate thickness induced the increase in tensile strength up to about 20% than fresh Zircaloy-4.

Influence of temperature on the Zircaloy-4 plastic anisotropy; Influence de la temperature sur l`anisotropie plastique du Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Limon, R.; Bechade, J.L.; Lehmann, S.; Maury, R.; Soniak, A. [CEA Centre d`Etudes de Saclay, 91 - Gif-sur-Yvette (France). Direction des Technologies Avancees; Mardon, J.P. [Societe Franco-Americaine de Constructions Atomiques (FRAMATOME), 92 - Paris-La-Defense (France)

1995-12-31

In order to improve the comportment modelling of PWR fuel pin, and more precisely their canning tubes, Framatome and the CEA have undertake an important study program of Zircaloy-4 mechanical properties. It includes in particular the study of the plasticity between 20 and 400 degree Celsius. This material being not isotropic because of the zirconium hexagonal crystal network and the texture presented by the canning tubes, its plastic anisotropy has been measured. The obtained results for the canning in *slack* and recrystallized before irradiation Zircaloy-4 are presented and the deformation systems able to explain the observed anisotropy is researched. (O.L.). 6 refs., 4 figs., 1 tab.

Oligo cyclic plastic fatigue of Zircaloy-4 under vacuum and in iodinated methanol; Fatigue plastique oligocyclique du Zircaloy-4 sous vide et dans le methanol iode

Energy Technology Data Exchange (ETDEWEB)

Beloucif, A.

1995-01-01

Our study was bound to the Zircaloy-4 fuel can damage in PWR type reactors. The topic was the damage mechanisms of Zircaloy-4 by oligo-cyclic plastic fatigue in inert atmosphere and in iodinated methanol. The oligo-cyclic plastic fatigue tests, under vacuum, were performed with steady plastic deformation and deformation speed. The corrosion fatigue tests in iodinated methanol put to the fore one obvious harmful part of iodine on Zircaloy-4 resistance to cyclic solicitations. The observations proved the existence of a very strong synergic effect between cyclic mechanical damage and corrosion. (MML). 84 refs., 117 figs., 3 tabs.

Development of a new bundle welding technology for CANDU fuels

International Nuclear Information System (INIS)

Kim, Soo Sung; Lee, D. Y.; Goo, D. S.

2010-01-01

The new technology of welding process for fuel bundle of CANDU nuclear fuels is considered important in respect to the soundness of weldments and the improvement of the performance of nuclear fuels during the operation in reactor. The probability of leakage of the fission products is mostly apt to occur at the weldments of fuel bundles, and it is connected directly with the safety and life prediction of the nuclear reactor in operation. The fuel bundles of CANDU nuclear fuels are welded by the electrical resistance method, connecting the endplates and endcaps with fuel rods. Therefore, the purpose of this study of the 2nd year is to select the proper welding parameters and to investigate the characteristics of the full-sized samples using the projection endplates and make some prototype samples for the endplate welding of CANDU nuclear fuels. This study will be also provide the fundamental data for the new design and fabrications of CANDU nuclear fuel bundles

Semi-empirical corrosion model for Zircaloy-4 cladding

International Nuclear Information System (INIS)

Nadeem Elahi, Waseem; Atif Rana, Muhammad

2015-01-01

The Zircaloy-4 cladding tube in Pressurize Water Reactors (PWRs) bears corrosion due to fast neutron flux, coolant temperature, and water chemistry. The thickness of Zircaloy-4 cladding tube may be decreased due to the increase in corrosion penetration which may affect the integrity of the fuel rod. The tin content and inter-metallic particles sizes has been found significantly in the magnitude of oxide thickness. In present study we have developed a Semiempirical corrosion model by modifying the Arrhenius equation for corrosion as a function of acceleration factor for tin content and accumulative annealing. This developed model has been incorporated into fuel performance computer code. The cladding oxide thickness data obtained from the Semi-empirical corrosion model has been compared with the experimental results i.e., numerous cases of measured cladding oxide thickness from UO 2 fuel rods, irradiated in various PWRs. The results of the both studies lie within the error band of 20μm, which confirms the validity of the developed Semi-empirical corrosion model. Key words: Corrosion, Zircaloy-4, tin content, accumulative annealing factor, Semi-empirical, PWR. (author)

Thermal creep of Zircaloy-4 cladding

International Nuclear Information System (INIS)

Murty, K.L.; Clevinger, G.S.; Papazoglou, T.P.

1977-01-01

Data on the hoop creep characteristics of Zircaloy tubing were collected at temperatures between 600 F and 800 F, and at stress levels ranging from 10 ksi to 25 ksi using internal pressurization tests. At low driving forces, exposures as long as 2000 hours were found insufficient to establish steady state creep. The experimental data at temperatures of 650 F to 800 F correlate well with an exponential stress dependence, and the activation energy for creep was found to be in excellent agreement with that for self-diffusion. The range of stresses and temperatures is too small to study the overall effect of these variables on the activation energy for creep. The experimental steady state creep-rates and those predicted from the creep equation used agree within a factor of 1.3. These correlations imply that the mechanism for hoop creep of Zircaloy-4 cladding is characterized by an activation energy of approximately 60 kcal/mole and an activation area of about 20b 3 . In addition, the exponential stress dependence implies that the activation area for creep is stress-independent. These results suggest that the climb of edge dislocations is the rate controlling mechanism for creep of Zircaloy-4. The transient creep regime was also analysed on the premise that primary creep is directly related to the rate of dispersal of dislocation entanglements by climb. (Auth.)

Recovery and recrystallisation of zircaloy-4

International Nuclear Information System (INIS)

Derep, J.L.; Rouby, D.; Fantozzi, G.

1981-01-01

Examination of the three mechanisms that control the recovery of zircaloy-4 workhardened by rolling: polygonisation leading to a cellular structure, annihilation of dislocations of opposite sign producing thinning of the cell walls, and growth of subgrains by coalescence [fr

Improvements in and relating to welding

International Nuclear Information System (INIS)

Taylor, B.D.

1979-01-01

This invention concerns apparatus for use in welding, particularly welding which must be effected in a predetermined, for example, inert atmosphere, e.g. the welding of reactive materials such as zircaloy, titanium, magnesium, aluminium, etc. (U.K.)

High-temperature oxidation of Zircaloy-2 and Zircaloy-4 in steam

International Nuclear Information System (INIS)

Urbanic, V.F.; Heidrick, T.R.

1978-01-01

At temperatures above the (α + β)/β transformation temperature for zirconium alloys, steam reacts with β-Zr to form a superficial layer of zirconium oxide (ZrO 2 ) and an intermediate layer of oxygen-stabilized α-Zr. Reaction kinetics and the rate of growth of the combined (ZrO 2 + α-Zr) layer for Zircaloy-2 and Zircaloy-4 oxidation in steam were measured over the temperature range 1050-1850 o C. The reaction rates for both alloys were similar, obeyed parabolic kinetics and were not limited by gas phase diffusion. The parabolic rate constants were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam. A discontinuity was found in the temperature dependence of both the reaction rate and the rate of growth of the combined (ZrO 2 + α-Zr) layer. The discontinuity is attributed to a change in the oxide microstructure at the discontinuity temperature, an observation which is consistent with the zirconium-oxygen phase diagram. (author)

Microstructural characterization of second phase irradiated Zircaloy-4 particles

International Nuclear Information System (INIS)

Flores, Alejandra V.; Vizcaino, Pablo; Banchik, Abraham D.; Bozzano, Patricia B.; Versaci, Raul A.

2007-01-01

X-ray diffraction diagrams of neutron irradiated Zircaloy-4 were obtained at the LNLS with the aim to obtain bulk information about the amorphization process in which the Zircaloy-4 second phase particles (SPPs) undergoes due to neutron irradiation. Owing to the low concentration of the SPPs in the alloy (∼0.4 V %), no data regarding to the bulk were obtained until now. The synchrotron experiences allowed to detect five of the more intense lines of the phase C 14 (SPPs structure) in unirradiated Zircaloy-4: (110) θ, (103) θ, (112) θ, (201) θ and (004) θ in the 34 degrees ≤ θ2≤45 degrees Bragg angle range and others of minor intensity. The diagrams of the samples irradiated at moderate doses (1020n/cm 2 ) show these lines even in the as received samples. In contrast, none of these lines are observed for high fluence samples (∼1022neutrons/cm 2 ). In addition, in similar high fluence samples annealed 24 h or 72 h at 600 C degrees the intensity rises just at the 2q range where the C 14 lines were observed, showing a wide peak. That peak is interpreted as a result of the superposition of unresolved diffraction lines corresponding to the Zircaloy SPPs which are in a reconstitution process of crystallization. Analytical Electron Microscopy techniques were used, in order to study the effects on the Zircaloy-4 SPPs and compared with samples of the same material without irradiation. Spots in SAD patterns of non irradiated SPPS, evidences the presence of a C 14 structure, but in irradiated SSP SAD patterns evidences the beginning of an amorphization process. Another important feature to point out is the different Fe / Cr ratio presented in both irradiated and non irradiated SSPs. In non irradiated precipitates the Fe / Cr ratio is approximately 1.5, while in irradiated precipitates the Fe / Cr ratio becomes near 1.0. (author) [es

Reactive endplate marrow changes: a systematic morphologic and epidemiologic evaluation

International Nuclear Information System (INIS)

Karchevsky, Michael; Schweitzer, Mark E.; Carrino, John A.; Zoga, Adam; Montgomery, Douglas; Parker, Laurence

2005-01-01

To evaluate the morphology and location of vertebral endplate changes, and to analyze their association with age, gender, and body mass index (BMI). At 1.5 T (T1-weighted, T2-weighted/STIR) 100 lumbar spines were evaluated separately by three observers. The readers classified the endplate bone marrow abnormalities on sagittal MR images according to the definitions of Modic et al. Findings were localized by disc segment; whether in the upper and/or lower endplate; and within each endplate divided into 15 segments. Disc space narrowing, as well as disc desiccation, was also noted at each vertebral level. In addition, endplate changes were correlated with age, gender, and BMI (weight(kg)/height(m) 2 ). A total of 15,000 data points were studied and 422 total changes recorded. A total of 99 vertebral levels were affected in 58 patients. Of these, 171 were of type I, 242 were of type II, and 9 were of type III. L4 - L5 and L5 - S1 vertebral levels were most commonly involved, having (142, 4.73%) and (116, 3.87%) changes respectively (P<0.0001). The upper and lower aspects of the endplate were affected similarly. Changes most frequently occurred at the anterior aspect of the endplate (P<0.0001). Endplate marrow changes were associated with increasing age (P<0.0001) and, surprisingly, male gender (P<0.0001). Endplate changes were not associated with BMI. The fatty pattern was most common, with the sclerotic pattern being rare. Endplate marrow changes most often occurred at the anterior aspect of the endplate, particularly at L4 - L5 and L5 - S1 levels. Modic changes occur more frequently with aging, evidence of their degenerative etiology. They were, however, not related to body habitus, but to weight and male gender. (orig.)

Reactive endplate marrow changes: a systematic morphologic and epidemiologic evaluation

Energy Technology Data Exchange (ETDEWEB)

Karchevsky, Michael [Hahnemann University Hospital, Department of Radiology, Philadelphia (United States); Thomas Jefferson University Hospital, Department of Radiology, Philadelphia (United States); Schweitzer, Mark E. [Hospital for Joint Diseases, Department of Radiology, New York (United States); Carrino, John A.; Zoga, Adam; Montgomery, Douglas; Parker, Laurence [Thomas Jefferson University Hospital, Department of Radiology, Philadelphia (United States)

2005-03-01

To evaluate the morphology and location of vertebral endplate changes, and to analyze their association with age, gender, and body mass index (BMI). At 1.5 T (T1-weighted, T2-weighted/STIR) 100 lumbar spines were evaluated separately by three observers. The readers classified the endplate bone marrow abnormalities on sagittal MR images according to the definitions of Modic et al. Findings were localized by disc segment; whether in the upper and/or lower endplate; and within each endplate divided into 15 segments. Disc space narrowing, as well as disc desiccation, was also noted at each vertebral level. In addition, endplate changes were correlated with age, gender, and BMI (weight(kg)/height(m){sup 2}). A total of 15,000 data points were studied and 422 total changes recorded. A total of 99 vertebral levels were affected in 58 patients. Of these, 171 were of type I, 242 were of type II, and 9 were of type III. L4 - L5 and L5 - S1 vertebral levels were most commonly involved, having (142, 4.73%) and (116, 3.87%) changes respectively (P<0.0001). The upper and lower aspects of the endplate were affected similarly. Changes most frequently occurred at the anterior aspect of the endplate (P<0.0001). Endplate marrow changes were associated with increasing age (P<0.0001) and, surprisingly, male gender (P<0.0001). Endplate changes were not associated with BMI. The fatty pattern was most common, with the sclerotic pattern being rare. Endplate marrow changes most often occurred at the anterior aspect of the endplate, particularly at L4 - L5 and L5 - S1 levels. Modic changes occur more frequently with aging, evidence of their degenerative etiology. They were, however, not related to body habitus, but to weight and male gender. (orig.)

Influence of temperature on the Zircaloy-4 plastic anisotropy

International Nuclear Information System (INIS)

Limon, R.; Bechade, J.L.; Lehmann, S.; Maury, R.; Soniak, A.

1995-01-01

In order to improve the comportment modelling of PWR fuel pin, and more precisely their canning tubes, Framatome and the CEA have undertake an important study program of Zircaloy-4 mechanical properties. It includes in particular the study of the plasticity between 20 and 400 degree Celsius. This material being not isotropic because of the zirconium hexagonal crystal network and the texture presented by the canning tubes, its plastic anisotropy has been measured. The obtained results for the canning in *slack* and recrystallized before irradiation Zircaloy-4 are presented and the deformation systems able to explain the observed anisotropy is researched. (O.L.). 6 refs., 4 figs., 1 tab

Cyclic deformation of zircaloy-4 at room temperature

International Nuclear Information System (INIS)

Armas, A. F; Herenu, S; Bolmaro, R; Alvarez-Armas, I

2003-01-01

Annealed materials hardens under low cyclic fatigue tests.However, FCC metals tested with medium strain amplitudes show an initial cyclic softening.That behaviour is related with the strong interstitial atom-dislocation interactions.For HCP materials the information is scarce.Commercial purity Zirconium and Zircaloy-4 alloys show also a pronounced cyclic softening, similar to Titanium alloys.Recently the rotation texture induced softening model has been proposed according to which the crystals are placed in a more favourable deformation orientation by prismatic slip due to the cyclic strain.The purpose of the current paper is the presentation of decisive results to discuss the causes for cyclic softening of Zircaloy-4. Low cycle fatigue tests were performed on recrystallized Zircaloy-4 samples.The cyclic behaviour shows an exponential softening at room temperature independently of the deformation range.Only at high temperature a cyclic hardening is shown at low number of cycles.Friction stresses, related with dislocation movement itself, and back stresses, related with dislocation pile-ups can be calculated from the stress-strain loops.The cyclic softening is due to diminishing friction stress while the starting hardening behaviour is due to increasing back stresses.The rotation texture induced softening model is ruled out assuming instead a model based on dislocation unlocking from interstitial oxygen solute atoms

Zircaloy 4 ingots' industrial fabrication

International Nuclear Information System (INIS)

Leyt, A.

1987-01-01

The technology developed for the industrial fabrication of Zircaloy-4 ingots is presented. According to the results obtained: a) the homogeneity of the ingots is analyzed, regarding the distribution of components (tin, iron, chromium, oxygen) and Brinell hardness as a function of different types of charge: zirconium sponge-recycling alloy material, sponge of zirconium-alloy; b) the distribution of the same parameters as a function of production is also analyzed. (Author)

Development of zircaloy deformation model to describe the zircaloy-4 cladding tube during accidents

International Nuclear Information System (INIS)

Raff, S.

1978-01-01

The development of a high-temperature deformation model for Zircaloy-4 cans is primarily based on numerous well-parametrized tensile tests to get the material behaviour including statistical variance. It is shown that plastic deformation may be described by a power creep law, the coefficients of which show strong dependence on temperature in the relevant temperature region. These coefficients have been determined. A model based on these coefficients has been established which, apart from best estimate deformation, gives upper and lower bounds of possible deformation. The model derived from isothermal uniaxial tests is being verified against isothermal and transient tube burst tests. The influence of preoxidation and increased oxygen concentration during deformation is modeled on the basis of the pseudobinary Zircaloy-oxygen phase diagram. (author)

Characterisation of metallic glass incorporated Zircaloy-2 weldments

International Nuclear Information System (INIS)

Mishra, S.; Savalia, R.T.; Bhanumurthy, K.; Dey, G.K.; Banerjee, S.

1995-01-01

In this study the effect of incorporation of Zr based Fe and Ni bearing metallic glass in spot welds in Zircaloy components has been examined. A comparison of strength and microstructure of the welded joint with and without glass has been carried out. The welded joint with metallic glass has been found to be stronger than the one without metallic glass. The microstructure of the welded region with metallic glass has been found to comprise a large region having martensite. This large martensitic region has also been found to have considerable amount of excess solute (Fe, Ni). The higher strength of the weld with metallic glass seems to originate due to solid solution strengthening, small grain size and the presence of martensitic structure over a large region. (orig.)

Octanol reduces end-plate channel lifetime

Science.gov (United States)

Gage, Peter W.; McBurney, Robert N.; Van Helden, Dirk

1978-01-01

1. Post-synaptic effects of n-octanol at concentrations of 0·1-1 mM were examined in toad sartorius muscles by use of extracellular and voltage-clamp techniques. 2. Octanol depressed the amplitude and duration of miniature end-plate currents and hence depressed neuromuscular transmission. 3. The decay of miniature end-plate currents remained exponential in octanol solutions even when the time constant of decay (τD) was decreased by 80-90%. 4. The lifetime of end-plate channels, obtained by analysis of acetylcholine noise, was also decreased by octanol. The average lifetime measured from noise spectra agreed reasonably well with the time constant of decay of miniature end-plate currents, both in control solution and in octanol solutions. 5. Octanol caused a reduction in the conductance of end-plate channels. Single channel conductance was on average about 25 pS in control solution and 20 pS in octanol. 6. In most cells the normal voltage sensitivity of the decay of miniature end-plate currents was retained in octanol solutions. The lifetime of end-plate channels measured from acetylcholine noise also remained voltage-sensitive in octanol solutions. In some experiments in which channel lifetime was exceptionally reduced the voltage sensitivity was less than normal. 7. In octanol solutions, τD was still very sensitive to temperature changes in most cells although in some the temperature sensitivity of τD was clearly reduced. Changes in τD with temperature could generally be fitted by the Arrhenius equation suggesting that a single step reaction controlled the decay of currents both in control and in octanol solutions. In some cells in which τD became less than 0·3 ms, the relationship between τD and temperature became inconsistent with the Arrhenius equation. 8. As the decay of end-plate currents in octanol solutions remains exponential, and the voltage and temperature sensitivity can be unchanged even when τD is significantly reduced, it seems likely that

Out-of-pile UO2/Zircaloy-4 experiments under severe fuel damage conditions

International Nuclear Information System (INIS)

Hofmann, P.

1983-01-01

Chemical interactions between UO 2 fuel and Zircaloy-4 cladding up to the melting point of zircaloy (Zry) are described. Out-of-pile UO 2 /zircaloy reaction experiments have been performed to investigate the chemical interaction behavior under possible severe fuel damage conditions (very high temperatures and external overpressure). The tests have been conducted in inert gas (1 to 80 bar) with 10-cm-long zircaloy cladding specimens filled with UO 2 pellets. The annealing temperature varied between 1000 and 1700 deg. C and the annealing period between 1 and 150 min. The extent of the chemical reaction depends decisively on whether or not good contact between UO 2 and zircaloy has been established. If solid contact exists, zircaloy reduces the UO 2 to form oxygen-stabilized α-Zr(O) and uranium metal. The uranium reacts with zircaloy to form a (U,Zr) alloy rich in uranium. The (U,Zr) alloy, which is liquid above approx. 1150 deg. C, lies between two α-Zr(O) layers. The UO 2 /zircaloy reaction obeys a parabolic rate law. The degree of chemical interaction is determined by the extent of oxygen diffusion into the cladding, and hence by the time and temperature. The affinity of zirconium for oxygen, which results in an oxygen gradient across the cladding, is the driving force for the reaction. The growth of the reaction layers can be represented in an Arrhenius diagram. The UO 2 /Zry-4 reaction occurs as rapidly as the steam/Zry-4 reaction above about 1100 deg. C. The extent of the interaction is independent of external pressure above about 10 bar at 1400 deg. C and 5 bar at 1700 deg. C. The maximum measured oxygen content of the cladding is approx. 6wt.%. Up to approx. 9 volume % of the UO 2 can be chemically dissolved by the zircaloy. In an actual fuel rod, complete release of the fission products in this region of the fuel must therefore be assumed. (author)

A technical report on the evaluation of the integrity for the TIG welded spacer grid

International Nuclear Information System (INIS)

Song, Kee Nam; Yoo, Ho Sik; Lww, Chang Woo

1994-07-01

The spacer grid, which supports fuel rods, guide thimble and instrumentation tube, is classified into two types according to their strap material,.ie. inconel and zircaloy spacer grid. KOFA fuel of 14 x 14 and 17 x 17 type has seven and eight spacer grid respectively. Zircaloy spacer grid is assembled by straps whose cross points are welded by TIG welding method. This technical report provides to give some information about structure and function of the spacer grid and the basis and characteristic of the TIG welding method. A series of test which is conducted to evaluate the integrity of TIG welded zircaloy spacer grid and their results have been also studied. (Author) 18 refs., 23 figs., 3 tabs

A technical report on the evaluation of the integrity for the TIG welded spacer grid

Energy Technology Data Exchange (ETDEWEB)

Song, Kee Nam; Yoo, Ho Sik; Lww, Chang Woo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-07-01

The spacer grid, which supports fuel rods, guide thimble and instrumentation tube, is classified into two types according to their strap material,.ie. inconel and zircaloy spacer grid. KOFA fuel of 14 x 14 and 17 x 17 type has seven and eight spacer grid respectively. Zircaloy spacer grid is assembled by straps whose cross points are welded by TIG welding method. This technical report provides to give some information about structure and function of the spacer grid and the basis and characteristic of the TIG welding method. A series of test which is conducted to evaluate the integrity of TIG welded zircaloy spacer grid and their results have been also studied. (Author) 18 refs., 23 figs., 3 tabs.

The oxidation kinetics of zircaloy - 4 under isothermal conditions

International Nuclear Information System (INIS)

Santos, A.M.M. dos; Cardoso, P.E.

1982-01-01

The oxidation kinetics of zircaloy-4 tubes was studied by means of isothermal tests in the temperature interval 500 0 C to 900 0 C. Dry oxygen and water steam, were used as oxidant agents. The results show that the oxidation kinetics law exhibits a behaviour from cubic to parabolic in the range of the time and temperatures of the experiment. Dry oxygen shows a stronger oxidation effect than water steam. A special mechanical test to study the embrittlement effect in the small samples of zircaloy tubes was used. (Author) [pt

Biaxial mechanical tests in zircaloy-4

International Nuclear Information System (INIS)

Mintzer, S.R.; Bordoni, R.A.A.; Falcone, J.M.

1980-01-01

The texture of the zircaloy-4 tubes used as cladding in nuclear fuel elements determines anisotropy of the mechanical properties. As a consequence, the uniaxial tests to determine the mechanical behaviour of the tubes are incomplete. Furthermore, the cladding in use is subject to creep with a state of biaxial tensions. For this reason it is also important to determine the biaxial mechanical properties. The creep tests were performed by internal pressure for a state of axial to circumferential tensions of 0.5. Among the experimental procedures are described: preparation of the test specimens, pressurizing equipment, and the implementation of a device that permits a permanent register of the deformation. For the non-irradiated Atucha type zircaloy-4 sheaths, experimental curves of circumferential deformation versus time were obtained, in tests at constant pressure and for different values of temperature and pressure. An empirical function was determined to adjust the experimental values for the speed of the circumferential deformation in terms of the initial tension applied, temperature and deformation, and the change of the corresponding parameters in accordance to the range of the tensions. Also the activation energy for creep was determined. (M.E.L.) [es

Endplates Changes Related to Age and Vertebral Segment

Directory of Open Access Journals (Sweden)

Carlos Fernando P. S. Herrero

2014-01-01

Full Text Available Endplate separations are defined as the presence of a space between the hyaline cartilage and the cortical bone of the adjacent vertebral body. This study evaluates endplate separations from the vertebral body and intervertebral discs and verifies if endplate separation is related to age and the spinal level. Groups were formed based on age (20–40 and 41–85 years old and the vertebral segment (T7-T8 and L4-L5 segments. Histological analysis included assessment of the length of the vertebral endplates, the number and dimensions of the separations, and orientation of the collagen fibers, in the mid-sagittal slice. Two indexes were created: the separation index (number of separations/vertebral length and separation extension index (sum of all separations/vertebral length. The results of the study demonstrated a direct relationship between the density of separations in the endplate and two variables: age and spinal level.

Spectrochemical determination of impurities in zircaloy 2 and 4

International Nuclear Information System (INIS)

Paula Reino, L.C. de; Lordello, A.R.

1987-06-01

A method has been developed for the determination of Hf,Co,Mo,Pb,Ti,V,Al,Si,W,Cu,Mg,Mn,B and Cd in zircaloy 2 and 4. For hafnium determination 10% CuF 2 is added as spectrographic buffer on a previously oxidized zircaloy; the samples are loaded in a shallow cup electrode of Scribner Mullins type and excited in a direct current arc. The carrier distillation technique has been used for the other elements. Better results were obtained with 25% AgCl as carrier. The precision of the method varies from 4% for copper to 29% for boron but it does not exceed 17% for most elements. (Author) [pt

Significance of angular mismatch between vertebral endplate and prosthetic endplate in lumbar total disc replacement.

Science.gov (United States)

Lee, Chong Suh; Chung, Sung Soo; Oh, Sung Kyun; You, Je Wook

2011-05-01

A retrospective study. To determine whether angular mismatch between the vertebral endplate and prosthetic endplate during lumbar total disc replacement (L-TDR) affects the radiological and clinical outcomes. A prosthesis anchored to the vertebral body by using a large central keel carries an inherent risk of angular mismatch between the vertebral endplate and prosthetic endplate at a segment with a greater degree of lordosis, such as L5-S1. Theoretically, this angular mismatch can cause several problems, such as segmental hyperlordosis, anterior positioning of the upper prosthesis, posterior prosthetic edge subsidence, decreased range of motion (ROM), and a poor clinical outcome. This study evaluated 64 prosthetic levels of 56 patients who were implanted with L-TDR between June 2002 and February 2006. There were 38 and 26 prosthetic levels at the L4-5 and L5-S1, respectively. The mean follow-up period was 25.6 (12 to 49) months. The angle of mismatch between the lower endplate of the upper vertebral body and the upper prosthetic plate, segmental flexion/extension ROM, segmental lordosis angle at extension, distance from the posterior wall of the vertebral body to the posterior prosthetic edge were measured by obtaining radiographs. Clinically, the Visual Analogue Scale and Oswestry Disability Index were also evaluated. The angular mismatches between the upper vertebra and prosthesis at L4-5 and L5-S1 were 1.6 degree and 5.6 degree, respectively (P body to the posterior edge of the prosthesis in L5-S1 were 6.8 degree (4 to 13), 12.8 degree (8 to 17), and 3.8 mm (1 to 6 mm) in patients with an angular mismatch of 0.05). Angular mismatch was more common in L5-S1 than in L4-5. L-TDR at the most lordotic level, L5-S1, and implantation of an upper prosthesis with a mismatched angle seem to be the causes of a reduced segmental ROM, increased segmental lordosis, and anterior malpositioning of the prosthesis. However, these changes do not affect the clinical outcomes of

Performance of indigenous resistance welding equipment for PHWR fuel fabrication in NFC

International Nuclear Information System (INIS)

Hemantha Rao, G.V.S.; Jayaraj, R.N.; Prakash, M.S.; Gupta, U.C.; Ganguly, C.

1999-01-01

Indigenisation of critical equipment for manufacturing of PHWR fuel and automation in the production line have been the main thrust in NFC in recent years. As part of this endeavour, resistance welding equipment for end plug welding of Zircaloy-4 clad Uranium Oxide fuel pin and end plates of 19-element fuel bundles have been developed. The paper discusses the equipment design features, critical operating parameters and performance of these indigenous welding machines. (author)

Quantification and characterization of zirconium hydrides in Zircaloy-4 by the image analysis method

International Nuclear Information System (INIS)

Zhang, J.H.; Groos, M.; Bredel, T.; Trotabas, M.; Combette, P.

1992-01-01

The image analysis method is used to determine the hydrogen content in specimens of Zircaloy-4. Two parameters, surface density of hydride, S v , and degree of orientation, Ω, are defined to represent separately the hydrogen content and the orientation of hydrides. By analysing the stress-relieved Zircaloy-4 specimens with known hydrogen content from 100 to 1000 ppm, a relationship is established between the parameter S v and the hydrogen content when the magnifications of the optical microscope are 1000 and 250. The degree of orientation for the hydride in the stress-relieved Zircaloy-4 cladding is about 0.3. (orig.)

PEM Low Cost Endplates. Final report

Energy Technology Data Exchange (ETDEWEB)

Nielsen, Martin; Clyens, S.; Steenstrup, F.R.; Christiansen, Jens [Danish Technological Institute. Plastics Technology, Taastrup (Denmark); Yde-Andersen, S. [IRD Fuel Cell A/S, Svendborg (Denmark)

2013-03-15

In the project, an endplate for the PEM-type fuel cells has been developed. The initial idea was to use an injection mouldable fibre reinforced polymer to produce the endplate and thereby exploit the opportunities of greater geometrical freedom to reduce weight and material consumption. Different PPS/glass-fibre compounds were produced and tested in order to use the results to optimize the results on the computer through FEM simulations. As it turned out, it was impossible to achieve adequate stiffness for the endplates within the given geometrical limitations. At the relatively high temperatures at which the endplates operate the material simply goes to soft. Material focus shifted to fibre reinforced high strength concrete composite. Test specimens were produced and tested so the results again could be used for FEM-simulations which also accounted for the technical limitations the concrete composite has regarding casting ability. In the process, the way the endplate is mounted was also alternated to better accommodate the properties of the concrete composite. A number of endplates were cast in specially produced moulds in order to map the optimum process parameters, and a final endplate was tested at IRD Fuel Cells A/S. The field test was in many aspects successful. However, the gas sealing and the surface finish can be further improved. The weight may still be an issue for some applications, even though it is lower than the endplate currently used. This issue can be addressed in a future project. The work has resulted in a new endplate design, which makes the stack assembly simpler and with fewer components. The endplates fabrication involves low cost methods, which can be scaled up as demand of fuel cells begin to take off. (Author)

Effect of a surface oxide-dispersion-strengthened layer on mechanical strength of zircaloy-4 tubes

Directory of Open Access Journals (Sweden)

Yang-Il Jung

2018-03-01

Full Text Available An oxide-dispersion-strengthened (ODS layer was formed on Zircaloy-4 tubes by a laser beam scanning process to increase mechanical strength. Laser beam was used to scan the yttrium oxide (Y2O3–coated Zircaloy-4 tube to induce the penetration of Y2O3 particles into Zircaloy-4. Laser surface treatment resulted in the formation of an ODS layer as well as microstructural phase transformation at the surface of the tube. The mechanical strength of Zircaloy-4 increased with the formation of the ODS layer. The ring-tensile strength of Zircaloy-4 increased from 790 to 870 MPa at room temperature, from 500 to 575 MPa at 380°C, and from 385 to 470 MPa at 500°C. Strengthening became more effective as the test temperature increased. It was noted that brittle fracture occurred at room temperature, which was not observed at elevated temperatures. Resistance to dynamic high-temperature bursting improved. The burst temperature increased from 760 to 830°C at a heating rate of 5°C/s and internal pressure of 8.3 MPa. The burst opening was also smaller than those in fresh Zircaloy-4 tubes. This method is expected to enhance the safety of Zr fuel cladding tubes owing to the improvement of their mechanical properties. Keywords: Laser Surface Treatment, Microstructure, Oxide Dispersion Strengthened Alloy, Tensile Strength, Zirconium Alloy

Development of remote laser welding technology

International Nuclear Information System (INIS)

Kim, Soo-Sung; Kim, Woong-Ki; Lee, Jung-Won; Yang, Myung-Seung; Park, Hyun-Soo

1999-01-01

Various welding processes are now available for end cap closure of nuclear fuel element such as TIG(Tungsten Inert Gas) welding, magnetic resistance welding and laser welding. Even though the resistance and TIG welding process are widely used for manufacturing of the commercial fuel elements, it can not be recommended for the remote seal welding of fuel element at PIE facility due to its complexity of the electrode alignment, difficulty in the replacement of parts in the remote manner and its large heat input for thin sheath. Therefore, Nd:YAG laser system using the optical fiber transmission was selected for Zircaloy-4 end cap welding. Remote laser welding apparatus is developed using a pulsed Nd:YAG laser of 500 watt average power with optical fiber transmission. The laser weldability is satisfactory in respect of the microstructures and mechanical properties comparing with the TIG and resistance welding. The optimum operation processes of laser welding and the optical fiber transmission system for hot cell operation in remote manner have been developed. (author)

Effect of the anodization variables in the corrosion resistence of the zircaloy-4 tubes

International Nuclear Information System (INIS)

Figueiredo, M.E.

1981-02-01

The anodization effect in the oxidation of the zircaloy-4 in steam atmosphere at 10,06MPa was investigated. It was also studied how the voltage and the types of electrolytes at several values of pH affect the growing of the anodic oxide film and the performance of the zircaloy-4 in relation to corrosion. Anodizations of zircaloy-4 tubes have been made with voltages ranging from zero to 280V and using electrolytic solutions of Na 2 B 4 O 7 , CH 3 COOH and NaOH in the concentrations of 1,0N, 0,1N and 0,01N. After anodization, the tubes were oxidized in autoclave under steam at 400 0 C and 10,06 MPa during 3 and 14 days. The results show that the anodization inhibit the oxidation process of zircaloy-4, and that this protection increases with the voltage applied for film formation. The relationship between the weight gain after oxidation in autoclave and the anodization voltage is of the exponential type: (σM/A) sub(AC) = Ce sup(-DV). The observed relationship between the applied voltage and the weight gain due to anodization is of the linear type: (σM/A) sub(AN) = aV. Concerning the influence of different electrolytes, it was observed a similar behaviour between them with respect to the thickness of the anodic oxide and the weight gain of zircaloy-4 after the autoclave test. (Author) [pt

The effect of second-phase particles on the corrosion and struture of Zircaloy-4

International Nuclear Information System (INIS)

Cortie, M.B.

1982-10-01

The effect of heat treatment and second-phase particles on the corrosion resistance and microstructure of Zircaloy-4 has been examined. In particular the effect of precipitates on the rate and mechanism of high-temperature, high-pressure water or steam corrosion is discussed. Measurements of corrosion rate are presented for specimens which have received various heat treatments. The heat treatments studied included a fast cool from the beta field, prolonged annealing at temperatures ranging from 500 degrees Celsius to 1 100 degrees Celsius as well as combinations of the above. The fabrication of a small quantity of Zircaloy-4 strip was undertaken and the methods used and observations made are recorded. The wide range of microstructures produced in Zircaloy-4 by the heat treatments and fabrication procedures utilized are described and discussed with optical or electron microscope photographs showing the important features. Qualitative and semi-quantitative chemical analyses of the second-phase particles were carried out by both the scanning electron microscope and Auger spectroscopy. Evidence for the existence of a tin-rich precipitate in Zircaloy-4 is presented and discussed

Fracture properties of hydrided Zircaloy-4 cladding in recrystallization and stress-relief anneal conditions

Energy Technology Data Exchange (ETDEWEB)

Hsu, Hsiao-Hung, E-mail: hhhsu@iner.gov.tw [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 325, Taiwan (China); Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan (China); Tsay, Leu-Wen [Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan (China)

2012-03-15

In this work, the stress-relieved (SRA) and recrystallized (RXA) Zircaloy-4 cladding specimens were hydrogen-charged to the target concentration of 300 wppm and then manufactured into X-specimens for fracture toughness test. The hydrogen embrittlement susceptibility of Zircaloy-4 cladding specimens in both SRA and RXA conditions were investigated. At the hydrogen concentration level of 300 wppm, J-integral values for RXA cladding were higher than those for SRA cladding at both 25 Degree-Sign C and 300 Degree-Sign C. The formation of brittle zirconium hydrides had a significant impact on the fracture toughness of Zircaloy-4 cladding in both SRA and RXA states, especially at 25 Degree-Sign C. Among all the tests, SRA cladding tested at 25 Degree-Sign C exhibited a great loss of the fracture toughness. The micrographic and fractographic observations further demonstrated that the fracture toughness of Zircaloy-4 cladding would be improved by the coarse grains in RXA cladding, but degraded by zirconium hydrides precipitated along the grain boundary.

Lifetime embrittlement of reactor core materials

International Nuclear Information System (INIS)

Kreyns, P.H..; Bourgeois, W.F.; Charpentier, P.L.; Kammenzind, B.F.; Franklin, D.G.; White, C.J.

1994-08-01

Over a core lifetime, the reactor materials Zircaloy-2, Zircaloy-4, and hafnium may become embrittled due to the absorption of corrosion- generated hydrogen and to neutron irradiation damage. Results are presented on the effects of fast fluence on the fracture toughness of wrought Zircaloy-2, Zircaloy-4, and hafnium; Zircaloy-4 to hafnium butt welds; and hydrogen precharged beta treated and weld metal Zircaloy-4 for fluences up to a maximum of approximately 150 x 10 24 n/M 2 (> 1 Mev). While Zircaloy-4 did not exhibit a decrement in K IC due to irradiation, hafnium and butt welds between hafnium and Zircaloy-4 are susceptible to embrittlement with irradiation. The embrittlement can be attributed to irradiation strengthening, which promotes cleavage fracture in hafnium and hafnium-Zircaloy welds, and, in part, to the lower chemical potential of hydrogen in Zircaloy-4 compared to hafnium, which causes hydrogen, over time, to drift from the hafnium end toward the Zircaloy-4 end and to precipitate at the interface between the weld and base-metal interface. Neutron radiation apparently affects the fracture toughness of Zircaloy-2, Zircaloy-4, and hafnium in different ways. Possible explanations for these differences are suggested. It was found that Zircaloy-4 is preferred over Zircaloy-2 in hafnium-to- Zircaloy butt-weld applications due to its absence of a radiation- induced reduction in K IC plus its lower hydrogen absorption characteristics compared with Zircaloy-2

Texture Of Zircaloy-4 Result Of Beta-Quenching, Cold Rolling And Recrystallization

International Nuclear Information System (INIS)

Futichah; Sulistioso

1998-01-01

Differences of crystallographic texture of zircaloy-4 plate depends on cold working and heat treatment.To determine the change of zircaloy-4 textures, the solid solution treatment process at beta phase which was followed by quenching on water was employed for this sample. The next step was cold rolling until deformation epsilon = 1.62. The specimens were recrystallized at 750 o C, for 2 hours. The result of beta-quench gave a spread and different orientations and the main orientation occurred at (0001)[1010] and (0001)[1120]. Result of cold rolling with epsilon = 1.39 and epsilon 1.62 is the deformation texture at the main orientation of (0001)[1010] with the angle of inclination was around 38 o. However, the result of Recrystallization process on 750 o C for 2 hours gave annealing textures with orientations of (0001)[1120]. It means that the recrystallization process of zircaloy-4 plate can not remove the deformation textures, but can change the crystallographic orientation

Modelling of zircaloy-4 corrosion in nitrogen and oxygen mixtures at high temperature

International Nuclear Information System (INIS)

Lasserre, M.; Peres, V.; Pijolat, M.; Coindreau, O.; Duriez, C.; Mardon, J.P.

2015-01-01

Previous studies of zircaloy-4 corrosion in air have shown accelerated corrosion in the 600-1000 Celsius degrees temperature range with Zr nitrides precipitating near the metal/oxide surface. The aim of this series of slides is to assess the influence of N 2 and O 2 partial pressures on the kinetic rate of growth of a new phase and to propose a kinetic modelling of zircaloy-4 corrosion

Electrolytic hydriding and hydride distribution in zircaloy-4

International Nuclear Information System (INIS)

Gomes, M.H.L.

1974-01-01

A study has been made of the electrolytic hydriding of zircaloy-4 in the range 20-80 0 C, for reaction times from 5 to 30 hours, and the effect of potential, pH and dissolved oxygen has been investigated. The hydriding reaction was more sensitive to time and temperature conditions than to the electrochemical variables. It has been shown that a controlled introduction of hydrides in zircaloy is feasible. Hydrides were found to be plate like shaped and distributed mainly along grain-boundaries. It has been shown that hydriding kinetics do not follow a simple law but may be described by a Johnson-Mehl empirical equation. On the basis of this equation an activation energy of 9.400 cal/mol has been determined, which is close to the activation energy for diffusion of hydrogen in the hydride. (author)

The anisotropic creep behaviour of zircaloy-4 fuel cladding at 1073 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Bowden, J.; Shewfelt, R.S.W.

1982-04-01

The anisotropy coefficients (F, G and H) of Hill's equation, suitably modified for creep deformation, have been determined for Zircaloy-4 fuel cladding from steady-state creep tests at an elevated temperature. Creep specimens were subjected to both uniaxial and biaxial loads (via internal pressure) at 1073 K and the strain measured concurrently in the axial and tangential directions. It has been found that Zircaloy-4 fuel cladding is almost, but not completely, isotropic at 1073 K; the values of F, G and H are 0.57, 0.48 and 0.45 respectively

High temperature interaction between Zircaloy-4 and stainless steel type 304

International Nuclear Information System (INIS)

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi

2001-03-01

The chemical interactions between Zircaloy-4 and stainless steel type 304 were investigated in the temperature range from 1273 to 1573 K to obtain the basic information on the melt progress in the fuel bundle during an LWR severe accident. Reaction layers were formed at the contact interface and grew as the temperature and the time increase. The Zircaloy was preferentially dissolved by the reaction. The SEM/EDX analyses showed that the main process of the reaction was diffusion of Fe, Cr and Ni into the Zircaloy which resulted in the formation of a Zr-rich eutectic through the tested temperature range. Reaction rates for decrease in the materials thickness were evaluated and the reaction generally obeyed a parabolic rate law. The reaction rate constant was determined at every examined temperature and Arrhenius type rate equations were estimated for the temperature range. (author)

Mechanisms of initial endplate failure in the human vertebral body.

Science.gov (United States)

Fields, Aaron J; Lee, Gideon L; Keaveny, Tony M

2010-12-01

Endplate failure occurs frequently in osteoporotic vertebral fractures and may be related to the development of high tensile strain. To determine whether the highest tensile strains in the vertebra occur in the endplates, and whether such high tensile strains are associated with the material behavior of the intervertebral disc, we used micro-CT-based finite element analysis to assess tissue-level strains in 22 elderly human vertebrae (81.5 ± 9.6 years) that were compressed through simulated intervertebral discs. In each vertebra, we compared the highest tensile and compressive strains across the different compartments: endplates, cortical shell, and trabecular bone. The influence of Poisson-type expansion of the disc on the results was determined by compressing the vertebrae a second time in which we suppressed the Poisson expansion. We found that the highest tensile strains occurred within the endplates whereas the highest compressive strains occurred within the trabecular bone. The ratio of strain to assumed tissue-level yield strain was the highest for the endplates, indicating that the endplates had the greatest risk of initial failure. Suppressing the Poisson expansion of the disc decreased the amount of highly tensile-strained tissue in the endplates by 79.4 ± 11.3%. These results indicate that the endplates are at the greatest risk of initial failure due to the development of high tensile strains, and that such high tensile strains are associated with the Poisson expansion of the disc. We conclude that initial failure of the vertebra is associated with high tensile strains in the endplates, which in turn are influenced by the material behavior of the disc. Copyright © 2010 Elsevier Ltd. All rights reserved.

Nondestructive hydrogen analysis of steam-oxidized Zircaloy-4 by wide-angle neutron scattering

Science.gov (United States)

Yan, Yong; Qian, Shuo; Garrison, Ben; Smith, Tyler; Kim, Peter

2018-04-01

A nondestructive neutron scattering method to precisely measure the hydrogen content in high-temperature steam-oxidized Zircaloy-4 cladding was developed. Zircaloy-4 cladding was used to produce hydrided specimens with hydrogen content up to ≈500 wppm. Following hydrogen charging, the hydrogen content of the hydrided specimens was measured using the vacuum hot extraction method, by which the samples with desired hydrogen concentrations were selected for the neutron study. The hydrided samples were then oxidized in steam up to ≈6.0 wt. % at 1100 °C. Optical microscopy shows that our hydriding procedure results in uniform distribution of circumferential hydrides across the wall thickness, and uniform oxide layers were formed on the sample surfaces by the steam oxidation. Small- and wide-angle neutron scattering were simultaneously performed to provide a quick (less than an hour per sample) measurement of the hydrogen content in various types of hydrided and oxidized Zircaloy-4. Our study demonstrates that the hydrogen in pre-oxidized Zircaloy-4 cladding can be measured very accurately by both small- and wide-angle neutron scattering. For steam-oxidized samples, the small-angle neutron scattering is contaminated with coherent scattering from additional structural features induced by the steam oxidation. However, the scattering intensity of the wide-angle neutron scattering increases proportionally with the hydrogen charged in the samples. The hydrogen content and wide-angle neutron scattering intensity are highly linearly correlated for the oxidized cladding samples examined in this work, and can be used to precisely determine the hydrogen content in steam-oxidized Zircaloy-4 samples. Hydrogen contents determined by neutron scattering of oxidation samples were also found to be consistent with the results of chemical analysis within acceptable margins for error.

Reaction behavior between B{sub 4}C, 304 grade of stainless steel and Zircaloy at 1473 K

Energy Technology Data Exchange (ETDEWEB)

Sasaki, Ryosuke [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan); Ueda, Shigeru, E-mail: tie@tagen.tohokku.ac.jp [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan); Kim, Sun-Joong [Dept. of Materials Science and Engineering, Chosun University, 309, Pilmun-daero, Dong-gu, Gwangju (Korea, Republic of); Gao, Xu; Kitamura, Shin-ya [Institute of Multidisciplinary Research Advanced Material, Tohoku University, 1-1 Katahira 2, Aoba-ku, Sendai (Japan)

2016-08-15

For a better understanding of the decommissioning of the Fukushima-daiichi nuclear power plant, the melting behavior of the control blade and the channel box should be clarified. In Fukushima nuclear reactor, the channel box was made of Zircaloy-4, and the control rode is made of B{sub 4}C together with stainless steel cladding and sheath. In the study, the interaction among B{sub 4}C, stainless steel (SUS), and Zircaloy-4 was investigated at 1473 K in either argon or air atmosphere. In argon, Zircaloy is melted by the diffusion of elements from SUS, and SUS was melted at 1473 K by the diffusion of C and B. In air, SUS reacted with B{sub 2}O{sub 3} and formed an oxides melt firstly. Then, the oxidized Zircaloy contacted with this melt and fused. Moreover, the progress of core melting process during severe accident under different atmospheres was firstly discussed. - Highlights: • The interaction among the system of B{sub 4}C, grade 304 stainless steel and Zircaloy-4 were studied at 1473 K in Ar and air. • In argon, Zircaloy is melted by the diffusion of elements from SUS, and SUS was melted by the diffusion of C and B. • In air, SUS reacted with B{sub 2}O{sub 3} and formed an oxides melt. Then, the oxidized Zircaloy contacted with this melt and fused.

Embrittlement by hydrogen in zircaloy-4

International Nuclear Information System (INIS)

Almendariz M, M.C.

1981-01-01

The brittleness study of zircaloy-4 (nuclear quality) by hydrogen in the lattice was carried out with the purpose to watch the alterations at mechanic properties and fracture appearance for different thermal treatments. We used a statistical experimental method to watch both alterations. Fracture toughness property was evaluated in a semiquantitative way, and this property was calculated by integral J method but at a modified version, this modification lies in the area calculation under the curve of load versus head displacement plot; we used Instron machine to evaluate it. Three points bending proof was carried out in accordance with the device that specify A.S.T.M. standards. The samples were treated with hydrogen by means of catodic charged method and subsequently mechanic proof was realized. We used statistical analysis to get information of experimental results, and the watched general behaviour was a great disminution of the fracture toughness (in relation to not treated hydrogen sample), always that the hydrogen is present in the lattice, likewise we did watch that hydrogen does not influence at fracture appearance change, further there is a threshold hydrogen concentration at wich it starts to brittle and prior not influence it. We did conclude of results analysis that the fracture toughness is reduced by hydrogen and threshold concentration is subject to thermal treatment. Experimental results can be considered as semiquantitatives, but they gave us an explicit idea of hydrogen effect in zircaloy-4. (author)

Effect of cyclic loading on the viscoplastic behaviour of Zircaloy 4 cladding tubes

International Nuclear Information System (INIS)

Bouffioux, P.; Gabriel, B.; Soniak, A.; Mardon, J.P.

1995-06-01

Most of the electricity being generated by nuclear energy load follow and remote control have become normal operating modes in the French PWR. In addition, EDF is developing a strategy of fuel sub-assembly burnup extension. Those operating conditions will lead to cyclic straining of the Zircaloy cladding tube which could induce damages. Therefore, EDF, CEA, and FRAMATOME has started a joint R and D cooperative program in order to investigate the mechanical behaviour of Zircaloy cladding tubes under cyclic loading. This paper is dealing with the effect of a pre-cyclic loading on the plasticity properties of Zircaloy 4 cladding tubes. Load controlled cyclic tests were carried out at 350 deg. C and 0.5 Hz in both axial and hoop directions. The Woehler curves were determined. Sequential tests combining pre-cyclic loading to 50 and 75 % fraction life with tension were then performed. It has ben noticed that the pre-cycling loading does not change the plastic flow curve of the Zircaloy 4 cladding tubes and therefore does not induce observable macroscopic damage. It has been concluded that a linear cumulative damage rule like ΣΔN(σ)/N r(σ) is very conservative. (author)

Plastic behaviour of Zircaloy-4 in the temperature range 77-1000 K

International Nuclear Information System (INIS)

Derep, J.L.; Ibrahim, S.; Rouby, D.; Fantozzi, G.; Gobin, P.

1979-01-01

Tensile tests were carried out on Zircaloy-4 over a temperature range 77-1000 K. So, we have determined the flow stress variations as a function of temperature and strain rate. Two thermally activated zones were observed between about 77 and 600 K, a plateau stress between 600 and 700 K and an other thermally activated zone above 700 K. The various mechanisms which can be responsible for the thermally activated and athermal zones are discussed in the light of experimental results. The mechanical behaviour of Zircaloy-4 appears similar to the zirconium-oxygen alloys one. (orig.) [de

Apparatus for spot welding sheathed thermocouples to the inside of small-diameter tubes at precise locations

International Nuclear Information System (INIS)

Baucum, W.E.; Dial, R.E.

1976-01-01

Equipment and procedures used to spot weld tantalum- or stainless-steel-sheathed thermocouples to the inside diameter of Zircaloy tubing to meet the requirements of the Multirod Burst Test (MRBT) Program at ORNL are described. Spot welding and oxide cleaning tools were fabricated to remove the oxide coating on the Zircaloy tubing at local areas and spot weld four thermocouples separated circumferentially by 90 0 at any axial distribution desired. It was found necessary to apply a nickel coating to stainless-steel-sheathed thermocouples to obtain acceptable welds. The material and shape of the inner electrode and resistance between inner and outer electrodes were found to be critical parameters in obtaining acceptable welds

Chemical interactions between as-received and pre-oxidized Zircaloy-4 and stainless steel at high temperatures

International Nuclear Information System (INIS)

Hofmann, P.

1994-05-01

The chemical reaction behavior between Zircaloy-4 and 1.4919 (AISI 316) stainless steel, which are used in absorber assemblies of Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR), has been studied in the temperature range 1000 - 1400 C. Zircaloy was used in the as-received, pre-oxidized and oxygen-containing condition. The maximum temperature was limited by the fast and complete liquefaction of the reaction couple as a result of eutectic chemical interactions. Liquefaction of the components occurs below their melting point. The effect of oxygen dissolved in Zircaloy plays an important role in the interaction; oxide layers on the Zircaloy surface delay the chemical interactions with stainless steel but cannot prevent them. Oxygen dissolved in Zircaloy reduces the reaction rates and shift the liquefaction temperature to slightly higher levels. The interaction experiments at the examined temperatures with or without pre-oxidized Zircaloy can be described by parabolic rate laws. The Arrhenius equations for the various conditions of interactions are given. (orig.) [de

Temperature measurement on Zircaloy-clad fuel pins during high temperature excursions

International Nuclear Information System (INIS)

Meservey, R.H.

1976-04-01

The development of a sheathed thermocouple suitable for attachment to zircaloy-clad fuel rods and for use during high temperature (2,800 0 F) excursions under loss-of-coolant accident conditions is described. Development, fabrication, and testing of the thermocouples is covered in detail. In addition, the development of a process for laser welding the thermocouples to fuel rods is discussed. The thermocouples and attachment welds have been tested for resistance to corrosion and nuclear radiation and have been subjected to fast thermal cycle, risetime, and blowdown accident tests

Apparatus for study of transient oxidation of Zircaloy-4 tubing

International Nuclear Information System (INIS)

Sagat, S.; Iglesias, F.C.; Newell, G.W.

1985-11-01

Complex transient oxidation tests on Zircaloy-4 tubing were performed to provide data for validation of the computer code FROM2. This code was developed to calculate oxygen distribution through oxidized Zircaloy tubing. The test temperature histories consisted of ramp, hold and cool cycles. The heating and cooling rates were in the range of 1 to 100 K/s and the maximum temperature was 1875 K. The apparatus developed to perform these experiments is described. In principle, Joule heating is used to heat the specimen and the temperature is controlled by a computer in conjunction with temperature and SCR power controllers. Using this combination, fast heating and cooling rates were achieved without sacrificing the accuracy of temperature control

Development of endplug welding technology for irradiation testing capsule

Energy Technology Data Exchange (ETDEWEB)

Lee, J. W.; Shin, Y. T.; Kim, S. S.; Kim, B. K.; Kang, Y. H. [KAERI, Taejon (Korea, Republic of)

2001-10-01

To evaluate the performance of newly developed nuclear fuel, it is necessary to irradiate the fuel at a research reactor and examine the irradiated fuel. For the irradiation test in a reasearch reactor, a fuel assembly which is generally called a capsule should be fabricated, considering the fuel irradiation plan and the characteristics of the reactor to be used. And also the fuel elements containing the developed fuel pellets should be made and assembled into a capsule. In this study, the welding method, welding equipment, welding conditions and parameters were developed to make fuel elements for the irradiation test at the HANARO research reactor. The TIG welding method using automatic orbital tube welding system was adopted and the welding joint design was developed for the fabrication of various kinds of irradiation fuel elements. And the optimal welding conditions and parameters were also established for the endplug welding of Zircaloy-4 cladding tube.

Zircaloy-4 and M5 high temperature oxidation and nitriding in air

Energy Technology Data Exchange (ETDEWEB)

Duriez, C. [Institut de Radioprotection et Surete Nucleaire, Direction de Prevention des Accidents Majeurs, Centre de Cadarache, 13115 St Paul Lez Durance (France)], E-mail: christian.duriez@irsn.fr; Dupont, T.; Schmet, B.; Enoch, F. [Universite Technologique de Troyes, BP 2060, 10010 Troyes (France)

2008-10-15

For the purpose of nuclear power plant severe accident analysis, degradation of Zircaloy-4 and M5 cladding tubes in air at high temperature was investigated by thermo-gravimetric analysis, in isothermal conditions, in a 600-1200 deg. C temperature range. Alloys were investigated either in a 'as received' bare state, or after steam pre-oxidation at 500 {sup o}C to simulate in-reactor corrosion. At the beginning of air exposure, the oxidation rate obeys a parabolic law, characteristic of solid-state diffusion limited regime. Parabolic rate constants compare, for Zircaloy-4 as well as for M5, with recently assessed correlations for high temperature Zircaloy-4 steam-oxidation. A thick layer of dense protective zirconia having a columnar structure forms during this diffusion-limited regime. Then, a kinetic transition (breakaway type) occurs, due to radial cracking along the columnar grain boundaries of this protective dense oxide scale. The breakaway is observed for a scale thickness that strongly increases with temperature. At the lowest temperatures, the M5 alloy appears to be breakaway-resistant, showing a delayed transition compared to Zircaloy-4. However, for both alloys, a pre-existing corrosion scale favours the transition, which occurs much earlier. The post transition kinetic regime is linear only for the lowest temperatures investigated. From 800 deg. C, a continuously accelerated regime is observed and is associated with formation of a strongly porous non-protective oxide. A mechanism of nitrogen-assisted oxide growth, involving formation and re-oxidation of ZrN particles, as well as nitrogen associated zirconia phase transformations, is proposed to be responsible for this accelerated degradation.

Four examples of non-ferrous metal electron beam welding

International Nuclear Information System (INIS)

Sommeria, J.

1989-01-01

The welding of superconducting cavity resonators made of niobium for particle accelerators is described. Then the welding of four plates in zircaloy 2 containing the fuel of the Orphee reactor is presented. The two other examples concern power transistor and motor support for planes. 9 figs [fr

In-situ neutron diffraction study of Zircaloy 4 subjected to biaxial tension

Energy Technology Data Exchange (ETDEWEB)

Gharghouri, M.A. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, ON (Canada); McDonald, D.; Xiao, L. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2014-07-01

Zircaloy-4 is widely used as fuel element cladding in nuclear reactors. Pellet-clad interaction (PCI) failure is a concern for many water reactor fuel designs. Extensive work on the mechanism of PCI failure has led to the conclusion that stress corrosion cracking (SCC) induced by iodine vapour in the temperature range relevant to fuel operation is the most probable cause of PCI failure in zirconium alloy fuel element cladding. In-situ neutron diffraction measurements performed on tubular Zircaloy-4 specimens simultaneously pulled in tension and pressurized internally will provide information on the effects of stress biaxiality on the distribution of stresses at the crystal level during loading. (author)

Effects of stress on the oxide layer thickness and post-oxidation creep strain of zircaloy-4

International Nuclear Information System (INIS)

Lim, Sang Ho; Yoon, Young Ku

1986-01-01

Effects of compressive stress generated in the oxide layer and its subsequent relief on oxidation rate and post-oxidation creep characteristics of zircaloy-4 were investigated by oxidation studies in steam with and without applied tensile stress and by creep testing at 700 deg C in high purity argon. The thickness of oxide layer increased with the magnitude of tensile stress applied during oxidation at 650 deg C in steam whereas similar phenomenon was not observed during oxidation at 800 deg C. Zircaloy-4 specimens oxidized at 600 deg C in steam without applied stress exhibited higher creep strain than that shown by unoxidized specimens when creep-tested in argon. Zircaloy-4 specimens oxidized at 600 deg C steam under the applied stress of 8.53MPa and oxidized at 800 deg C under the applied stress of 0 and 8.53MPa exhibited lower strain than that shown by unoxidized specimen. The above experimental results were accounted for on the basis of interactions among applied stress during oxidation, compressive stress generated in the oxide layer and elasticity of zircaloy-4 matrix. (Author)

Thermomechanical treatment of {beta}-treated Zircaloy-4 within the upper {alpha}-range; Traitements thermomecaniques dans le haut domaine {alpha} du zircaloy-4 trempe-{beta}

Energy Technology Data Exchange (ETDEWEB)

Chauvy, C

2004-09-15

Zircaloy-4 is a Zr base alloy mainly used for nuclear applications. This study deals with its behaviour during forming, with a special interest for physical mechanisms acting in the upper {alpha}-range. The {beta}-treated Zircaloy-4 is first described in terms of microstructure and texture. The {alpha} plates are organised in colonies with alternating crystallographic orientations. The Burgers relationships have also been checked. The mechanical behaviour shows two distinct domains (with or without hardening). This could be linked to interactions between the solutes (Sn, O) and mobile dislocations. The evolution of crystallographic texture is characterised by X-ray diffraction. At 550 C, twinning is shown to be the main deformation mechanism under specific experimental conditions. Globularization of the initial lamellar structure is identified as a continuous dynamic recrystallization process, more efficient at higher temperature. High strains are necessary to achieve this phenomenon but meta-dynamic recrystallization can also be used to obtain an equiaxed microstructure after limited strains. (author)

The steady-state creep of zircaloy-4 fuel cladding from 940 to 1873 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Bera, P.C.; Clendening, W.R.

1978-11-01

The steady-state creep rates of as-received Zircaloy-4 fuel cladding have been determined in the α-Zr phase (940 -6 and 10 -3 s -1 were determined under constant uniaxial load conditions. Assuming that creep rates can be described by a power law - Arrhenius equation, the creep rate for α-phase Zircaloy-4 is given by: epsilon sub(ss) = 2000σ sup(5.32) exp (-284 600/kT) s -1 and for the β-phase Zircaloy-4 is given by: epsilon sub(ss) = 8.1σ sup(3.79) exp (-142 300/kT) s -1 . For both the α-Zr and β-Zr phases, the activation energies for creep are in agreement with those for self-diffusion of zirconium and the rate-controlling mechanism is attributed to dislocation climb. Because of the scarcity of data, it is not possible to determine the rate equation unambiguously, nor to identify the mechanism for creep in the mixed α + β phase region. (author)

Electrochemical Studies on Important Elements for Zirconium Recovery Form Irradiated Zircaloy-4 Cladding

International Nuclear Information System (INIS)

Park, J.; Sohn, S.; Hwang, I.S.

2015-01-01

Since Zircaloy cladding accounts for about 16 wt. % of used nuclear fuel assembly, decontamination process is required to reduce the final waste volume from spent nuclear fuel. To develop Zircaloy-4 electrorefining process as an irradiated Zircaloy cladding decontamination process, electrochemical studies on Sn, Cr, Fe and Co which are major or important elements in the irradiated cladding were conducted based on cyclic voltammetry in LiCl-KCl at 500 deg. C. Cyclic voltammetry for Sn, Fe, Cr and Co elements that should be eliminated was conducted and revealed that redox reactions of these ions are much simpler than Zr and more reductive than Zr. The reliability of cyclic voltammetry was verified by comparing diffusion coefficients and formal reduction potentials of these ions obtained in this study to previous studies. (authors)

Adsorption and diffusion of hydrogen in Zircaloy-4

International Nuclear Information System (INIS)

Torres, E.; Desquines, J.; Baietto, M.C.; Coret, M.; Wehling, F.; Blat-Yrieix, M.; Ambard, A.

2015-01-01

Hydrogen in zirconium alloys is considered in many nuclear safety issues. Below 500 Celsius degrees, rather limited knowledge is available on the combined hydrogen adsorption at the sample surface and diffusion in the metal. A modeling of hydrogen gaseous charging has been established starting with a set of relevant laws and parameters derived from open literature. Simulating the hydrogen charging process requires simultaneous analysis of gaseous surface adsorption, hydrogen solid-solution diffusion and precipitation, when exceeding the material solubility limit. The modeling has been extended to reproduce the solid-gas exchange. Gaseous charging experiments have been performed at 420 C. degrees on Stress Relieved Annealed (SRA) Zircaloy-4 cladding samples to validate the model. The sample hydrogen content has been systematically measured after charging and compared to the calculated value thus providing a validation of the adsorption modeling. Complementary tests have been carried out on Recrystallized Annealed (RXA) Zircaloy-4 rods to characterize the combined diffusion and adsorption process. The hydrogen concentration distribution has been characterized using an inverse technique based on destructive analyses of the samples. This additional set of data was relevant for the validation of the hydrogen combined adsorption/diffusion modeling up to 420 C. degrees. (authors)

Chemical interactions between as-received and pre-oxidized Zircaloy-4 and Inconel-718 at high temperatures

International Nuclear Information System (INIS)

Hofmann, P.; Markiewicz, M.

1994-06-01

Isothermal reaction experiments were performed in the temperature range of 1000 - 1300 C in order to determine the chemical interactions between Zircaloy-4 fuel rod cladding and Inconel-718 spacer grids of Pressurized Water Reactors (PWR) under severe accident conditions. It was not possible to apply even higher temperatures since fast and complete liquefaction of the components occurred as a result of eutectic interactions during heatup. The liquid reaction products formed enhance and accelerate the degradation of the material couples and the fuel elements, respectively. Only small amounts of Inconel are necessary to liquefy large amounts of Zircaloy. Thin oxide layers on the Zircaloy surface delay the beginning of the chemical interactions with Inconel but cannot prevent them. In this work the reaction kinetics have been determined for the system: as-received and pre-oxidized Zircaloy-4/Inconel 718. The interactions can be described by parabolic rate laws; the Arrhenius equations for the various interactions are given. (orig.) [de

Texture, morphology and deformation mechanisms in β-transformed Zircaloy-4

International Nuclear Information System (INIS)

Ciurchea, D.; Furtuna, I.; Todica, M.; Roth, M.

1996-01-01

The morphology of the β(bcc) transformed Zircaloy-4 may be treated as a lenticular-twinned martensite. The texture is a consequence of the degeneration of the left angle 0001 right angle α , left angle 1010 right angle α and left angle 1011 right angle α directions into left angle 110 right angle β directions. The crystallographic mechanisms implied in the accommodation of the microscopic Bain strain are (1010) left angle 1120 right angle prism slip, (1012) left angle 101 1 right angle twinning and (1011) left angle 1012 right angle twinning. This degeneration explains the 'parallel plate' and 'basketweave' morphologies observed by microscopy and the texture of the β transformed tube. The macroscopic Bain strain was calculated and agrees with the dimensional measurements. The deformation mechanisms of β transformed Zircaloy-4 are identified from the new texture and from deformation experiments as twinning and interplatelet glide. The interplatelet glide induces a fragile character of fracture in the 'parallel plate' morphology. (orig.)

The corrosion of Zircaloy-4 fuel cladding in pressurized water reactors

International Nuclear Information System (INIS)

Van Swam, L.F.P.; Shann, S.H.

1991-01-01

This paper reports on the effects of thermo-mechanical processing of cladding on the corrosion of Zircaloy-4 in commercial PWRs that have been investigated. Visual observations and nondestructive measurements at poolside, augmented by observations in the hot cell, indicate that the initial black oxide transforms into a grey or tan later white oxide layer at a thickness of 10 to 15 μm independent of the thermal processing history of the tubing. At an oxide layer thickness of 60 to 80 μm, the oxide may spall depending somewhat on the particular oxide morphology formed and possibly on the frequency of power and temperature changes of the fuel rods. Because spalling of oxide lowers the metal-to-oxide interface temperature of fuel rods, it reduces the corrosion rate and is beneficial from that point of view. To determine the effect of thermo-mechanical processing on in-reactor corrosion of Zircaloy-4, oxide thickness measurements at poolside and in the hot cell have been analyzed with the MATPRO corrosion model. A calibrated corrosion parameter in this model provides a measure of the corrosion susceptibility of the Zircaloy-4 cladding. It was found necessary to modify the MATPRO equations with a burnup dependent term to obtain a near constant value of the corrosion parameter over a burnup range of approximately 10 to 45 MWd/kgU. Different calculational tests were performed to confirm that the modified model accurately predicts the corrosion behavior of fuel rods

Comparison of the air oxidation behaviors of Zircaloy-4 implanted with yttrium and cerium ions at 500 deg. C

International Nuclear Information System (INIS)

Chen, X.W.; Bai, X.D.; Xu, J.; Zhou, Q.G.; Chen, B.S.

2002-01-01

As a valuable process for surface modification of materials, ion implantation is eminent to improve mechanical properties, electrochemical corrosion resistance and oxidation behaviors of varieties of materials. To investigate and compare the oxidation behaviors of Zircaloy-4, implantation of yttrium ion and cerium ion were respectively employed by using an MEVVA source at the energy of 40 keV with a dose ranging from 1x10 16 to 1x10 17 ions/cm 2 . Subsequently, weight gain curves of the different specimens including as-received Zircaloy-4 and Zircaloy-4 specimens implanted with the different ions were measured after oxidation in air at 500 deg. C for 100 min. It was obviously found that a significant improvement was achieved in the oxidation behaviors of implanted Zircaloy-4 compared with that of the as-received Zircaloy-4, and the oxidation behavior of cerium-implanted Zircaloy-4 was somewhat better than that of yttrium-implanted specimen. To obtain the valence and the composition of the oxides in the scale, X-ray photoemission spectroscopy was used in the present study. Glancing angle X-ray diffraction, employed to analyze the phase transformation in the oxide films, showed that the addition of yttrium transformed the phase from monoclinic zirconia to tetragonal zirconia, yet the addition of cerium transformed the phase from monoclinic zirconia to hexagonal zirconia. In the end, the mechanism of the improvement of the oxidation behavior was discussed

Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4; Filage et etirage du zircaloy 2. Realisation des tubes de force pour EL-4

Energy Technology Data Exchange (ETDEWEB)

Thevenet, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires; Buffet, J [Cefilac (France)

1964-07-01

The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the {phi} = 340 ingot into {phi} = 220 billets, cutting into lengths and hot drilling at {phi} = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes ({phi} =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [French] Les auteurs donnent un resume succint des proprietes physiques mecaniques et chimiques du zircaloy 2 en ce qui concerne la transformation de cet alliage. Filage: Apres quelques generalites sur le filage et le cofilage, dont une comparaison entre les resistances a la deformation des metaux de gainage et du zircaloy 2, on etudie successivement: - les

High temperature interaction between UO2 and Zircaloy-4/silver mixture

International Nuclear Information System (INIS)

Uetsuka, Hiroshi; Nagase, Fumihisa; Otomo, Takashi

1995-12-01

The reaction between UO 2 and Zircaloy is a main material interaction in the reactor core during a severe accident of LWR. With a view of examining the influence of the core materials having low melting temperatures on the reaction, the effect of silver that is main component of PWR control rod alloy was investigated in the temperature range from 1373 to 1703K. Zircaloy was completely liquefied by the same weight of liquid silver at tested temperatures. The reaction between UO 2 and (Zircaloy+silver) mixture roughly obeyed a parabolic rate law. The determined reaction rate below about 1600K was much lower than that obtained by Hofmann et al. for the reaction between UO 2 and Zircaloy. However, it sharply increased with temperature and became comparable with the rate of UO 2 /Zircaloy reaction at about 1700K. Metallurgical examination including EPMA analysis revealed that Zr(O) layer formed at the reaction interface only for the tests below about 1600K correlated with the discontinuity of the temperature dependence of reaction rate. (author)

Resistance welding equipment manufacturing capability for exports

Energy Technology Data Exchange (ETDEWEB)

Sastry, V.S.; Raju, Y.S.; Somani, A.K.; Setty, D.S.; Rameswara Raw, A.; Hermantha Rao, G.V.S.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. of Atomic Energy, Hyderbad (India)

2010-07-01

Indian Pressurised Heavy Water Reactor (PHWR) fuel bundle is fully welded and is unique in its design. Appendage welding, end closure welding, and end plate welding is carried out using resistance welding technique. Out of many joining processes available, resistance-welding process is reliable, environment friendly and best suitable for mass production applications. Nuclear Fuel Complex (NFC), an industrial unit is established in Hyderabad, under the aegis of the Dept of Atomic Energy to manufacture fuel for Pressurised Heavy Water Reactors. From inception, NFC has given importance for self-reliance and indigenization with respect to manufacturing process and equipment. Sintering furnaces, centreless grinders, appendage-welding machines, end-closure welding equipment and end-plate welding equipments, which were initially imported, are either indigenized or designed and manufactured in house. NFC has designed, manufactured a new appendage-welding machine for manufacturing 37 element fuel bundles. Recently NFC has bagged an order from IAEA through international bidding for design, manufacture, supply, erection and commissioning of end-closure welding equipment. The paper gives in detail the salient features of these welding equipment. (author)

Instrumented impact properties of zircaloy-oxygen and zircaloy-hydrogen alloys

Energy Technology Data Exchange (ETDEWEB)

Garde, A.M.; Kassner, T.F.

1980-04-01

Instrumented-impact tests were performed on subsize Charpy speciments of Zircaloy-2 and -4 with up to approx. 1.3 wt % oxygen and approx. 2500 wt ppM hydrogen at temperatures between 373 and 823/sup 0/K. Self-consistent criteria for the ductile-to-brittle transition, based upon a total absorbed energy of approx. 1.3 x 10/sup 4/ J/m/sup 2/, a dynamic fracture toughness of approx. 10 MPa.m/sup 1/2/, and a ductility index of approx. 0, were established relative to the temperature and oxygen concentration of the transformed BETA-phase material. The effect of hydrogen concentration and hydride morphology, produced by cooling Zircaloy-2 specimens through the temperature range of the BETA ..-->.. ..cap alpha..' = hydride phase transformation at approx. 0.3 and 3 K/s, on the impact properties was determined at temperatures between 373 and 673 K. On an atom fraction basis, oxygen has a greater effect than hydrogen on the impact properties of Zircaloy at temperatures between approx. 400 and 600 K. 34 figures.

Effects of deposited pyrolytic carbon on some mechanical properties of zircaloy-4 tubes. Vol. 3

Energy Technology Data Exchange (ETDEWEB)

Shrkawy, S W; Abdel-razek, I D; El-Sayed, H A [Metallurgy Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

1996-03-01

Zircaloy cladding tubes are not compatible with the uranium fuel pellets as they suffer from failure due to pelletclad interaction (PCI). A carbon coating, as used in the canadian CANLUB fuel elements, is thought to improve the cladding performance with respect to the PCI problem. In this paper pyrolytic carbon coating was deposited on zircaloy-4 cladding tubes by the thermal cracking of commercial butant gas at the temperature range 250-450 degree C. In order to evaluate the effect of gaseous species on the mechanical properties of the tubes tensile and microhardness testing measurements were performed on samples prepared from the coated tubes. The fractured surface of the tensile zircaloy tubes and the deposited carbon coating, both, were examined by the SEM. The results of the tensile tests of zircaloy-4 tubes indicated that the coating process has insignificant effect on the ultimate strength of the tubes tested. The values of Vickers hardness numbers were not significantly changed across the tubes thickness. The microstructure of deposited carbon, due to the cracking process, was granular in all the temperature range (250-450 degree C) studied. 9 figs., 1 tab.

A unified model to describe the anisotropic viscoplastic behavior of Zircaloy-4 cladding tubes

International Nuclear Information System (INIS)

Delobelle, P.; Robinet, P.; Bouffioux, P.; Geyer, P.; Pichon, I. Le

1996-01-01

This paper presents the constitutive equations of a unified viscoplastic model and its validation with experimental data. The mechanical tests were carried out in a temperature range of 20 to 400 C on both cold-worked stress-relieved and fully annealed Zircaloy-4 tubes. Although their geometry (14.3 by 1.2 mm) is different, the crystallographic texture was close to that expected in the cladding tubes. To characterize the anisotropy, mechanical tests were performed under both monotonic and cyclic uni- and bi-directional loadings, i.e., tension-compression, tension-torsion, and tension-internal pressure tests. The results obtained at ambient temperatures and the independence of the ratio R p = var-epsilon θθ p /var-epsilon zz p , with respect to temperature would seem to indicate that the set of anisotropy coefficients does not depend on temperature. Zircaloy-4 material also has a slight supplementary hardening during out-of-phase cyclic loading. The authors propose to extend the formulation of a unified viscoplastic model, developed and identified elsewhere for other initially isotropic materials, to the case of Zircaloy-4. Generally speaking, anisotropy is introduced through fourth order tensors affecting the flow directions, the linear kinematical hardening components, as well as the dynamic and static recoveries of the forementioned hardening variables. The ability of the model to describe all the mechanical properties of the material is shown. The application of the model to simulate mechanical tests (tension, creep, and relaxation) performed on true CWSR Zircaloy-4 cladding tubes with low tin content is also presented

Investigation of the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4

International Nuclear Information System (INIS)

Soares, M.I.

1981-12-01

To investigate the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4 tubes, deformation tests under pressure of samples hydrided in autoclave and of samples containing iodine were carried out, in order to simulate the fission product. The same tests were carried out in samples without hydride and iodine contents that were used as reference samples in the temperature range of 650 0 C-950 0 C. The hydrided samples and the samples containing iodine tested at 650 0 C and 750 0 C showed a higher ductility than the samples of reference. The hydrided samples tested at 850 0 C and 950 0 C showed a higher embritlement than the samples of reference and than the samples containing iodine tested at the same temperatures. A mechanical test has been developed to investigate the effect of hydride and iodine on the mechanical behaviour of the zircaloy-4 tubes. The mechanical test were carried out at room temperature. At room temperature the hydrition decreased the ductility of zircaloy-4. At room temperature the sample containing iodine showed a higher ductility than the sample without iodine. The combined action of hydrogen and iodine at room temperature enhanced the embrittlment of the samples zircaloy-4. (Author) [pt

The characteristics of surface oxidation and corrosion resistance of nitrogen implanted zircaloy-4

International Nuclear Information System (INIS)

Tang, G.; Choi, B.H.; Kim, W.; Jung, K.S.; Kwon, H.S.; Lee, S.J.; Lee, J.H.; Song, T.Y.; Shon, D.H.; Han, J.G.

1997-01-01

This work is concerned with the development and application of ion implantation techniques for improving the corrosion resistance of zircaloy-4. The corrosion resistance in nitrogen implanted zircaloy-4 under a 120 keV nitrogen ion beam at an ion dose of 3 x 10 17 cm -2 depends on the implantation temperature. The characteristics of surface oxidation and corrosion resistance were analyzed with the change of implantation temperature. It is shown that as implantation temperature rises from 100 to 724 C, the colour of specimen surface changes from its original colour to light yellow at 100 C, golden at 175 C, pink at 300 C, blue at 440 C and dark blue at 550 C. As the implantation temperature goes above 640 C, the colour of surface changes to light black, and the surface becomes a little rough. The corrosion resistance of zircaloy-4 implanted with nitrogen is sensitive to the implantation temperature. The pitting potential of specimens increases from 176 to 900 mV (SCE) as the implantation temperature increases from 100 to 300 C, and decreases from 900 to 90 mV(SCE) as the implantation temperature increases from 300 to 640 C. The microstructure, the distribution of oxygen, nitrogen and carbon elements, the oxide grain size and the feature of the precipitation in the implanted surface were investigated by optical microscope, TEM, EDS, XRD and AES. The experimental results reveal that the ZrO 2 is distributed mainly on the outer surface. The ZrN is distributed under the ZrO 2 layer. The characteristics of the distribution of ZrO 2 and ZrN in the nitrogen-implanted zircaloy-4 is influenced by the implantation temperature of the sample, and in turn the corrosion resistance is influenced. (orig.)

Irradiation creep and growth of zircaloy-4 tubes

International Nuclear Information System (INIS)

Lansiart, S.; Darchis, L.; Pelchat, J.

1990-01-01

The influence of temperature and fast neutron flux on irradiation creep and growth of stress relieved zircaloy-4 pressurized tubes has been derived from experimental irradiations in NaK, performed up to 2.5 10 25 n.m -2 in the temperature range [280, 350] 0 C. A significant influence of temperature on axial growth has been observed: at 280 0 C the elongation can no longer be expressed as a linear function of fluence as for the 350 0 C irradiation temperature; diametral growth, on the other hand, always appears negligible. Irradiation creep obviously depends on temperature too; the diametral strain (including thermal part) has been modelled as a sum of primary and secondary terms, the former being independent of fluence. For the tubing considered it is observed that the ranking of the different batches, with respect to diametral creep resistance, is the same before and under irradiation. Concerning axial creep strain the stress relieved material behaves as does an isotropic tube. This is not the case of recrystallized zircaloy-4 F, which shows a non negligible axial deformation, related to the diametral creep one, even though this diametral irradiation creep strain is strongly reduced comparatively to that of the stress relieved material. The comparison of the two materials growth rates is more complex since their dependence on temperature and flux differs

Experiments on ballooning in pressurized and transiently heated Zircaloy-4 tubes

International Nuclear Information System (INIS)

Markiewicz, M.E.; Erbacher, F.J.

1988-02-01

Single-rod burst tests were performed with Atucha I Zircaloy-4 cladding tubes in the REBEKA burst equipment of KfK. The objective was to investigate the ballooning and burst behavior of argentine cladding tubes obtained from NRG, Germany and CONVAR, Argentina. The burst data were compared with those of cladding tubes used in german PWR's. It was found that the burst data e.g. burst temperature, circumferential burst strain and its response to azimuthal temperature differences are identical for the Argentine and German tubing quality. The burst data are in good agreement with those of German PWR-Zircaloy tubes. Thus, the fuel rod behavior codes developed for German PWR's can also be used for the Argentine reactor Atucha I. (orig.) [de

MICROSTRUCTURE AND MECHANICAL STRENGTH OF SURFACE ODS TREATED ZIRCALOY-4 SHEET USING LASER BEAM SCANNING

Directory of Open Access Journals (Sweden)

HYUN-GIL KIM

2014-08-01

Full Text Available The surface modification of engineering materials by laser beam scanning (LBS allows the improvement of properties in terms of reduced wear, increased corrosion resistance, and better strength. In this study, the laser beam scan method was applied to produce an oxide dispersion strengthened (ODS structure on a zirconium metal surface. A recrystallized Zircaloy-4 alloy sheet with a thickness of 2 mm, and Y2O3 particles of 10 μm were selected for ODS treatment using LBS. Through the LBS method, the Y2O3 particles were dispersed in the Zircaloy-4 sheet surface at a thickness of 0.4 mm, which was about 20% when compared to the initial sheet thickness. The mean size of the dispersive particles was 20 nm, and the yield strength of the ODS treated plate at 500°C was increased more than 65 % when compared to the initial state. This strength increase was caused by dispersive Y2O3 particles in the matrix and the martensite transformation of Zircaloy-4 matrix by the LBS.

Nucleation and growth of intermetallic precipitates in Zircaloy-2 and zircaloy-4 and correlation to nodular corrosion behavior

International Nuclear Information System (INIS)

Maussner, G.; Steinberg, E.; Tenckhoff, E.

1987-01-01

One of the fundamental aspects in the history of the development of zirconium alloys for nuclear applications is the corrosion behavior under in-pile conditions. In boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) the zirconium alloys Zircaloy-2 and Zircaloy-4 are the most commonly used materials, permitting attainment of a very high level of integrity and reliability. Nevertheless, efforts are required to optimize these well-established alloys with regard to their resistance to nodular corrosion, where improvements will give long-term advantages in fuel integrity and fuel economy. Phenomenological studies allow correlation of the nodular corrosion behavior with the morphological appearance of precipitated intermetallic phases in the microstructures of Zry-2 and Zry-4. To understand the fundamental processes of precipitation, particle nucleation-and-growth studies were made with Zry-2 and Zry-4 in different fabrication dimensions and with variations in β-quenching rates followed by isothermal and isochronical heat treatments. The microstructural characteristics of the precipitates were investigated by optical and transmission-electron microscopy. The macroscopic behavior was studied by electrical-resistivity measurements and hardness measurements. The nodular-corrosion susceptibility was determined by weight-gain and nodule distribution measurements after a 500 0 C laboratory-autoclave test

Young's modulus of crystal bar zirconium and zirconium alloys (zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium) to 1000 K

International Nuclear Information System (INIS)

Rosinger, H.E.; Ritchie, I.G.; Shillinglaw, A.J.

1975-09-01

This report contains experimentally determined data on the dynamic elastic moduli of zircaloy-2, zircaloy-4, zirconium-2.5wt% niobium and Marz grade crystal bar zirconium. Data on both the dynamic Young's moduli and shear moduli of the alloys have been measured at room temperature and Young's modulus as a function of temperature has been determined over the temperature range 300 K to 1000 K. In every case, Young's modulus decreases linearly with increasing temperature and is expressed by an empirical equation fitted to the data. Differences in Young's modulus values determined from specimens with longitudinal axes parallel and perpendicular to the rolling direction are small, as are the differences between Young's moduli determined from strip, bar stock and fuel sheathing. (author)

The steam pressure effect on high temperature corrosion of zircaloy-4

International Nuclear Information System (INIS)

Kim, K. P.; Park, G. H.

1998-01-01

To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The degree of high temperature oxidation of zircaloy-4 was measured at three different conditions, high pressure steam, high pressure Ar gas with small amount of steam, and 1 atm steam. All the measurements were done at 750 deg C. The oxide thickness is much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. No effect was observed in the case of high pressure Ar containing small amount of steam. Many cracks exist on the surface of specimens oxidized at high pressure steam, which come from the enhanced tetragonal to monoclinic phase transformation due to high pressure steam. The enhanced oxidation seems to oxide cracking

Diffusionless bonding of aluminum to Zircaloy-2

International Nuclear Information System (INIS)

Watson, R.D.

1965-04-01

Aluminum can be bonded to zirconium without difficulty even when a thin layer of oxide is present on the surface of the zirconium . No detectable diffusion takes place during the bonding process. The bond layer can be stretched as much. as 8% without affecting the bond. The bond can be heated for 1000 hours at 260 o C (500 o F), and can be water quenched from 260 o C (500 o F) without any noticeable change in the bond strength. An extrusion technique has been devised for making transition sections of aluminum bonded to zirconium which can then be used to join these metals by conventional welding. Welding can be done close to the bond zone without seriously affecting the integrity of the bond. This method of bonding aluminum to Zircaloy-2 is covered by Canadian patent 702,438 January 26, 1965. (author)

Effect of deformation on crystallite characteristic and yield stress of zircaloy-4

International Nuclear Information System (INIS)

Sugondo; Futichah

2007-01-01

The effect of deformation (rolling) on micro strain, crystallite size, crystallite density, and yield strength of Zircaloy-4 was characterized by x-ray diffraction. The goal of this investigation is to characterize the cladding materials of PWR and the target is to have data on the crystallography of Zircaloy-4. The as-received material with the composition 1.3% Sn, 0.22% Fe, 0.1% Cr, and Zr balanced was cut 10 mm × 100 mm in size using diamond blade. The samples were cleaned and heated at 1100 °C for 2 hours and then quenched in cold water. Then the sample were cleaned and heated at 750 °C for 2 hours. Afterward the samples were cold rolled with 40%, 75%, and 80% reduction in thickness. After the preparation was completed, the crystals of the samples were characterized using X-ray diffraction. The processes being analysed were quenching followed by annealing, plastic deformation of annealing and reduction from 40% to 80%, and the constancy of the c/a ratio. From the analyses, three conclusions were obtained. Firstly, the annealing process at 750 °C of Zircaloy-4 from the quenched samples resulted in the recrystallization and the grain growth which was proven by the increase of micro strain from 25.05% to 32.83%, the increase of crystallite size from 10.1015 Å to 287.4798 Å, the decrease of dislocation density from 2.94E+16 m/m3 to 3.63E+13 m/m3, and the decrease of yield strength from 1125.52 MPa to 304.44 MPa. Secondly, the process of reduction of Zircaloy-4 from the annealed samples reduced to 80% resulted in the plastic deformation and crystallite which was shown by the decrease of micro strain from 32.83% to 3.15%, the decrease of crystallite size from 287.4798 Å to 10.9764 Å, the increase of dislocation density from 3.63E+13 m/m3 to 2.49E+16 m/m3, and the increase of yield strength from 304.44 MPa to 1057.69 MPa. Thirdly, the process of plastic deformation of Zircaloy-4 was limited by the constancy of the c/a ratio which was verified by the decrease

Fatigue testing on samples from Zircaloy-4 tubes type SEU-43

International Nuclear Information System (INIS)

Olaru, V.; Ionescu, V.; Nitu, A.; Ionescu, D.; Voicu, F.

2016-01-01

The paper presents the testing of samples worked from Zicaloy-4 tubes (as-received.. metallurgical state), utilized in the composition of the CANDU SEU-43 fuel bundle. These tests are intended to simulate their behaviour in a power cycling process inside the reactor. The testing process is of low cycle fatigue type, done outside of the reactor, on ''C-ring'' samples, cut along the transversal direction. These samples are tested at 1%, 2% and 3% amplitude deformation, at room temperature. The calibration curves for both types of tube (small and big diameter) are determined by using the finite element analyses with the ANSYS computer code. The cycling test results are in the form of a fatigue life curve (N-e) for zircaloy-4 used in the SEU-43 fuel bundle. The curve is determined by the experimental dependency between the number of cycles to fracture and the deformation amplitude. The low cycle fatigue mechanical tests done at room temperature together with electronic microscopy analyses have reflected the characteristic behaviour of the zircaloy-4 metal in the given environment conditions. (authors)

Pool boiling CHF enhancement by micro/nanoscale modification of zircaloy-4 surface

International Nuclear Information System (INIS)

Ahn, Ho Seon; Lee, Chan; Kim, Hyungdae; Jo, HangJin; Kang, SoonHo; Kim, Joonwon; Shin, Jeongseob; Kim, Moo Hwan

2010-01-01

Consideration of the critical heat flux (CHF) requires difficult compromises between economy and safety in many types of thermal systems, including nuclear power plants. Much research has been directed towards enhancing the CHF, and many recent studies have revealed that the significant CHF enhancement in nanofluids is due to surface deposition of nanoparticles. The surface deposition of nanoparticles influenced various surface characteristics. This fact indicated that the surface wettability is a key parameter for CHF enhancement and so is the surface morphology. In this study, surface wettability of zircaloy-4 used as cladding material of fuel rods in nuclear power plants was modified using surface treatment technique (i.e. anodization). Pool boiling experiments of distilled water on the prepared surfaces was conducted at atmospheric and saturated conditions to examine effects of the surface modification on CHF. The experimental results showed that CHF of zircaloy-4 can be significantly enhanced by the improvement in surface wettability using the surface modification, but only the wettability effect cannot explain the CHF increase on the treated zircaloy-4 surfaces completely. It was found that below a critical value of contact angle (10 o ), micro/nanostructures created by the surface treatment increased spreadability of liquid on the surface, which could lead to further increase in CHF even beyond the prediction caused only by the wettability improvement. These micro/nanostructures with multiscale on heated surface induced more significant CHF enhancement than it based on the wettability effect, due to liquid spreadability.

Effects of operating conditions on molten-salt electrorefining for zirconium recovery from irradiated Zircaloy-4 cladding of pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Park, Jaeyeong, E-mail: d486916@snu.ac.kr [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Choi, Sungyeol [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Sohn, Sungjune [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of); Kim, Kwang-Rag [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Hwang, Il Soon [Department of Nuclear Engineering, Seoul National University, 1 Gwanak-ro, Gwanak-gu, Seoul 151-742 (Korea, Republic of)

2014-08-15

Highlights: • Computational simulation on electrorefining of irradiated Zircaloy-4 cladding. • Composition of irradiated Zircaloy-4 cladding of pressurized water reactor. • Redox behavior of elements in irradiated Zircaloy cladding during electrorefining. • Effect of electrorefining operating conditions on decontamination factor. - Abstract: To reduce the final waste volume from used nuclear fuel assembly, it is significant to decontaminate irradiated cladding. Electrorefining in high temperature molten salt could be one of volume decontamination processes for the cladding. This study examines the effect of operating conditions on decontamination factor in electrorefining of irradiated Zircaloy-4 cladding of pressurized water reactor. One-dimensional time-dependent electrochemical reaction code, REFIN, was utilized for simulating irradiated cladding electrorefining. Composition of irradiated Zircaloy was estimated based on ORIGEN-2 and other literatures. Co and U were considered in electrorefining simulation with major elements of Zircaloy-4 to represent activation products and actinides penetrating into the cladding respectively. Total 240 cases of electrorefining are simulated including 8 diffusion boundary layer thicknesses, 10 concentrations of contaminated molten salt and 3 termination conditions. Decontamination factors for each case were evaluated and it is revealed that the radioactivity of Co-60 in recovered zirconium on cathode could decrease below the clearance level when initial concentration of chlorides except ZrCl{sub 4} is lower than 1 × 10{sup −11} weight fraction if electrorefining is finished before anode potential reaches −1.8 V (vs. Cl{sub 2}/Cl{sup −})

The influence of hydride on fracture toughness of recrystallized Zircaloy-4 cladding

Energy Technology Data Exchange (ETDEWEB)

Hsu, Hsiao-Hung, E-mail: 175877@mail.csc.com.tw [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 32546, Taiwan, ROC (China); China Steel Corporation, Hsiao Kang District, Kaohsiung 81233, Taiwan, ROC (China); Chiang, Ming-Feng [China Steel Corporation, Hsiao Kang District, Kaohsiung 81233, Taiwan, ROC (China); Chen, Yen-Chen [Institute of Nuclear Energy Research (INER), Lungtan Township, Taoyuan County 32546, Taiwan, ROC (China)

2014-04-01

In this work, RXA cladding tubes were hydrogen-charged to target hydrogen content levels between 150 and 800 wppm (part per million by weight). The strings of zirconium hydrides observed in the cross sections are mostly oriented in the circumferential direction. The fracture toughness of hydrided RXA Zircaloy-4 cladding was measured to evaluate its hydride embrittlement susceptibility. With increasing hydrogen content, the fracture toughness of hydrided RXA cladding decreases at both 25 °C and 300 °C. Moreover, highly localized hydrides (forming a hydride rim) aggravate the degradation of the fracture properties of RXA Zircaloy-4 cladding at both 25 °C and 300 °C. Brittle features in the form of quasi-cleavages and secondary cracks were observed on the fracture surface of the hydride rim, even for RXA cladding tested at 300 °C.

Effects of Nitrogen Implantation on the Resistance to Localized Corrosion of Zircaloy-4 in a Chloride Solution

International Nuclear Information System (INIS)

Lee, Sung Joon; Kwon, Hyuk Sang; Kim, Wan; Choi, Byung Ho

1996-01-01

The influences of ion dose and substrate temperature on the resistance to localized corrosion of nitrogen-implanted Zircaloy-4 are examined in terms of potentiodynamic anodic polarization tests in deaerated 4M NaCl solution at 80 .deg. C. Nitrogen implantations into the Zircaloy-4 were performed under conditions of varying the ion dose from 3 x 10 17 to 1.2 x 10 18 ions/cm 2 and of maintaining the substrate temperatures respectively at 100, 200, and 300 .deg. C by controlling the current density of ion beam. The resistance to localized corrosion of Zircaloy-4 was significantly increased with increasing the ion dose when implanted at substrate temperatures above 200 .deg. C. However, it was not almost improved by implantation at 100 .deg. C. Specifically, the pitting potential increased from 350mV (vs. SCE) for the unimplanted to values of 900 to about 1400mV (vs. SCE) for the implanted alloy depending on the nitrogen dose. This significant improvement in the resistance to localized corrosion of the implanted Zircaloy-4 was found to be associate with the formation of compound layers of ZrO 2 + ZrN during the implantation. The galvanostatic anodization tests on the nitrogen-implanted Zircaloy-4 in 1M H 2 SO 4 at 20 .deg. C demonstrated that an increase in the ion dose and also in the substrate temperature increased the thickness of the compound layer of ZrO 2 + ZrN, and hence increased the pitting potential of the alloy. The low resistance to localized and general corrosion of the alloy implanted at 100 .deg. C was attributed to the increase in surface defect density and also to thinner implanted layer compared with those formed at higher temperatures

Cervical endplate and facet arthrosis: an anatomic study of cadaveric specimens.

Science.gov (United States)

Master, Daniel L; Toy, Jason O; Eubanks, Jason D; Ahn, Nicholas U

2012-10-01

An anatomic, epidemiologic study of cervical endplate and facet arthrosis in cadaveric spines. To determine the prevalence of cervical endplate and facet arthrosis and the relationship between these 2 entities in a large population sample. Cervical endplate and facet arthrosis are common radiographic findings, which have both been linked to pain. However, the prevalence and relationship between cervical endplate and facet arthrosis has not been clearly defined. The cervical vertebrae from 234 cadaveric spines were examined by a single investigator for evidence of endplate and facet arthrosis. Arthrosis at each endplate and facet was graded on a continuum from 0 to IV. Race, age at death, and sex of each specimen were also recorded. Stepwise multiple linear regression was used to analyze any association between race, age, sex, endplate arthrosis, and facet arthrosis. Factors with P values arthrosis severity scores among patients within the same decades of life. Concurrent cervical endplate and facet arthrosis was present in 77% of the study population. Stepwise multiple linear regression revealed significant (Parthrosis and between age and facet arthrosis. Race and sex did not correlate with facet arthrosis. In addition, patients in age groups 30 to 39, 40 to 49, 50 to 59, 60 to 69, 70 to 79, and 80 to 89 demonstrated more severe (Parthrosis in comparison with facet arthrosis. Concurrent cervical endplate and facet arthrosis is a common condition. Cervical endplate arthrosis and advancing age are associated with cervical facet arthrosis independent of race and sex. Cervical endplate arthrosis precedes facet arthrosis.

Brazing characteristics of a Zr-Ti-Cu-Fe eutectic alloy filler metal for Zircaloy-4

Science.gov (United States)

Lee, Jung G.; Lim, C. H.; Kim, K. H.; Park, S. S.; Lee, M. K.; Rhee, C. K.

2013-10-01

A Zr-Ti-Cu-Fe quaternary eutectic alloy was employed as a new Be-free brazing filler metal for Zircaloy-4 to supersede physically vapor-deposited Be coatings used conventionally with several disadvantages. The quaternary eutectic composition of Zr58Ti16Cu10Fe16 (at.%) showing a low melting temperature range from 832 °C to 853 °C was designed by a partial substitution of Zr with Ti based on a Zr-Cu-Fe ternary eutectic system. By applying an alloy ribbon with the determined composition, a highly reliable joint was obtained with a homogeneous formation of predominantly grown α-Zr phases owing to a complete isothermal solidification, exhibiting strength higher than that of Zircaloy-4. The homogenization of the joint was rate-controlled by the diffusion of the filler elements (Ti, Cu, and Fe) into the Zircaloy-4 base metal, and the detrimental segregation of the Zr2Fe phase in the central zone was completely eliminated by an isothermal holding at a brazing temperature of 920 °C for 10 min.

Stress corrosion cracking of Zircaloy-4 in non-aqueous iodine solutions

International Nuclear Information System (INIS)

Gomez Sanchez, Andrea V.

2006-01-01

In the present work the susceptibility to intergranular attack and stress corrosion cracking of Zircaloy-4 in different iodine alcoholic solutions was studied. The influence of different variables such as the molecular weight of the alcohols, the water content of the solutions, the alcohol type (primary, secondary or tertiary) and the temperature was evaluated. To determine the susceptibility to stress corrosion cracking the slow strain rate technique was used. Specimens of Zircaloy-4 were also exposed between 0.5 and 300 hours to the solutions without applied stress to evaluate the susceptibility to intergranular attack. The electrochemical behavior of the material in the corrosive media was studied by potentiodynamic polarization tests. It was determined that the active species responsible for the stress corrosion cracking of Zircaloy-4 in iodine alcoholic solutions is a molecular complex between the alcohol and iodine. The intergranular attack precedes the 'true' stress corrosion cracking phenomenon (which is associated to the transgranular propagation of the crack) and it is controlled by the diffusion of the active specie to the tip of the crack. Water acts as inhibitor to intergranular attack. Except for methanolic solutions, the minimum water content necessary to inhibit stress corrosion cracking was determined. This critical water content decreases when increasing the molecular weight of the alcohol. An explanation for this behavior is proposed. The susceptibility to stress corrosion cracking also depends on the type of the alcohol used as solvent. The temperature dependence of the crack propagation rate is in agreement with a thermal activated process, and the activation energy is consistent with a process controlled by the volume diffusion of the active species. (author) [es

Air Oxidation Behaviors of Zircaloy-4 Cladding During a LOCA In Spent Fuel Pool

International Nuclear Information System (INIS)

Bang, Je Geon; Chun, Tae Hyun; Kim, Sun Ki; Koo, Yang Hyun

2014-01-01

It is well known that air oxidation induces a serious degradation of the Zircaloy cladding material, compared with steam oxidation. From the oxidant point of view, in comparison with steam, chemical heat release during oxidation in air is higher by 80%, which may lead to a more rapid degradation of the Zircaloy cladding, and further evolution of the accident.. Additionally, the oxidation kinetics in air is much faster than in steam due to the formation of non-protective oxide layer. From the safety point of view, the barrier effect of the cladding against release of fission products is lost much earlier in air compared to steam. The objective of this study is to investigate the oxidation behaviors of fuel cladding in two different conditions such as isothermal and transient condition and to generate its kinetic data under an accident condition in the spent fuel pool. In this study, the oxidation behaviors and its kinetics of the Zircaloy-4 were investigated in air environment for various air ingress scenarios in the temperature range 600 .deg. C-1,400 .deg. C by thermo-gravimetric analysis. In this study, the oxidation behaviors of the Zircaloy-4 for both isothermal condition and transient condition were investigated in air environment. In comparison with isothermal condition, the retardation of oxidation rate in transient condition was observed at both 1,200 .deg. C and 1,400 .deg. C. This seems to be ascribed to the effect of thin oxide formed during a heating

The effect of zinc addition on PWR corrosion product deposition on zircaloy-4

International Nuclear Information System (INIS)

Walters, W.S.; Page, J.D.; Gaffka, A.P.; Kingsbury, A.F.; Foster, J.; Anderson, A.; Wickenden, D.; Henshaw, J.; Zmitko, M.; Masarik, V.; Svarc, V.

2002-01-01

During the period 1995 to 2001 a programme of loop irradiation tests have been performed to confirm the effectiveness of zinc additions on PWR circuit chemistry and corrosion. The programme included two loop irradiation experiments, and subsequent PIE; the experiments were a baseline test (no added zinc) and a test with added zinc (10 ppb). This paper addresses the findings regarding corrosion product deposition and activation on irradiated Zircaloy-4 surfaces. The findings are relevant to overall corrosion of the reactor primary circuit, the use of zinc as a corrosion inhibitor, and activation and transport of corrosion products. The irradiation experience provides information on the equilibration of the loop chemistry, with deliberate injection of zinc. The PIE used novel and innovative techniques (described below) to obtain samples of the oxide from the irradiated Zircaloy. The results of the PIE, under normal chemistry and zinc chemistry, indicate the effect of zinc on the deposition and activation of corrosion products on Zircaloy. It was found that corrosion product deposition on Zircaloy is enhanced by the addition of zinc (but corrosion product deposition on other materials was reduced in the presence of zinc). Chemical analysis and radioisotope gamma counting results are presented, to interpret the findings. A computer model has also been used to simulate the corrosion product deposition and activation, to assist in the interpretation of the results. (authors)

Substructure evolution of Zircaloy-4 during creep and implications for the Modified Jogged-Screw model

Energy Technology Data Exchange (ETDEWEB)

Morrow, B.M., E-mail: morrow@lanl.gov [The Ohio State University, 2041 College Rd., 477 Watts Hall, Columbus, OH 43210 (United States); Los Alamos National Laboratory, P.O. Box 1663, MS G755, Los Alamos, NM 87545 (United States); Kozar, R.W.; Anderson, K.R. [Bettis Laboratory, Bechtel Marine Propulsion Corp., West Mifflin, PA 15122 (United States); Mills, M.J., E-mail: millsmj@mse.osu.edu [The Ohio State University, 2041 College Rd., 477 Watts Hall, Columbus, OH 43210 (United States)

2016-05-17

Several specimens of Zircaloy-4 were creep tested at a single stress-temperature condition, and interrupted at different accumulated strain levels. Substructural observations were performed using bright field scanning transmission electron microscopy (BF STEM). The dislocation substructure was characterized to ascertain how creep strain evolution impacts the Modified Jogged-Screw (MJS) model, which has previously been utilized to predict steady-state strain rates in Zircaloy-4. Special attention was paid to the evolution of individual model parameters with increasing strain. Results of model parameter measurements are reported and discussed, along with possible extensions to the MJS model.

Temperature effect on Zircaloy-4 stress corrosion cracking

International Nuclear Information System (INIS)

Farina, Silvia B.; Duffo, Gustavo S.; Galvele, Jose R.

1999-01-01

Stress corrosion cracking (SCC) susceptibility of Zircaloy-4 alloy in chloride, bromide and iodide solutions with variables as applied electrode potential, deformation rate and temperature have been studied. In those three halide solutions the susceptibility to SCC is only observed at potentials close to pitting potential, the crack propagation rate increases with the increase of deformation rate, and that the temperature has a notable effect only for iodide solutions. For chloride and bromide solutions and temperatures ranging between 20 to 90 C degrees it was not found measurable changes in crack propagation rates. (author)

Irradiation creep and growth behavior of Zircaloy-4 inner shell of HANARO

Energy Technology Data Exchange (ETDEWEB)

Chung, Jong-Ha; Cho, Yeong-Garp; Kim, Jong-In [Korea Atomic Energy Research Inst., Daejeon (Korea, Republic of)

2012-03-15

The inner shell of the reflector vessel of HANARO was made of Zircaloy-4 rolled plate. Zircaloy-4 rolled plate shows highly anisotropic behavior by fast neutron irradiation. This paper describes the analysis method for the irradiation induced creep and growth of the inner shell of HANARO. The anisotropic irradiation creep behavior was modeled as uniaxial strain-hardening power law modified by Hill's stress potential and the anisotropic irradiation growth was modeled by using volumetric swelling with anisotropic strain rate. In this study, the irradiation induced creep and growth behavior of the inner shell of the HANARO reflector vessel was re-evaluated. The rolling direction, the fast neutron flux, and the boundary conditions were applied with the same conditions as the actual inner shell. Analysis results show that deformation of the inner shell due to irradiation does not raise any problem for the lifetime of HANARO. (author)

Mechanical behavior of zircaloy-4 tubes under complexe state of stress

International Nuclear Information System (INIS)

Costa Viana, C.S. da

1980-01-01

The use of zircaloy-4 tubing as cladding material for fuel elements is reviewed with respect to its microstructural, textural and loading conditions. Its anisotropic plastic behaviour is studied through the experimental determination of its yield locus by mechanical testing and Knoop hardness and compared to Hill's anisotropic yield criterion. (Author) [pt

Stress corrosion of Zircaloy-4. Fracture mechanics study of the intergranular - transgranular transition

International Nuclear Information System (INIS)

Farina, Silvia B.; Duffo, Gustavo S.

2003-01-01

Stress corrosion cracking susceptibility of Zircaloy-4 wires was studied in 1M NaCl, 1M KBr and 1M KI aqueous solutions, and in iodine alcoholic solutions. In all cases, intergranular attack preceded transgranular propagation. It is generally accepted that the intergranular-transgranular transition occurs when a critical value of the stress intensity factor is reached. In the present work it was confirmed that the transition from intergranular to transgranular propagation cracking in Zircaloy-4 wires also occurs when a critical value of the stress intensity factor is reached. This critical stress intensity factor in wire samples is independent of the solution tested and close to 10 MPa.m-1/2. This value is in good agreement with those reported in the literature measured by different techniques. (author)

Thermomechanical treatment of β-treated Zircaloy-4 within the upper α-range

International Nuclear Information System (INIS)

Chauvy, C.

2004-09-01

Zircaloy-4 is a Zr base alloy mainly used for nuclear applications. This study deals with its behaviour during forming, with a special interest for physical mechanisms acting in the upper α-range. The β-treated Zircaloy-4 is first described in terms of microstructure and texture. The α plates are organised in colonies with alternating crystallographic orientations. The Burgers relationships have also been checked. The mechanical behaviour shows two distinct domains (with or without hardening). This could be linked to interactions between the solutes (Sn, O) and mobile dislocations. The evolution of crystallographic texture is characterised by X-ray diffraction. At 550 C, twinning is shown to be the main deformation mechanism under specific experimental conditions. Globularization of the initial lamellar structure is identified as a continuous dynamic recrystallization process, more efficient at higher temperature. High strains are necessary to achieve this phenomenon but meta-dynamic recrystallization can also be used to obtain an equiaxed microstructure after limited strains. (author)

Studies on the Electrochemical Dissolution for the Treatment of 10 g-Scale Zircaloy-4 Cladding Hull Wastes in LiCl-KCl Molten Salts

International Nuclear Information System (INIS)

Lee, You Lee; Lee, Jang Hwa; Jeon, Min Ku; Kang, Kweon Ho

2012-01-01

The electrochemical behaviors of 10 g-scale fresh and oxidized Zircaloy-4 cladding hulls were examined in 500 degree C LiCl-KCl molten salts to confirm the feasibility of the electrorefining process for the treatment of hull wastes. In the results of measuring the potential-current response using a stainless steel basket filled with oxidized Zircaloy-4 hull specimens, the oxidation peak of Zr appears to be at -0.7 to -0.8 V vs. Ag/AgCl, which is similar to that of fresh Zircaloy-4 hulls, while the oxidation current is found to be much smaller than that of fresh Zircaloy-4 hulls. These results are congruent with the outcome of current-time curves at -0.78 V and of measuring the change in the average weight and thickness after the electrochemical dissolution process. Although the oxide layer on the surface affects the uniformity and rate of dissolution by decreasing the conductivity of Zircaloy-4 hulls, electrochemical dissolution is considered to occur owing to the defect of the surface and phase properties of the Zr oxide layer.

Investigation of effect of air flow rate on Zircaloy-4 oxidation kinetics and breakaway phenomenon in air at 850 .deg. C

International Nuclear Information System (INIS)

Maeng, Yunhwan; Lee, Jaeyoung; Park, Sanggil

2016-01-01

This paper analyzed an effect of flow rate on oxidation kinetics of Zircaloy-4 in air at 850 .deg. C. In case of the oxidation of Zircaloy-4 in air at 850 .deg. C, acceleration of oxidation kinetics from parabolic to linear (breakaway phenomenon) occurs. Oxidation and breakaway kinetics of the Zircaloy-4 in air was experimentally studied by changing a flow rate of argon/air mixture. Tests were conducted at 850 .deg. C under constant ratio of argon and air. The effects of flow rate on the oxidation and breakaway kinetics was observed. This paper is based on a revised and considerably extended presentation given at the 21 st International Quench Workshop. The effects of flow conditions on the oxidation kinetics of Zircaloy-4 samples were explained with residence time and percent flow efficiency. In addition, several issues were observed from this study, interdiffusion at breakaway and deformation of oxide structure by breakaway phenomenon

Effect of annealing temperature on the mechanical properties of Zircaloy-4 cladding

International Nuclear Information System (INIS)

Beauregard, R.J.; Clevinger, G.S.; Murty, K.L.

1977-01-01

The mechanical properties of Zircaloy cladding materials are sensitive to those fabrication variables which have an effect on the preferred crystallographic orientation or texture of the finished tube. The effect of one such variable, the final annealing temperature, on various mechanical properties is examined using tube reduced Zircaloy-4 fuel rod cladding annealed at temperatures from 905F to 1060F. This temperature range provides cladding with varying degrees of recrystallization including full recrystallization. The burst strength of the cladding at 650F decreased with the annealing temperature reaching a saturation value at approximately 1000F. The total circumferential elongation increased with the annealing temperature reaching a maximum at approximately 1000F and decreasing at higher temperatures. Hoop creep characteristics of Zircaloy cladding were studied as a function of the annealing temperature using closed-end internal pressurization tests at 750F and hoop stresses of 10, 15, 20 and 25 ksi. The effect of annealing temperature on the room temperature mechanical anisotropy parameters, R and P, was studied. The R-parameter was essentially independent of the annealing temperature while the P-parameter increased with annealing temperature. The mechanical anisotropy parameters were also studied as a function of the test temperature from ambient to approximately 800F using continuously monitored high precision extensometry. (Auth.)

Out-of-pile experiments on the high-temperature behavior of Zircaloy-4 clad fuel rods

International Nuclear Information System (INIS)

Hagen, S.

1984-01-01

Out-of-pile experiments have been performed to investigate the escalation in temperature of Zircaloy-clad fuel rods during heatup in steam due to the exothermal Zircaloy steam reaction. In these tests single Zircaloy/uranium dioxide (UO 2 ) fuel rod simulators surrounded with a Zircaloy shroud--simulating the Zircaloy of neighboring rods--were heated inside a fiber ceramic insulation. The initial heating rates were varied from 0.3 to 2.5 K/s. In every test an escalation of the temperature rise rate was observed. The maximum measured surface temperature was about 2200 0 C. The temperature decreased after the maximum had been reached without decreasing the input electric power. The temperature decreases were due to inherent processes including the runoff of molten Zircaloy. The escalation process was influenced by the temperature behavior of the shroud, which was itself affected by the insulation and steam cooling. Damage to the fuel rods increased with increasing heatup rate. Fro slow heatup rates nearly no interaction between the oxidized cladding and UO 2 was observed, while for fast heatup rates the entire annular pellet was dissolved by molten Zircaloy

Mechanical properties and structure of Zircaloy attached by UO2+x and fission products

International Nuclear Information System (INIS)

Holub, F.

1987-08-01

The aim of this project was to determine the combined long-term effect of simulated fission products and hyperstoichiometric uranium dioxide on the mechanical properties and structure of Zircaloy. Three groups of fission product elements or compounds were defined: The rare earth oxides CeO 2 , La 2 O 3 , Nd 2 O 3 , Y 2 O 3 ; The metals No, Ru, Ag; The low melting elements Te, Sb and Cd. Each of these groups of fission products was mixed with UO 2+x in proportion related for burnups of 5, 10 and 30%. The simulated fuel mixtures were filled into tubular Zircaloy casings, plugged and welded. These specimens were annealed at 350, 500 and 700 deg. C up to 17,500 hours. The test results indicate different kinds of action of the simulated fuel constituents. Mixtures of rare earth oxides and UO 2+x embrittle Zircaloy drastically at higher temperatures. There exists a mutual intensifying effect of rare earth oxides and UO 2+x . UO 2+x and (Mo + Ru + Ag) and their mixtures act very similar on Zircaloy. The low melting fission products (Te + Sb + Cd) influence the ductility of Zircaloy in an advantageous manner, compared to pure UO 2+x fuel. The layer of zirconium tellurides seems to protect the Zircaloy metal against the embrittling attack of oxygen from UO 2+x . The most important events of tensile tests at 400 deg. C are the high values of the elongation of specimens which are brittled at room temperature. It should guarantee the integrity of fuel elements, which have been attacked chemically by fission products at temperatures of 400 deg. C and higher

Strengthening of Zircaloy-4 using Oxide Particles by Laser Beam Treatment

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

Oxide particles such as Y{sub 2}O{sub 3} and CeO{sub 2} were dispersed homogeneously in a Zircaloy-4 plate surface using an LBS method. From the tensile test at 380 .deg. C, the strength of laser ODS alloying on the Zircaloy-4 sheet was increased more than 50% when compared to the initial state of the sheet, although the ODS alloyed layer was less than 20% of the specimen thickness. This technology showed a good opportunity to increase the strength without major changes in the substrates of zirconium-based alloys. Accident tolerant fuel (ATF) cladding is being developed globally after the Fukushima accident with the demands for the nuclear fuel having higher safety at normal operation conditions as well as even in a severe accident conditions. Korea Atomic Energy Research Institute (KAERI) is one of the leading organizations for developing ATF claddings. One concept is to form an oxidation-resistant layer on Zr cladding surface. The other is to increase high-temperature mechanical strength of Zr tube. The oxide dispersion strengthened (ODS) zirconium was proposed to increase the strength of the Zr-based alloy up to high temperatures.

Effect of ageing time and temperature on the strain ageing behaviour of quenched zircaloy-4

International Nuclear Information System (INIS)

Rheem, K.S.; Park, W.K.; Yook, C.C.

1977-01-01

The strain ageing behaviour of quenched Zircaloy-4 has been studied as a function of ageing time and temperature in the temperature range 523-588 K for a short-ageing time of 1 to 52 seconds. A the test conditions, the strain ageing stress increased with ageing time and temperature at a strain rate of 5.55x10 -4 sec -1 . Applying stress on the quenched Zircaloy-4, the strain ageing effect indicated following two states: an initial stage having an activation energy of 0.39ev considered to be due to Snoek type ordering of interstitial oxygen atoms in the stress field of a dislocaiton and a second stage havingan activation energy of 0.60 ev, due to mainly long range diffusion of oxygen atoms. (author)

Friction welding method

International Nuclear Information System (INIS)

Ishida, Ryuichi; Hatanaka, Tatsuo.

1969-01-01

A friction welding method for forming a lattice-shaped base and tie plate supporter for fuel elements is disclosed in which a plate formed with a concavity along its edge is pressure welded to a rotating member such as a boss by longitudinally contacting the projecting surfaces remaining on either side of the concavity with the rotating member during the high speed rotation thereof in the presence of an inert gas. Since only the two projecting surfaces of the plate are fused by friction to the rotary member, heat expansion is absorbed by the concavity to prevent distortion; moreover, a two point contact surface assures a stable fitting and promotes the construction of a rigid lattice in which a number of the abovementioned plates are friction welded between rotating members to form any desired complex arrangement. The inert has serves to protect the material quality of the contacting surfaces from air during the welding step. The present invention thus provides a method in which even Zircaloy may be friction welded in place of casting stainless steel in the construction of supporting lattices to thereby enhance neutron economy. (K. J. Owens)

Determination of lower bound crystallographic yield loci of zircaloy-4 tubes

International Nuclear Information System (INIS)

Costa Viana, C.S. da

1980-01-01

The use of zircaloy-4 tubing in fuel elements of water cooled reactors is discussed with respect to its mechanisms of deformation and also its resulting anisotropic plastic behaviour. A method for obtaining lower bound crystallographic yield loci of α-Zr is presented and applied to individual crystal orientations and to a real texture described by the main components observed on a direct pole figure. (Author) [pt

Effects of oxidation in the mechanical behavior of zircaloy-4 tubes

International Nuclear Information System (INIS)

Santos, A.M.M. dos.

1981-07-01

The kinetics of oxidation of zircaloy-4 is isothermally studied utilizing discontinous gravimetric method under two different oxidizing conditions, using gaseous oxigen and steam. The total weight gain during oxidation occurs in two different way: formation of oxide and solid solution. A mechanical test for studying the effect of embrittlement due to the absorption of oxygen in small zircalloy tubes have been developed. (Author) [pt

Effect of annealing temperature on the mechanical properties of zircaloy-4 cladding

International Nuclear Information System (INIS)

Beauregard, R.J.; Clevinger, G.S.; Murty, K.L.

1977-01-01

The mechanical properties of zircaloy cladding materials are sensitive to those fabrication variables which have an effect on the preferred crystallographic orientation or texture of the finished tube. The effect of one such variable, the final annealing temperature, on various mechanical properties is examined using tube reduced zircaloy-4 fuel rod cladding annealed at temperatures from 905F to 1060F. This temperature range provides cladding with varying degrees of recrystallization including full recrystallization. Hoop creep characteristics of zircaloy cladding were studied as a function of the annealing temperature using closed-end internal pressurization tests at 750F and hoop stresses of 10, 15, 20 and 25 ksi. The critical annealing temperature at which a minimum creep strain occurs decreases as the applied stress increases. An additional test at 700F and 30 ksi hoop stress was conducted to demonstrate that the critical annealing temperature is essentially independent of the test temperature. Plausible explanations based on differing substructures developed in cold-worked stress-relieved material are forwarded. The effect of annealing temperature on the room temperature mechanical anisotropy parameters, R and P, was studied. R-parameters were determined from in situ transverse strain gage measurements in uniaxial tensile tests. P-parameters were calculated from uniaxial test data (R and yield stress) and hoop yield stress determined in biaxial, closed-end internal pressurization tests

Oxidation of Zircaloy-4 under limited steam supply at 1000 and 13000C

International Nuclear Information System (INIS)

Uetsuka, H.

1984-12-01

With the view of examining the oxidation behavior of Zircaloy-4 under limited steam supply occurring in severe accidents of LWRs, Zircaloy-4 cladding specimens were examined at the isothermal oxidation temperatures of 1000 and 1300 0 C under a steam atmosphere, flowing at a reduced and constant rate in the range of 3proportional170 mg/cm 2 xmin. The effect of steam starvation, which was restricted to very low levels of steam supply rate, was observed at the two examined temperatures. And the critical supply rate of steam starvation was evaluated to be 13 and 20 mg/cm 2 xmin for the oxidation at 1000 and 1300 0 C, respectively. Variation of the oxidation duration between 2 and 60 min at 1000 0 C allowed to compare the reaction kinetics for three different rates of steam supply. The short-term results confirmed the reduced reaction rates for the lower steam supplies. At the longer times, however, a clear trend towards linear kinetics was observed for the lower supplies. This can be interpreted as the result of earlier breakaway transition under limited steam supply. In the test at 1300 0 C, an acceleration of the oxidation rate was measured for the specified steam supply rate between 20 and 60 mg/cm 2 xmin. This related strongly with high hydrogen concentration in the atmosphere. Hydrogen blanketing - the retarding effect of hydrogen on Zircaloy oxidation - was not identified in the examined temperature range. (orig./HP) [de

The increase in fatigue crack growth rates observed for Zircaloy-4 in a PWR environment

Science.gov (United States)

Cockeram, B. V.; Kammenzind, B. F.

2018-02-01

Cyclic stresses produced during the operation of nuclear reactors can result in the extension of cracks by processes of fatigue. Although fatigue crack growth rate (FCGR) data for Zircaloy-4 in air are available, little testing has been performed in a PWR primary water environment. Test programs have been performed by Gee et al., in 1989 and Picker and Pickles in 1984 by the UK Atomic Energy Authority, and by Wisner et al., in 1994, that have shown an enhancement in FCGR for Zircaloy-2 and Zircaloy-4 in high-temperature water. In this work, FCGR testing is performed on Zircaloy-4 in a PWR environment in the hydrided and non-hydrided condition over a range of stress-intensity. Measurements of crack extension are performed using a direct current potential drop (DCPD) method. The cyclic rate in the PWR primary water environment is varied between 1 cycle per minute to 0.1 cycle per minute. Faster FCGR rates are observed in water in comparison to FCGR testing performed in air for the hydrided material. Hydrided and non-hydrided materials had similar FCGR values in air, but the non-hydrided material exhibited much lower rates of FCGR in a PWR primary water environment than for hydrided material. Hydrides are shown to exhibit an increased tendency for cracking or decohesion in a PWR primary water environment that results in an enhancement in FCGR values. The FCGR in the PWR primary water only increased slightly with decreasing cycle frequency in the range of 1 cycle per minute to 0.1 cycle per minute. Comparisons between the FCGR in water and air show the enhancement from the PWR environment is affected by the applied stress intensity.

Understanding thermally activated plastic deformation behavior of Zircaloy-4

Science.gov (United States)

Kumar, N.; Alomari, A.; Murty, K. L.

2018-06-01

Understanding micromechanics of plastic deformation of existing materials is essential for improving their properties further and/or developing advanced materials for much more severe load bearing applications. The objective of the present work was to understand micromechanics of plastic deformation of Zircaloy-4, a zirconium-based alloy used as fuel cladding and channel (in BWRs) material in nuclear reactors. The Zircaloy-4 in recrystallized (at 973 K for 4 h) condition was subjected to uniaxial tensile testing at a constant cross-head velocity at temperatures in the range 293 K-1073 K and repeated stress relaxation tests at 293 K, 573 K, and 773 K. The minimum in the total elongation was indicative of dynamic strain aging phenomenon in this alloy in the intermediate temperature regime. The yield stress of the alloy was separated into effective and athermal components and the transition from thermally activated dislocation glide to athermal regime took place at around 673 K with the athermal stress estimated to be 115 MPa. The activation volume was found to be in the range of 40 b3 to 160 b3. The activation volume values and the data analyses using the solid-solution models in literature indicated dislocation-solute interaction to be a potential deformation mechanism in thermally activated regime. The activation energy calculated at 573 K was very close to that found for diffusivity of oxygen in α-Zr that was suggestive of dislocations-oxygen interaction during plastic deformation. This type of information may be helpful in alloy design in selecting different elements to control the deformation behavior of the material and impart desired mechanical properties in those materials for specific applications.

Endplate effect on aerodynamic characteristics of three-dimensional wings in close free surface proximity

Directory of Open Access Journals (Sweden)

Jae Hwan Jung

2012-12-01

Full Text Available We investigated the aerodynamic characteristics of a three-dimensional (3D wing with an endplate in the vicinity of the free surface by solving incompressible Navier-Stokes equations with the turbulence closure model. The endplate causes a blockage effect on the flow, and an additional viscous effect especially near the endplate. These combined effects of the endplate significantly reduce the magnitudes of the velocities under the lower surface of the wing, thereby enhancing aerodynamic performance in terms of the force coefficients. The maximum lift-to-drag ratio of a wing with an endplate is increased 46% compared to that of wing without an endplate at the lowest clearance. The tip vortex of a wing-with-endplate (WWE moved laterally to a greater extent than that of a wing-without-endplate (WOE. This causes a decrease in the induced drag, resulting in a reduction in the total drag.

Contribution to study on recovery and recrystallization of cold rolling zircaloy-4

International Nuclear Information System (INIS)

Persiano, A.I.C.

1977-01-01

Recovery and recrystallization of work-hardened (40-60% - Cold rolling) Zircaloy-4 were studied between 200 and 600 0 C with times varying from 15 to 240 minutes, from electrical resistance and hardness measurements. Activation energy calculation for the recovery and recrystallization processes using the samples work-hardened 60% gave 0,7 and 2,1 eV. (author)

A study of stress reorientation of hydrides in zircaloy

Energy Technology Data Exchange (ETDEWEB)

Yourong, Jiang; Bangxin, Zhou [Nuclear Power Inst. of China, Chengdu, SC (China)

1994-10-01

Under the conditions of circumferential tensile stress from 70 to 180 MPa for Zircaloy tubes or the tensile stress from 55 to 180 MPa for Zircaloy-4 plates and temperature cycling between 150 and 400 degree C, the effects of stress and the number of temperature cycling on hydride reorientation in Zircaloy-4 tubes and plates and Zircaloy-2 tubes containing about 220 {mu}g/g hydrogen have been investigated. With the increase of stress and/or the number of temperature cycling, the level of hydride reorientation increases. When hydride reorientation takes place, there is a threshold stress concerned with the number of temperature cycling. Below the threshold stress, hydride reorientation is not obvious. When applied stress is higher than the threshold stress, the level of hydride reorientation increases with the increase of stress and the number of temperature cycling. Hydride reorientation in Zircaloy-4 tubes develops gradually from the outer surface to inner surface. It might be related to the difference of texture between outer surface and inner surface. The threshold stress is affected by both the texture and the value of B. So controlling texture could still restrict hydride reorientation under tensile stress.

Behaviour of High Strength Steel Endplate Connections in Fire and after Fire

NARCIS (Netherlands)

Qiang, X.

2013-01-01

The aim of this research is to reveal more information and understanding on behaviour and failure mechanisms of high strength steel endplate connections (combining high strength steel endplates with either mild steel or high strength steel beams and columns in endplate connections) in fire and after

Review of zircaloy oxidation

Energy Technology Data Exchange (ETDEWEB)

Iglesias, F.C. [Royal Military College of Canada, Kingston, Ontario (Canada); Lewis, B.J. [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

2013-07-01

This paper provides an overview of the kinetics for Zircaloy clad oxidation behaviour in steam and air during reactor accident conditions. The generation of chemical heat from metal/water reaction is considered. The effect of internal clad oxidation due to Zircaloy/UO{sub 2} interaction is also discussed. Low-temperature oxidation of Zircaloy due to water-side corrosion is further described. (author)

Experimental studies on the crystallographic and plastic anisotropies of zircaloy-4 tubes

International Nuclear Information System (INIS)

Costa Viana, C.S. da

1982-01-01

The crystallographic and plastic anisotropies of a zircaloy-4 tubing using direct pole figures and experimental yield loci are analyzed. Tensile and plane-strain compression tests were used to assess the mecahnical behaviour. The results are discussed with respect to the dimensional stability and mechanical behaviour expected for the tube in its use in the core of pressurized water cooled reactors. (Author) [pt

Forming Limit Diagrams of Zircaloy-4 and Zirlo Sheets for Stamping of Spacer Grids of Nuclear Fuel Rods

International Nuclear Information System (INIS)

Seo, Yun Mi; Hyun, Hong Chul; Lee, Hyung Yil; Kim, Nak Soo

2011-01-01

In this work, we investigated the theoretical forming limit models for Zircaloy-4 and Zirlo used for spacer grid of nuclear fuel rods. Tensile and anisotropy tests were performed to obtain stress-strain curves and anisotropic coefficients. The experimental forming limit diagrams (FLD) for two materials were obtained by dome stretching tests following NUMISHEET 96. Theoretical FLD depends on FL models and yield criteria. To obtain the right hand side (RHS) of FLD, we applied the FL models (Swift's diffuse necking, M-K theory, S-R vertex theory) to Zircaloy-4 and Zirlo sheets. Hill's local necking theory was adopted for the left hand side (LHS) of FLD. To consider the anisotropy of sheets, the yield criteria of Hill and Hosford were applied. Comparing the predicted curves with the experimental data, we found that the RHS of FLD for Zircaloy-4 can be described by the Swift model (with the Hill's criterion), while the LHS of the FLD can be explained by Hill model. The FLD for Zirlo can be explained by the S-R model and the Hosford's criterion (a = 8)

In-reactor oxidation of zircaloy-4 under low water vapor pressures

Science.gov (United States)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 °C). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr-4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

In-reactor oxidation of zircaloy-4 under low water vapor pressures

International Nuclear Information System (INIS)

Luscher, Walter G.; Senor, David J.; Clayton, Kevin K.; Longhurst, Glen R.

2015-01-01

Complementary in- and ex-reactor oxidation tests have been performed to evaluate the oxidation and hydrogen absorption performance of Zircaloy-4 (Zr-4) under relatively low partial pressures (300 and 1000 Pa) of water vapor at specified test temperatures (330 and 370 ℃). Data from these tests will be used to support the fabrication of components intended for isotope-producing targets and provide information regarding the temperature and pressure dependence of oxidation and hydrogen absorption of Zr- 4 over the specified range of test conditions. Comparisons between in- and ex-reactor test results were performed to evaluate the influence of irradiation.

Stress corrosion cracking of Zircaloys. Final report

International Nuclear Information System (INIS)

Cubicciotti, D.; Jones, R.L.; Syrett, B.C.

1980-03-01

The overall aim has been to develop an improved understanding of the stress corrosion cracking (SCC) mechanism considered to be responsible for pellet-cladding interaction (PCI) failures of nuclear fuel rods. The objective of the present phase of the project was to investigate the potential for improving the resistance of Zircaloy to iodine-induced SCC by modifying the manufacturing techniques used in the commercial production of fuel cladding. Several aspects of iodine SCC behavior of potential relevance to cladding performance were experimentally investigated. It was found that the SCC susceptibility of Zircaloy tubing is sensitive to crystallographic texture, surface condition, and residual stress distribution and that current specifications for Zircaloy tubing provide no assurance of an optimum resistance to SCC. Additional evidence was found that iodine-induced cracks initiate at local chemical inhomogeneities in the Zircaloy surface, but laser melting to produce a homogenized surface layer did not improve the SCC resistance. Several results were obtained that should be considered in models of PCI failure. The ratio of axial to hoop stress and the temperature were both shown to affect the SCC resistance whereas the difference in composition between Zircaloy-2 and Zircaloy-4 had no detectable effect. Damage accumulation during iodine SCC was found to be nonlinear: generally, a given life fraction at low stress was more damaging than the same life fraction at higher stress. Studies of the thermochemistry of the zirconium-iodine system (performed under US Department of Energy sponsorship) revealed many errors in the literature and provided important new insights into the mechanism of iodine SCC of Zircaloys

Determination of the initial oxidation behavior of Zircaloy-4 by in-situ TEM

International Nuclear Information System (INIS)

Harlow, Wayne; Ghassemi, Hessam; Taheri, Mitra L.

2016-01-01

The corrosion behavior of Zircaloy-4 (Zry-4), specifically by oxidation, is a problem of great importance as this material is critical for current nuclear reactor cladding. The early formation behavior and structure of the oxide layer during oxidation was studied using in-situ TEM techniques that allowed for Zry-4 to be monitored during corrosion. These environmental exposure experiments were coupled with precession electron diffraction to identify and quantify the phases present in the samples before and after the oxidation. Following short-term, high temperature oxidation, the dominant phase was revealed to be monoclinic ZrO 2 in a columnar structure. These samples oxidized in-situ contained structures that correlated well with bulk Zry-4 subjected to autoclave treatment, which were used for comparison and validation of this technique. By using in-situ TEM the effect of microstructure features, such as grain boundaries, on oxidation behavior of an alloy can be studied. The technique presented herein holds the potential to be applied any alloy system to study these effects. - Highlights: • In-situ TEM was used to oxidize samples of Zircaloy-4. • Similar behavior was found in the in-situ oxidized and autoclave-oxidized samples. • Precession diffraction was used to characterize oxide phase and texture.

The effect of papaine on the time course of the end-plate current.

Science.gov (United States)

Humar, M; Kordas, M; Melik, Z

1980-07-01

Papaine is known to detach cholinesterases from the synaptic cleft. It could be expected that this would result in an increase of the amplitude and half-time of the end-plate current. Thus, the effect of papaine on the end-plate current. Thus, the effect of papaine on the end-plate current should be similar to the effect of anticholinesterase methanesulfonylfluoride. The end-plate current was recorded in frog skeletal muscle at various levels of membrane potential, before and after papaine was added to the bath. The effect of papaine was an increase of the half-time of the end-plate current, similarly as after treatment of the muscle by methanesulfonylfluoride. It seems that both papaine and methanesulfonylfluoride have a similar mechanism of action. In either experimental condition hydrolysis of transmitter is decreased or abolished, which results in an increase of the half-time of the end-plate current.

Influence of Zircaloy cladding composition on hydride formation during aqueous hydrogen charging

Energy Technology Data Exchange (ETDEWEB)

Rajasekhara, S. [Intel Corporation, 2501 NW 229th Av., Hillsboro, OR 97124 (United States); Kotula, P.G.; Enos, D.G.; Doyle, B.L. [Sandia National Laboratories, Albuquerque, NM, 87185 (United States); Clark, B.G., E-mail: blyclar@sandia.gov [Sandia National Laboratories, Albuquerque, NM, 87185 (United States)

2017-06-15

Although hydrogen uptake in Zirconium (Zr) based claddings has been a topic of many studies, hydrogen uptake as a function of alloy composition has received little attention. In this work, commercial Zr-based cladding alloys (Zircaloy-2, Zircaloy-4 and ZIRLO™), differing in composition but with similar initial textures, grain sizes, and surface roughness, were aqueously charged with hydrogen for 100, 300, and 1000 s at nominally 90 °C to produce hydride layers of varying thicknesses. Transmission electron microscope characterization following aqueous charging showed hydride phase and orientation relationship were identical in all three alloys. However, elastic recoil detection measurements confirmed that surface hydride layers in Zircaloy-2 and Zircaloy-4 were an order of magnitude thicker relative to ZIRLO™. - Highlights: •Aqueous charging was performed to produce a layer of zirconium hydride for three different Zr-alloy claddings. •Hydride thicknesses were analyzed by elastic recoil detection and transmission electron microscopy. •Zircaloy-2 and Zircaloy-4 formed thicker hydride layers than ZIRLO™ for the same charging durations.

The long range migration of hydrogen through Zircaloy in response to tensile and compressive stress gradients

International Nuclear Information System (INIS)

Kammenzind, B.F.; Berquist, B.M.; Bajaj, R.; Kreyns, P.H.; Franklin, D.G.

1998-01-01

Zircaloy-4, which is used widely as a core structural material in pressurized water reactors (PWRs), picks up hydrogen during service. Hydrogen solubility in Zircaloy-4 is low and zirconium hydride phases precipitate after the Zircaloy-4 lattice becomes supersaturated with hydrogen. These hydrides embrittle the Zircaloy-4, degrading its mechanical performance as a structural material. Because hydrogen can move rapidly through the Zircaloy-4 lattice, the potential exists for large concentrations of hydride to accumulate in local regions of a Zircaloy component remote from its point of entry into the component. Much has been reported in the literature regarding the long range migration of hydrogen through Zircaloy under concentration gradients and temperature gradients. Relatively little has been reported, however, regarding the long range migration of hydrogen under stress gradients. This paper presents experimental results regarding the long range migration of hydrogen through Zircaloy in response to both tensile and compressive stress gradients. The importance of this driving force for hydrogen migration relative to concentration and thermal gradients is discussed

Extrusion and drawing of zircaloy 2. Production of pressure tubes for EL-4

International Nuclear Information System (INIS)

Thevenet, J.

1964-01-01

The authors give briefly the physical mechanical and chemical properties of zircaloy 2, as far as the transformation of this alloy is concerned. Extrusion: After a few general remarks concerning the extrusion and co-extrusion, including a comparison of the deformation resistance of canning metals and of zircaloy 2, the following points are considered: - the difficulties occurring because of the use of this alloy: - atmosphere protection - adjustment on to the machine tools - low thermal conductivity - economy of the metal (price) - the factors affecting the quality of the extruded products extrusion under a copper can and under lubricant glass - fine grain structure - temperature homogeneity - working temperature The transformation cycle - '550 kg ingot - preliminary shape 'for drawing of EL-4 tubes (112 x 120 L 12 m)' - is described in detail (extrusion or forging of the φ = 340 ingot into φ = 220 billets, cutting into lengths and hot drilling at φ = 125, fixing into a copper can and rough extrusion). Drawing: The main difficulties are due to seizing of the tools and to the necessity of protecting the alloy from the atmosphere during annealings. A brief description is given of drawing out on a short mandrel, on a long mandrel, of laminating on a reducing machine and of the carrying out of an annealing, as well as of the production of EL-4 tubes (φ =107 x 113 L 430 m) by drawing out shapes having a size of 112 x 120 on long mandrels. Conclusion: It is possible by extrusion and drawing to produce zircaloy 2 tubes similar to those which may be obtained normally using stainless steel. (authors) [fr

Zircaloy-steam reaction under a simulated loss-of-coolant accident

International Nuclear Information System (INIS)

Kawasaki, Satoru; Furuta, Teruo; Hashimoto, Masao

1975-07-01

Under a simulated loss-of-coolant condition, the reaction between zircaloy and steam and the embrittlement of the zircaloy oxidized by this reaction have been studied. The parabolic rate constant, ksub(p), in the zircaloy-steam reaction is represented as ksub(p)=3.24x10 6 exp(-40500/RT) (mg 2 /cm 4 . sec) Ring compression test was made on the steam-reacted zircaloy tubes, and following results were obtained: Embrittlement of the steam-reacted zircaloy tube increases with oxidation at each oxidation temperature. For a given quantity of the oxidation, the incursion of α-phase into β-phase is more remarkable in the specimens reacted at low temperatures than those at high temperatures. The embrittlement, however, is larger in the specimens oxidized at high temperatures than those at low temperatures. (auth.)

Performance assessment on high strength steel endplate connections after fire

NARCIS (Netherlands)

Qiang, X.; Wu, N.; Jiang, X.; Bijlaard, F.S.K.; Kolstein, M.H.

2017-01-01

Purpose – This study aims to reveal more information and understanding on performance and failure mechanisms of high strength steel endplate connections after fire. Design/methodology/approach – An experimental and numerical study on seven endplate connections after

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

International Nuclear Information System (INIS)

Vaibhaw, Kumar; Rao, S.V.R.; Jha, S.K.; Saibaba, N.; Jayaraj, R.N.

2008-01-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (∼300 deg. C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation (F n ) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process

Texture and hydride orientation relationship of Zircaloy-4 fuel clad tube during its fabrication for pressurized heavy water reactors

Science.gov (United States)

Vaibhaw, Kumar; Rao, S. V. R.; Jha, S. K.; Saibaba, N.; Jayaraj, R. N.

2008-12-01

Zircaloy-4 material is used for cladding tube in pressurized heavy water reactors (PHWRs) of 220 MWe and 540 MWe capacity in India. These tubes are fabricated by using various combinations of thermo-mechanical processes to achieve desired mechanical and corrosion properties. Cladding tube develops crystallographic texture during its fabrication, which has significant influence on its in-reactor performance. Due to radiolytic decomposition of water Zircaloy-4 picks-up hydrogen. This hydrogen in excess of its maximum solubility in reactor operating condition (˜300 °C), precipitates as zirconium hydrides causing embrittlement of cladding tube. Hydride orientation in the radial direction of the tube limits the service life and lowers the fuel burn-up in reactor. The orientation of the hydride primarily depends on texture developed during fabrication. A correlation between hydride orientation ( F n) with the texture in the tube during its fabrication has been developed using a second order polynomial. The present work is aimed at quantification and correlation of texture evolved in Zircaloy-4 cladding tube using Kearn's f-parameter during its fabrication process.

Effect of current density on the anodic behaviour of zircaloy-4 and niobium: a comparative study

International Nuclear Information System (INIS)

Raghunath Reddy, G.; Lavanya, A.; Ch Anjaneyulu

2004-01-01

The kinetics of anodic oxidation of zircaloy-4 and niobium have been studied at current densities ranging from 2 to 14 mA.cm -2 at room temperature in order to investigate the dependence of ionic current density on the field across the oxide film. Thickness of the anodic films were estimated from capacitance data. The formation rate, current efficiency and differential field were found to increase with increase in the ionic current density for both zircaloy-4 and niobium. Plots of the logarithm of formation rate vs. logarithm of the current density are fairly linear. From linear plots of logarithm of ionic current density vs. differential field, and applying the Cabrera-Mott theory, the half-jump distance and the height of the energy barrier are deduced and compared. (author)

Precipitates in irradiated Zircaloy

International Nuclear Information System (INIS)

Chung, H.M.

1985-10-01

Precipitates in high-burnup (>20 MWd/kg U) Zircaloy spent-fuel cladding discharged from commercial boiling- and pressurized-water reactors have been characterized by TEM-HVEM. Three classes of primary precipitates were observed in the irradiated Zircaloys: Zr 3 O (2 to 6 nm), cubic-ZrO 2 (greater than or equal to 10 nm), and delta-hydride (35 to 100 nm). The former two precipitations appears to be irradiation induced in nature. Zr(Fe/sub x/Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/Ni/sub 1-x/) intermetallics, which are the primary precipitates in unirradiated Zircaloys, were largely dissolved after the high burnup. It seems, therefore, that the influence of the size and distribution of the intermetallics on the corrosion behavior may be quite different for the irradiated Zircaloys

Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

International Nuclear Information System (INIS)

Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho

2016-01-01

Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

Crack behavior of oxidation resistant coating layer on Zircaloy-4 for accident tolerant fuel claddings

Energy Technology Data Exchange (ETDEWEB)

Park, Jung Hwan; Kim, Eui Jung; Jung, Yang Il; Park, Dong Jun; Kim, Hyun Gil; Park, Jeong Yong; Yang, Jae Ho [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Terrani et al. reported the oxidation resistance of Fe-based alloys for protecting zirconium alloys from the rapid oxidation in a high-temperature steam environment. Kim and co-workers also reported the corrosion behavior of Cr coated zirconium alloy using a plasma spray and laser beam scanning. Cracks are developed by tensile stress, and this significantly deteriorates the oxidation resistance. This tensile stress is possibly generated by the thermal cycle or bending or the irradiation growth of zirconium. In this study, Cr was deposited by AIP on to Zircaloy-4 plate, and the crack behavior of Cr coated Zircaloy-4 under uni-axial tensile strain was observed. In addition, the strain of the as-deposited state was calculated by iso-inclination method. Coating began to crack at 8% of applied strain. It is assumed that a well-densified structure by AIP tends to be resistant to cracking under tensile strain.

Obtaining zircaloy powder through hydriding

International Nuclear Information System (INIS)

Dupim, Ivaldete da Silva; Moreira, Joao M.L.

2009-01-01

Zirconium alloys are good options for the metal matrix in dispersion fuels for power reactors due to their low thermal neutron absorption cross-section, good corrosion resistance, good mechanical strength and high thermal conductivity. A necessary step for obtaining such fuels is producing Zr alloy powder for the metal matrix composite material. This article presents results from the Zircaloy-4 hydrogenation tests with the purpose to embrittle the alloy as a first step for comminuting. Several hydrogenation tests were performed and studied through thermogravimetric analysis. They included H 2 pressures of 25 and 50 kPa and temperatures ranging between from 20 to 670 deg C. X-ray diffraction analysis showed in the hydrogenated samples the predominant presence of ZrH 2 and some ZrO 2 . Some kinetics parameters for the Zircaloy-4 hydrogenation reaction were obtained: the time required to reach the equilibrium state at the dwell temperature was about 100 minutes; the hydrogenation rate during the heating process from 20 to 670 deg C was about 21 mg/h, and at constant temperature of 670 deg C, the hydride rate was about 1.15 mg/h. The hydrogenation rate is largest during the heating process and most of it occurs during this period. After hydrogenated, the samples could easily be comminuted indicating that this is a possible technology to obtain Zircaloy powder. The results show that only few minutes of hydrogenation are necessary to reach the hydride levels required for comminuting the Zircaloy. The final hydride stoichiometry was between 2.7 and 2.8 H for each Zr atom in the sample (author)

High temperature properties of Zircaloy--oxygen alloys

International Nuclear Information System (INIS)

Mellinger, G.B.; Bates, J.L.

1977-03-01

The effect of oxygen on three properties of Zircaloy-4 cladding relevant to LOCA evaluation codes was determined. Thermal expansion, elastic moduli, and thermal diffusivity were measured over the range room temperature--1200 0 C (2192 0 F) and 0.7 to 28 at.% oxygen. Thermal expansion and elastic moduli showed increases with oxygen concentration, while thermal diffusivity tended to decrease. Zircaloy-2 was examined over the same temperature range, but only to 5 at.% oxygen, differences in the properties between the two alloys were minor. The thermal emittance of Zircaloy-4 was measured in argon over the wavelength range 1.5 to 2.5 μm on previously oxidized tubing and on surfaces in the process of oxidizing in unlimited steam. For the latter, a high emittance (approximately 0.9) was reached at an oxide thickness of about 100 mg/dm 2 , and the tubing surface remained black and substoichiometric as oxidation continued at temperatures to 1200 0 C

Preliminary design and manufacturing feasibility study for a machined Zircaloy triangular pitch fuel rod support system (grids) (AWBA development program)

International Nuclear Information System (INIS)

Horwood, W.A.

1981-07-01

General design features and manufacturing operations for a high precision machined Zircaloy fuel rod support grid intended for use in advanced light water prebreeder or breeder reactor designs are described. The grid system consists of a Zircaloy main body with fuel rod and guide tube cells machined using wire EDM, a separate AM-350 stainless steel insert spring which fits into a full length T-slot in each fuel rod cell, and a thin (0.025'' or 0.040'' thick) wire EDM machined Zircaloy coverplate laser welded to each side of the grid body to retain the insert springs. The fuel rods are placed in a triangular pitch array with a tight rod-to-rod spacing of 0.063 inch nominal. Two dimples are positioned at the mid-thickness of the grid (single level) with a 90 0 included angle. Data is provided on the effectiveness of the manufacturing operations chosen for grid machining and assembly

Study of the mechanisms controlling the oxide growth under irradiation: characterization of irradiated zircaloy-4 and Zr-1 Nb-O oxide scales

International Nuclear Information System (INIS)

Bossis, Ph.; Thomazet, J.; Lefebvre, F.

2002-01-01

In PWRs, the Zr-1Nb-O alloy shows a marked enhancement in corrosion resistance in comparison with Zircaloy-4. The aim of this work is to analyze the reasons for these different behaviors and to determine the respective nature of the oxide growth controlling mechanisms under irradiation. Samples taken from Zircaloy-4 irradiated 1, 2, and 4 cycles and Zr-1Nb-O irradiated 1 and 3 cycles have been systematically characterized by optical microscopy, SEM coupled with image analysis, hydride distribution, and XRD. Specific TEM characterizations have been performed on the Zr-1Nb-O samples. A XPS analysis of a nonirradiated sample is also reported. It has been shown that under irradiation the slow oxidation kinetics of the Zr-1Nb-O alloy is associated with very regular metal-oxide interface and oxide layer. On the contrary, the accelerated oxidation kinetics of Zircaloy-4 is associated with highly perturbed metal-oxide interface and oxide layer. On both irradiated alloys, cracks are observed to initiate preferentially above the delayed parts of the oxidation front. Hydrogen intake during water oxidation in PWR environment is found to be much lower on the Zr-1Nb-O alloy than on Zircaloy-4. More β-ZrO 2 is found on the oxide layer formed on Zircaloy-4 than on Zr-1NbO after oxidation in PWR. Classical irradiation-induced microstructural evolution is observed in the Zr-1Nb-O metallic alloy after 3 cycles, i.e., a fine β-Nb precipitation. β-Nb precipitates are observed to undergo a delayed oxidation associated with a crystalline to amorphous transformation. After water oxidation in autoclave, a pronounced Nb segregation is detected on the oxide surface of a Zr-1Nb-O sample. These results suggest that the oxidation kinetics of Zircaloy-4 is controlled essentially by oxygen diffusion through the inner barrier layer, which is significantly accelerated under irradiation. The oxidation kinetics of Zr-1Nb-O is controlled by both oxygen diffusion through the inner barrier and by

X-ray radiography of Ti6Al4V welded by plasma tungsten arc (PTA) welding

Energy Technology Data Exchange (ETDEWEB)

Dikbas, Halil; Caligulu, Ugur; Taskin, Mustafa; Turkmen, Mustafa [Firat Univ., Elazig (Turkey). Metallurgy Dept.

2013-03-01

In this study, X-ray radiographic tests of Ti6Al4V alloys welded by plasma tungsten arc welding (PTA) were investigated. PTA welding experiments were carried out under argon shielding gas atmosphere, at 1400-1600 W and 1800 W welding powers as well as 1 m/min, 0.75 m/min, and 0.50 m/min welding speeds. After this process, radiography of the welded joints was performed by X-ray diffraction. The result of the radiographic tests indicated that by increasing welding power the widths of deep penetration increased in all specimens. On the contrary, increasing welding speeds decreases the widths deep penetration. The best properties of Ti6Al4V joints were observed for specimens welded at 1800 W welding power and at 0.50 m/min welding speed. (orig.)

A review on the welding technology for the sealing of irradiation test fuel element

Energy Technology Data Exchange (ETDEWEB)

Lee, J. W.; Kang, Y. H.; Kim, B. G.; Joo, K. N.; Oh, J. M.; Park, S. J.; Shin, Y. T

2000-02-01

For the irradiation test of nuclear fuel in a research reactor, the fuel manufacturing technology should be developed in advance. Highly radioactive fission products are produced and can be released from the fuel materials during irradiation. Therefore, The sealing of the test is one of the most important procedure among the test fuel manufacturing processes, considering its impacts on the safety of a reactor operation.many welding techniques such as TIG, EBW, LBW, upset butt welding and flash welding are applied in sealing the end of fuel elements. These welding techniques are adopted in conjunction with the weld material, weldability, weld joint design and cost effectiveness. For fuel irradiation test, the centerline temperature of fuel pellets is one of the important item to be measured. For this, a thermocouple is installed into the center of the fuel pellet. The sealing of the penetration hole of the thermocouple sheath should be conducted and the hole should be perfectly sealed using the dissimilar metal joining technique. For this purpose, the dissimilar metal welding between zircaloy-4 and Inconel or stainless steel is needed to be developed. This report describes the techniques sealing the end cap and the penetration of a thermocouple sheath by welding. (author)

Zircaloy-4 stress corrosion by iodine: crack kinetics and influence of irradiation on the crack initiation

International Nuclear Information System (INIS)

Serres, A.

2008-01-01

During the PWR power transients, iodine-induced stress corrosion cracking (I-SCC) is one of the potential failure modes of Zircaloy-4 fuel claddings under Pellet-Cladding Interaction conditions. The primary objective of this study is to distinguish the parameters that contribute to the I-SCC phenomenon in iodized methanol solutions at ambient temperature, on notched tensile specimens, using crack growth rate measurements provided by Direct Current Potential Drop. The results show that for a KI lower than 20 MPa.m 1/2 , the IG and mixed IG/TG velocity of propagation is a linear function of KI, regardless of the propagation mode. Between 20 and 25 MPa.m 1/2 , the TG crack growth rate also depends linearly on KI, but increases at a faster rate with respect to KI than during the IG and mixed IG/TG propagation steps. The crack propagation direction and plane (LT and TL) have an impact on the propagation modes, but no impact on the kinetics. The increase of iodine content induces an increase of the crack growth rate for a given KI, and a decrease of the KI, threshold, allowing the crack propagation. This work enables us to quantify the effect of iodine content and of KI on the crack propagation step, propose a propagation law taking into accounts these parameters, and improve the I-SCC description for models. During operation, a zirconium cladding is neutron-irradiated, modifying its microstructure and deformation modes. The second objective of the study is therefore to investigate the impact of these modifications on I-SCC. For that purpose, smooth specimens in recrystallized Zircaloy-4 are proton-irradiated to 2 dpa at 305 C, the microstructure and deformation modes of unirradiated and irradiated Zircaloy-4 are characterized by TEM and SEM, and the influence of these radiation-induced modifications on the I-SCC susceptibility is studied. The Laves phases precipitates are slightly modified by irradiation. The formation of P -type dislocation loops correlated with

Spectrophotometric determination of uranium traces in zircaloy-4 and zirconium sponge

International Nuclear Information System (INIS)

Correia, R.J.; Weber de D'Alessio, Ana; Zucal, R.H.

1980-01-01

The uranium contents of the zircaloy-4 which is used for the fabrication of the fuel cans for the PHWR Atucha and Embalse nuclear power stations must not exceed 3.ppM. A method was developed for performing that control, involving the separation of the uranium from its matrix by partition chromatography and its determination by spectrophotometry with Arsenazo (III). This method is applied within the range of 0.2 to 10 ppM, obtaining a relative standard deviation of 6% for U contents of 3 ppm. (M.E.L.) [es

Determinations of the temperature of terminal solid solubility in dissolution and precipitation of hydrogen/deuterium in irradiated Zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Vizcaino, P [CNEA-CONICET, Centro Atomico Ezeiza (Argentina)

2012-07-01

The proposed plan is an approach to the metallurgical consequences of the high neutron fluencies (10''2''2 n/cm''2) on the hydrogen behavior in zirconium based alloys, based on the significance of the microstructural behavior of the high burn up fuel claddings during the dry storage period. The studies are focused on Zircaloy-4, concerning to two processes: Neutron irradiation damage; Hydrogen pick up. The Zircaloy-4 was taken from cooling channels of the PHWR Atucha 1. These components remained more than 10 years in service, reaching neutron fluencies up to 10''2''2 n/cm''2. In the last recent years, measurements of the hydride dissolution temperatures have shown that hydrogen solubility is affected by the neutron irradiation, increasing it respect to the unirradiated Zircaloy solubility. In addition, in this material the amorphization/dissolution of the second phase particles (SPPs) was observed, being proposed an interaction between the hydrogen atoms, the SPPs and the irradiation defects as a possible explanation of the observed behavior. For the present case, attention will be focused on the hydride precipitation process, since it is strongly related with delay hydrogen cracking initiation, a problem of direct concern for the dry storage. The goal of the present proposal is to make an approach to the source of the observed effect, applying several specific techniques as differential scanning calorimetry (DSC), high resolution x-ray diffraction and transmission electron microscopy. The objectives can be divided as follows: Determination of the temperatures of terminal solid solubility in dissolution (TTSSd) and in precipitation (TTSSp) in high fluency irradiated Zircaloy-4, reproducing the temperatures at which the Zircaloy fuel claddings remain during dry storage by an annealing program during the DSC experiments; Observations by optical and transmission electron microscopy of the hydride distribution before (as received material) and after high temperature

Effect of water content on the stress corrosion cracking susceptibility of Zircaloy-4 in iodine-alcoholic solutions

International Nuclear Information System (INIS)

Gomez Sanchez, Andrea; Farina, Silvia B.; Duffo, Gustavo S.

2005-01-01

The stress corrosion cracking (SCC) susceptibility of Zircaloy-4 (UNS R60804) was studied in 10 g/L iodine dissolved in various alcohols: methanol, ethanol, 1 propanol, 1-butanol, 1-pentanol and 1-octanol. SCC was observed in all the systems studied and it was found that the higher the size of alcohol molecule, the lower the SCC susceptibility. The existence of intergranular attack -controlled by the diffusion of the active species- is a condition for the SCC process to occur. In the present work the inhibiting effect of water on the SCC susceptibility of Zircaloy-4 in iodine-alcoholic solutions was also investigated and the results showed that the minimum water content to inhibit the SCC process depends on the type of alcohol used as a solvent. (author) [es

Pressure effects on high temperature steam oxidation of Zircaloy-4

International Nuclear Information System (INIS)

Park, Kwangheon; Kim, Kwangpyo; Ryu, Taegeun

2000-01-01

The pressure effects on Zircaloy-4 (Zry-4) cladding in high temperature steam have been analyzed. A double layer autoclave was made for the high pressure, high temperature oxidation tests. The experimental test temperature range was 700 - 900 deg C, and pressures were 0.1 - 15 MPa. Steam partial pressure turns out to be an important one rather than total pressure. Steam pressure enhances the oxidation rate of Zry-4 exponentially. The enhancement depends on the temperature, and the maximum exists between 750 - 800 deg C. Pre-existing oxide layer decreases the enhancement about 40 - 60%. The acceleration of oxidation rate by high pressure team seems to be originated from the formation of cracks by abrupt transformation of tetragonal phase in oxide, where the un-stability of tetragonal phase comes from the reduction of surface energy by steam. (author)

Relationship between screw sagittal angle and stress on endplate of adjacent segments after anterior cervical corpectomy and fusion with internal fixation: a Chinese finite element study.

Science.gov (United States)

Zhang, Yu; Tang, Yibo; Shen, Hongxing

2017-12-01

In order to reduce the incidence of adjacent segment disease (ASD), the current study was designed to establish Chinese finite element models of normal 3rd~7th cervical vertebrae (C3-C7) and anterior cervical corpectomy and fusion (ACCF) with internal fixation , and analyze the influence of screw sagittal angle (SSA) on stress on endplate of adjacent cervical segments. Mimics 8.1 and Abaqus/CAE 6.10 softwares were adopted to establish finite element models. For C4 superior endplate and C6 inferior endplate, their anterior areas had the maximum stress in anteflexion position, and their posterior areas had the maximum stress in posterior extension position. As SSA increased, the stress reduced. With an increase of 10° in SSA, the stress on anterior areas of C4 superior endplate and C6 inferior endplate reduced by 12.67% and 7.99% in anteflexion position, respectively. With an increase of 10° in SSA, the stress on posterior areas of C4 superior endplate and C6 inferior endplate reduced by 9.68% and 10.22% in posterior extension position, respectively. The current study established Chinese finite element models of normal C3-C7 and ACCF with internal fixation , and demonstrated that as SSA increased, the stress on endplate of adjacent cervical segments decreased. In clinical surgery, increased SSA is able to play important role in protecting the adjacent cervical segments and reducing the incidence of ASD.

Analysis of the tensile behaviour of zircaloy-4 in the region of dynamic strain aging

International Nuclear Information System (INIS)

Dellaretti Filho, O.

1974-01-01

An analysis of the tensile behavior of Zircaloy 4, centering around the influence of dynamic strain aging and strain rate history, is presented. This analysis is based on techniques introduced by Jaoul-Crussard and Reed-Hill. An attempt is also made to assess the experimental errors that influence these methods. (author)

The effects of irradiation and temperature on the growth of Zircaloy-4 tubes

International Nuclear Information System (INIS)

Kendoush, A.A.

1987-01-01

The growth strain was measured after irradiation for 16 Zircaloy-4 tubes of the recrystallised and stress relieved types. The operating temperature during irradiation ranged between 317 and 344 0 C. The average fast neutron fluence was 9.6x10 20 n/cm 2 . Experimental results indicated the dependence of the growth on the irradiation temperature. The stress relieved result was compared with data of the literature. (orig.)

Research progress of aluminium alloy endplates for PEMFCs

Energy Technology Data Exchange (ETDEWEB)

Fu, Yu.; Hou, Junbo [Fuel Cell system and Engineering Laboratory, Dalian Institute of Chemical and Physics, Chinese Academy of Sciences, Dalian 116023 (China); Graduate University of Chinese Academy of Sciences, Beijing 100049 (China); Hou, Ming; Yan, Xiqiang; Luo, Xiaokuan; Shao, Zhigang; Yi, Baolian [Fuel Cell system and Engineering Laboratory, Dalian Institute of Chemical and Physics, Chinese Academy of Sciences, Dalian 116023 (China)

2007-04-15

The endplate is a crucial component in a proton exchange membrane fuel cell (PEMFC) stack. It can provide the necessary rigidity and strength for the stack. An aluminium alloy is one of the ideal materials for PEMFC endplates because of its low density and high rigidity. But it does not meet the requirements of corrosion resistance and electrical insulation in PEMFC environments. In this work, methods of sealing treatments and the conditions of aluminium alloy anodization were investigated. Corrosion resistances of the samples prepared by different technologies were evaluated in simulated PEMFC environments. The results showed that the corrosion resistance of the samples sealed by epoxy resin was greatly improved compared with those sealed in boiling water, and the samples anodized at a constant current density performed better than those anodized at a constant voltage. By insulation measurements, all of the samples showed good electrical insulation. The aluminium alloy endplate anodized at a constant current density and sealed with thermosetting bisphenol-A epoxy resin exhibited promising potential for practical applications by assembling it in a PEMFC stack and applying a life test. (author)

Cladding the inside surface of a 3 1/4 in. ID Zircaloy-2 pressure tube with 1S aluminum

International Nuclear Information System (INIS)

Watson, R.D.

1966-09-01

A hot-press sizing technique has been developed for cladding the inside surface of Zircaloy-2 pressure tubes with 1S aluminum. The process is performed in air with the Zircaloy-2 and aluminum at a temperature of approximately 950 o F. A controlled atmosphere is not required, either during preheating or while the cladding is being applied. Tubes 30 inches long and 3 1/4 inches ID have been coated with 1S aluminum in thicknesses ranging from 0.005 inches to more than 0.02 inches; tubes longer than 30 inches have not been attempted. The lining of aluminum is firmly attached to the Zircaloy-2 at all points in the tube but the bond strength varies considerably - from. 6500 to 28000 lbf/in 2 . This work is the subject of Canadian Patent Application No. 955,358 filed March 21, 1966. (author)

Heat transfer coefficient between UO2 and Zircaloy-2

International Nuclear Information System (INIS)

Ross, A.M.; Stoute, R.L.

1962-06-01

This paper provides some experimental values of the heat-transfer coefficient between UO 2 and Zircaloy-2 surfaces in contact under conditions of interfacial pressure, temperature, surface roughness and interface atmosphere, that are relevant to UO 2 /Zircaloy-2 fuel elements operating in pressurized-water power reactors. Coefficients were obtained from eight UO 2 / Zircaloy-2 pairs in atmospheres of helium, argon, krypton or xenon, at atmosphere pressure and in vacuum. Interfacial pressures were varied from 50 to 550 kgf/cm 2 while surface roughness heights were in the range 0.2 x 10 -4 to 3.5 x 10 -4 cm. The effect on the coefficients of cycling the interfacial pressure, of interface gas pressure and of temperature were examined. The experimental values of the coefficients were used to test the predictions of expressions for the heat-transfer between two solids in contact. For the particular UO 2 / Zircaloy-2 pairs examined, numerical values were assigned to several parameters that related the surface roughnesses to either the radius of solid/solid contact spots or to the mean thickness of the interface voids and that accounted for the imperfect accommodation of the void gas on the test surfaces. (author)

Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Gu, E-mail: jglee88@ulsan.ac.kr [School of Materials Science and Engineering, University of Ulsan, Ulsan 44610 (Korea, Republic of); Lee, Gyoung-Ja; Park, Jin-Ju [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 34057 (Korea, Republic of); Lee, Min-Ku, E-mail: leeminku@kaeri.re.kr [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 34057 (Korea, Republic of)

2017-05-15

The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments. - Highlights: •Corrosion of Zircaloy-4 joints brazed with Zr-Cu-X filler alloys was investigated. •Alloyed Al deteriorated the overall nobility of joints by microgalvanic reaction. •Compositional gradient of Al in joints was the driving force for galvanic corrosion. •Cu and Fe did not influence the electrochemical stability of joints. •Zr-Cu-Fe filler alloy yielded excellent high-temperature corrosion resistance.

Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

International Nuclear Information System (INIS)

Lee, Jung Gu; Lee, Gyoung-Ja; Park, Jin-Ju; Lee, Min-Ku

2017-01-01

The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments. - Highlights: •Corrosion of Zircaloy-4 joints brazed with Zr-Cu-X filler alloys was investigated. •Alloyed Al deteriorated the overall nobility of joints by microgalvanic reaction. •Compositional gradient of Al in joints was the driving force for galvanic corrosion. •Cu and Fe did not influence the electrochemical stability of joints. •Zr-Cu-Fe filler alloy yielded excellent high-temperature corrosion resistance.

Corrosion kinetic of 2 and 4 zircaloys in air at high temperatures

International Nuclear Information System (INIS)

Goncalves, A.C.; Goncalves, Z.C.

1986-01-01

The corrosion results of 2 and 4 zircaloys obtained in a thermal balance between 500 and 850 0 C are discussed based on the model of 'reduction of diffusion path'. The behaviour of both alloys has shown almost similar in this interval of temperature, proving that the corrosion is processed by an identical kinetic mechanism. It is still analysed the formation of superposed layer of porous oxide and the possible influence of the oxygen partial pressure in inversion velocities between 750 and 800 0 C. (Author) [pt

Residual stresses in 2 1/4Cr1Mo welds

International Nuclear Information System (INIS)

Fidler, R.; Jerram, K.

1978-01-01

Two separate investigations, initiated in an attempt to explain the large amount of residual stress scatter previously observed in the weld metal of eighteen nominally identical thick-section 2 1/4Cr1Mo butt welds, are described in this paper. The first examined the detailed surface residual stress distributions in 2 1/4Cr1Mo manual arc circumferential butt welds in 80mm and 100mm thick 1/2Cr1/2Mo1/4V steam pipe. High residual stresses were found in the regions of overlap between adjacent weld beads, with low values in virgin weld metal. The second utilised single pass manual metal arc bead-in-groove welds to investigate the effects of preheat and weld metal composition on weld metal residual stresses. In four weld metals, mild steel, 1/2Cr1/2Mo1/4V, 1Cr1/2Mo, and 2 1/4Cr1Mo, the residual stresses were very similar, becoming less tensile (or more compressive) with increase of preheat, while the residual stresses in the fifth weld metal (12Cr) were significantly different, being compressive and less affected by preheat. In both investigations the effects have been described in terms of the basic metallurgical phenomena occurring in the weld metal. (author)

Investigation of flow condition on the oxidation of Zircaloy-4 in air at 850 and 1100 .deg. C

Energy Technology Data Exchange (ETDEWEB)

Maeng, Yun Hwan; Lee, Jae Young [Hangdong Global University, Pohang (Korea, Republic of); Park, Sang Gil [ACT Co. Ltd, Daejeon (Korea, Republic of)

2016-05-15

An oxidation behavior of the Zircaloy-4 was experimentally studied by varying a flow rate and partial pressure of air. Tests were conducted at two distinct temperatures in which a kinetic transition was occurred, or not: 850 .deg. C and 1100 .deg. C. The effects of flow rate and partial pressure of air was studied by a measurement of mass gain using thermogravimetric analyzer (TGA). After experiments, samples were observed with macrophotography and metallography using optical microscopy. The effect of flow rate and partial pressure of air were qualitatively analyzed with those methods. The effects of flow conditions on the oxidation kinetics of Zircaloy-4 samples were qualitatively studied. The flow rate and the partial pressure of air were changed and their effects was different when the temperature was changed.

Behavior and properties of Zircaloys in power reactors: A short review of pertinent aspects in LWR fuel

International Nuclear Information System (INIS)

Garzarolli, F.; Stehle, H.; Steinberg, E.

1996-01-01

Zircaloy-2 and -4, developed mainly in the US, have been used in Germany for fuel rod claddings and in-core structural components from the beginning of reactor technology. Extensive studies of the material properties of the Zircaloys have been performed in Siemens laboratories since 1957. Zircaloy-2 and -4 turned out to be very reliable materials that fulfilled all requirements for normal operation and likewise the requirements for postulated accidental conditions and for intermediate storage for many years. Optimization of Zircaloy-2 and -4 during recent years includes both optimization of microstructure and of chemical composition. BWRs and PWRs need differently optimized materials. Today's more demanding operation conditions and discharge burnups required a further optimization of the Zircaloys and for hot PWRs even the development of more corrosion-resistant Zr alloys. A significant improvement of PWR corrosion behavior can be achieved with Zr alloys using the alloying elements of Zircaloy with somewhat modified concentrations. Sn should be below or at least in the lower range of the ASTM specification range for Zircaloy-4, Fe and Cr should be somewhat higher, and Si should be specified as an alloying element rather than as an impurity

Influence of surface treatment on the oxidation behavior of zirconium and zircaloy-4

International Nuclear Information System (INIS)

Costa, I.; Ramanathan, L.V.

1986-01-01

The influence of fluoride concentration in surface treatment solutions on the oxidation behavior of Zr and Zircaloy-4 in the temperature range 350-760 0 C have been studied by means of thermogravimetric analysis. Two solutions containing different concentrations of hydrofluoric acid have been used for surface treatments, following which surface roughness measurements were also carried out. The influence of fluoride ion concentration on oxidation behavior has been found to be significant at higher temperatures. (Author) [pt

Hydride precipitation, fracture and plasticity mechanisms in pure zirconium and Zircaloy-4 at temperatures typical for the postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Pshenichnikov, Anton; Stuckert, Juri; Walter, Mario

2016-01-01

Highlights: • All δ-hydrides in Zr and Zircaloy-4 have basal or pyramidal types of habit planes. • Seven orientation relationships for δ-hydrides in Zr matrix were detected. • Decohesion fracture mechanism of hydrogenated Zr was investigated by fractography. - Abstract: The results of investigations of samples of zirconium and its alloy Zircaloy-4, hydrogenated at temperatures 900–1200 K (typical temperatures for loss-of-coolant accidents) are presented. The analyses, based on a range of complementary techniques (X-ray diffraction, scanning electron microscopy, electron backscatter diffraction) reveals the direct interrelation of internal structure transformation and hydride distribution with the degradation of mechanical properties. Formation of small-scale zirconium hydrides and their bulk distribution in zirconium and Zircaloy-4 were investigated. Fractographical analysis was performed on the ruptured samples tested in a tensile machine at room temperature. The already-known hydrogen embrittlement mechanisms based on hydride formation and hydrogen-enhanced decohesion and the applicability of them in the case of zirconium and its alloys is discussed.

Numerical Study on Ultimate Behaviour of Bolted End-Plate Steel Connections

Directory of Open Access Journals (Sweden)

R.E.S. Ismail

Full Text Available Abstract Bolted end-plate steel connections have become more popular due to ease of fabrication. This paper presents a three dimension Finite Element Model (FEM, using the multi-purpose software ABAQUS, to study the effect of different geometrical parameters on the ultimate behavior of the connection. The proposed model takes into account material and geometrical non-linearities, initial imperfection, contact between adjacent surfaces and the pretension force in the bolts. The Finite Element results are calibrated with published experimental results ''briefly reviewed in this paper'' and verified that the numerical model can simulate and analyze the overall and detailed behavior of different types of bolted end-plate steel connections. Using verified FEM, parametric study is then carried out to study the ultimate behavior with variations in: bolt diameter, end-plate thickness, length of column stiffener, angle of rib stiffener. The results are examined with respect to the failure modes, the evolution of the resistance, the initial stiffness, and the rotation capacity. Finally, the ultimate behavior of the bolted end-plate steel connection is discussed in detail, and recommendations for the design purpose are made.

Laser and Pressure Resistance Weld of Thin-Wall Cladding for LWR Accident-Tolerant Fuels

Science.gov (United States)

Gan, J.; Jerred, N.; Perez, E.; Haggard, D. C.

2018-02-01

FeCrAl alloy with typical composition of approximately Fe-15Cr-5Al is considered a primary candidate cladding material for light water reactor accident-tolerant fuel because of its superior resistance to oxidation in high-temperature steam compared with Zircaloy cladding. Thin-walled FeCrAl cladding at 350 μm wall thickness is required, and techniques for joining endplug to cladding need to be developed. Fusion-based laser weld and solid-state joining with pressure resistance weld were investigated in this study. The results of microstructural characterization, mechanical property evaluation by tensile testing, and hydraulic pressure burst testing of the welds for the cladding-endplug specimen are discussed.

Aerosol material release rates from zircaloy-4 at temperatures from 2000 to 22000C

International Nuclear Information System (INIS)

Mulpuru, S.R.; Wren, D.J.; Rondeau, R.K.

1987-01-01

During some postulated severe accidents involving loss of coolant and loss of emergency coolant injection, the temperatures in a CANDU reactor fuel channel become high enough to cause failure and melting of the Zircaloy fuel cladding. At such high temperatures, vapors of fission products and structural (fuel and cladding) materials will be released into the coolant steam and hydrogen mixture. These vapors will condense as cooler conditions are encountered downstream. The vapors from structural materials are relatively involatile; therefore, they will condense readily into aerosol particles. These particles, in turn, will provide sites for the condensation of the more volatile fission products. The aerosol transport of fission products in the primary heat transport system (PHTS) will thus be influenced by the structural material release rates. As part of an ongoing program to develop predictive tools for aerosol and associated fission product transport through the PHTS, experiments have been conducted to measure the vapor mass release rates of the alloying elements from Zircaloy-4 at high temperatures. The paper presents the results and analysis of these experiments

DECONTAMINATION OF ZIRCALOY SPENT FUEL CLADDING HULLS

International Nuclear Information System (INIS)

Rudisill, T; John Mickalonis, J

2006-01-01

The reprocessing of commercial spent nuclear fuel (SNF) generates a Zircaloy cladding hull waste which requires disposal as a high level waste in the geologic repository. The hulls are primarily contaminated with fission products and actinides from the fuel. During fuel irradiation, these contaminants are deposited in a thin layer of zirconium oxide (ZrO 2 ) which forms on the cladding surface at the elevated temperatures present in a nuclear reactor. Therefore, if the hulls are treated to remove the ZrO 2 layer, a majority of the contamination will be removed and the hulls could potentially meet acceptance criteria for disposal as a low level waste (LLW). Discard of the hulls as a LLW would result in significant savings due to the high costs associated with geologic disposal. To assess the feasibility of decontaminating spent fuel cladding hulls, two treatment processes developed for dissolving fuels containing zirconium (Zr) metal or alloys were evaluated. Small-scale dissolution experiments were performed using the ZIRFLEX process which employs a boiling ammonium fluoride (NH 4 F)/ammonium nitrate (NH 4 NO 3 ) solution to dissolve Zr or Zircaloy cladding and a hydrofluoric acid (HF) process developed for complete dissolution of Zr-containing fuels. The feasibility experiments were performed using Zircaloy-4 metal coupons which were electrochemically oxidized to produce a thin ZrO 2 layer on the surface. Once the oxide layer was in place, the ease of removing the layer using methods based on the two processes was evaluated. The ZIRFLEX and HF dissolution processes were both successful in removing a 0.2 mm (thick) oxide layer from Zircaloy-4 coupons. Although the ZIRFLEX process was effective in removing the oxide layer, two potential shortcomings were identified. The formation of ammonium hexafluorozirconate ((NH 4 ) 2 ZrF 6 ) on the metal surface prior to dissolution in the bulk solution could hinder the decontamination process by obstructing the removal of

Experimental investigation of strain, damage and failure of hydrided zircaloy-4 with various hydride orientations

International Nuclear Information System (INIS)

Racine, A; Catherine, C.S.; Cappelaere, C.; Bornert, M.; Caldemaison, D.

2005-01-01

This experimental investigation is devoted to the influence of the orientation of hydrides on the mechanical response of Zircaloy-4. Ring tensile tests are performed on unirradiated CWSR Zircaloy-4, charged with about 200 or 500wppm hydrogen. Hydrides are oriented either parallel ('tangential'), or perpendicular ('radial') to the circumferential tensile direction. Tangential hydrides are usually observed in cladding tubes, however, hydrides can be reoriented after cooling under stress to become radial and then trigger brittle behavior. In this investigation, we perform, 'macroscopic' or SEM in-situ tensile tests on smooth rings, at room temperature. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. The results lead to the following conclusions: neither the tensile stress-strain response nor the strain modes are affected by hydrogen content or hydride orientation, but the failure modes are. Indeed, only 200wppm radial hydrides embrittle Zy-4: sample fails in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases samples reach at least 750 MPa before failure, with ductile or brittle mode. (authors)

Study of mechanical properties from 77 to 900 K and of recovery-recrystallization of zircaloy 4

International Nuclear Information System (INIS)

Derep, J.L.

1981-09-01

Tensile tests carried out on zircaloy-4 between 77 and 900 K show five deformation domains: 1) 77-100 K the deformation is controlled by dislocation-interstitial interaction; 2) 180-205 K alternally activated transition zone with a Orowan process; 3) 250-600 K a thermally activated zone with mobile dislocations through interstitial clusters; 4) 600-700 K two mechanisms are superposed: Orowan or Frank and Read for the migration of dislocations fixed to the Frank network and a Snoek order associated to a dynamic hardening; 5) above 700 K another thermally activated zone. In the study of recovery-recrystallization the mechanisms were determined. Recovery is controlled by polygonisation leading to a cellular structure and annihilation of dislocations of opposite sign producing the growth of the cells. In the crystallisation the germination is controlled by the same mechanism and the growth by interactions between boundaries and second phase precipitates (Zrsub(x)-Fe 5 -Cr 2 ). It is then evident that zircaloy-4 plates used in PWR reactors between 350 and 450 0 C will evolve with time (even without the supplementary effect of irradiation) [fr

The accelerated oxidation of zircaloy-4 at 700∼900 .deg. C in high pressure steam

International Nuclear Information System (INIS)

Kim, K. P.; Park, K. H.

1999-01-01

To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The specimens used in experiments are commercially available Zircaloy-4 used in Kori nuclear power plants. All the measurements were done at 700∼900 .deg. C in steam. Pressure effects were noticed. The oxide thickness was much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. The enhancement of oxide growth rate at 700∼900 .deg. C in high pressure steam is approximately propotion to the power of 1.0∼1.6 of the ratio of experimental steam pressure to critical steam pressure. There is a critical steam pressure above that the oxidation rate enhances. The critical steam pressure was measured as 30∼40 bar. The enhanced oxidation seems from the oxide cracking due to the tetragonal to monoclinic phase transformation at high pressure steam

Interaction between zircaloy tube and inconel spacer grid at high temperature

International Nuclear Information System (INIS)

Nagase, Fumihisa; Otomo, Takashi; Uetsuka, Hiroshi; Furuta, Teruo

1990-09-01

In order to investigate the interaction between fuel cladding and spacer grid of the pressurized water reactor during a severe accident, isothermal reaction tests were performed at the temperature range from 1248 to 1673K. A specimen consisted of a short Zircaloy-4 cladding tube and a piece of spacer grid of Inconel-718. In the tests in an argon atmosphere, eutectic reaction between Zircaloy and Inconel was observed at the contact points at 1248K. Rapid reaction was observed at higher test temperatures. For example, in the test at 1373K for 300s, Zircaloy reacted with Inconel over the entire thickness (0.62mm) of the tube in the vicinity of the contact point. In the present tests, Zircaloy which has higher melting point than Inconel was dissolved preferentially due to eutectic formation. In the tests in an oxygen atmosphere, no eutectic reaction was observed at temperatures below 1437K. A trace of interaction was found at the contact point of specimen heated at 1573 and 1623K. However, decrease in Zircaloy thickness was not measured. The possibility of eutectic reaction between Zircaloy cladding and Inconel spacer grid seems to be quite limited when sufficient oxygen is supplied. (author)

Technology of electron beam welding for Zr-4 alloy spacer grid

International Nuclear Information System (INIS)

Pei Qiusheng; Wu Xueyi; Yang Qishun

1989-10-01

The welding technology for Zr-4 alloy spacer grid by using vacuum electron beam was studied. Through a series of welding technological experiments, metallographic examinations of seam structure and detecting tests for welding defect by X-ray defectoscopy, a good welding technology was selected to meet the requirements. The experimental results indicated that the Zr-4 alloy spacer grid welded by vacuum electron beam welding is feasible

Vertebral endplate (modic) changes and the treatment of back pain using antibiotics

DEFF Research Database (Denmark)

Manniche, Claus

2014-01-01

Vertebral end-plate changes/modic changes are the MRI-image of inflammatory vertebral endplate damage, most often related to general disc degeneration. However, in a subgroup of patients disc infection may be the causal factor. In patients with prolonged back pain, the prevalence of modic changes......'. Most commonly, Propionibacterium acnes is involved. Long-term antibiotics may be effective....

Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

International Nuclear Information System (INIS)

Scheuer, A.; Gutsmiedl, E.

1999-01-01

The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256 deg. C and 250 deg. C. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was take into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256 deg. C and 150 deg. C to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to take into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼ 1x10 22 n/cm 2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture

Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

International Nuclear Information System (INIS)

Gutsmiedl, Erwin

2001-01-01

The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256degC and 250degC. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was taken into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256degC and 150degC to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to taken into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼1·10 22 n/cm 2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture criteria of

Statistics of the acoustic emission signals parameters from Zircaloy-4 fuel cladding

International Nuclear Information System (INIS)

Oliveto, Maria E.; Lopez Pumarega, Maria I.; Ruzzante, Jose E.

2000-01-01

Statistic analysis of acoustic emission signals parameters: amplitude, duration and risetime was carried out. CANDU type Zircaloy-4 fuel claddings were pressurized up to rupture, one set of five normal pieces and six with defects included, acoustic emission was used on-line. Amplitude and duration frequency distributions were fitted with lognormal distribution functions, and risetime with an exponential one. Using analysis of variance, acoustic emission was appropriated to distinguish between defective and non-defective subsets. Clusters analysis applied on mean values of acoustic emission signal parameters were not effective to distinguish two sets of fuel claddings studied. (author)

Stress corrosion crack growth in unirradiated zircaloy

International Nuclear Information System (INIS)

Pettersson, K.

1978-10-01

Experimental techniques suitable for the determination of stress corrosion crack growth rates in irradiated Zircaloy tube have been developed. The techniques have been tested on unirradiated. Zircaloy and it was found that the results were in good agreement with the results of other investigations. Some of the results were obtained at very low stress intensities and the crack growth rates observed, gave no indication of the existance of a K sub(ISCC) for iodine induced stress corrosion cracking in Zircaloy. This is of importance both for fuel rod behavior after a power ramp and for long term storage of spent Zircaloy-clad fuel. (author)

Behaviour of MZFR-type Zircaloy-4 cans under tensile stress

International Nuclear Information System (INIS)

Bordoni, R.A.; Casario, J.A.; Coroli, Graciela; Povolo, Francisco

1981-01-01

The paper describes the experimental procedure and results from the tensile tests of Zircaloy-4 fuel cans of the MZFR-type, performed at temperatures ranging from 250 to 450 deg C and for a relative deformation velocity of about 0.5%/min. In the representation of the results by a curve of the type sigma = K epsilon/sup n/, two different stages are observed. By statistically fitting the experimental curves, the values for the K and n parameters were obtained for each stage as a function of temperature. The results are discussed and compared with similar data found in current literature. It is concluded that new tests on tubes of different characteristics are necessary in order to obtain a clearer idea about the mechanical behaviour of these materials. (C.A.K.) [es

Delayed hydride cracking of Zircaloy-4 fuel cladding

International Nuclear Information System (INIS)

Pizarro, Luis M.; Fernandez, Silvia; Lafont, Claudio; Mizrahi, Rafael; Haddad, Roberto

2007-01-01

Crack propagation rates, grown by the delayed hydride cracking mechanism, were measured in Zircaloy-4 fuel cladding, according to a Coordinated Research Project (CRP) sponsored by the International Atomic Energy Agency (IAEA). During the first stage of the program a Round Robin Testing was performed on fuel cladding samples provided by Studsvik (Sweden), of the type used in PWR reactors. Crack growth in the axial direction is obtained through the specially developed 'pin load testing' (PLT) device. In these tests, crack propagation rates were determined at 250 C degrees on several samples of the material described above, obtaining a mean value of about 2.5 x 10 -8 m s -1 . The results were analyzed and compared satisfactorily with those obtained by the other laboratories participating in the CRP. At the present moment, similar tests on CANDU and Atucha I type fuel cladding are being performed. It is thought that the obtained results will give valuable information concerning the analysis of possible failures affecting fuel cladding under reactor operation. (author) [es

Chemical and microstructural characterization of recycled zircaloy

International Nuclear Information System (INIS)

Martinez, Luis G.; Pereira, Luiz A.T.; Rossi, Jesualdo L.; Takiishi, Hidetoshi; Sato, Ivone M.; Scapin, Marcos A.; Orlando, Marcos T.D.

2011-01-01

PWR reactors employ as nuclear fuel UO 2 pellets with Zircaloy clad. Brazil is autonomous in the nuclear fuel cycle, from uranium mining to enrichment and nuclear fuel manufacture. However, the industrial production of nuclear zirconium alloys does not meet the demand, leading to importation of Zircaloy for fuel manufacturing. In the fabrication of fuel elements parts, machining chips of alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is strategic in economical and environmental aspects. In this work are described two methods that are being developed to recycle Zircaloy chips. The first method the Zircaloy machining chips are melted using an electric arc furnace to obtain small laboratory ingots. The second method uses powder metallurgy technique. By this later method, the Zircaloy chips are submitted to a hydriding process and the resulting material is milled in a high-energy ball mill. The powder is cold isostatically pressed and vacuum sintered. The elemental composition of the materials obtained using both methods is being determined using X-ray fluorescence techniques and compared to the specifications of nuclear grade Zircaloy and to the composition of the starting chips. The phase composition of the laboratory ingots was determined using X-ray diffraction. The ingots were vacuum annealed and the microstructures resulting from both processing methods before and after heat treatments were characterized using optical and scanning electron microscopy. The hardness of the materials was evaluated. A methodology of chemical analysis using X-ray fluorescence spectrometry, for composition certification, was established and tested. The results showed that recycled Zircaloy presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding cap-ends, using near net shape sintering. (author)

Biaxial creep deformation of Zircaloy-4 in the high alpha phase temperature range

International Nuclear Information System (INIS)

Donaldson, A.T.; Horwood, R.A.; Healey, T.

1983-01-01

The ballooning response of Zircaloy-4 fuel tubes during a postulated loss-of-coolant accident may be calculated from a knowledge of the thermal environment of the rods and the creep deformation characteristics of the cladding. In support of such calculations biaxial creep studies have been performed on fuel tubes supplied by Westinghouse, Wolverine and Sandvik of temperatures in the alpha phase range. This paper presents the results of an investigation of their respective creep behaviour which has resulted in the formulation of equations for use in LOCA fuel ballooning codes. (author)

The effect of repeated melting of zircaloy-4 to the distribution of volatile constituents

International Nuclear Information System (INIS)

Johneri, E.; Wijaksana; Badruzzaman, M.

1996-01-01

The effect of repeated fusion on the composition and distribution of zircaloy volatile elemental constituents (especially Sn) has been investigated. The results showed that the higher the number of repeated fusion is, the more evenly distributed the constituents are, but the composition decreased until reached constant values. This phenomenon occurred due to the relatively faster diffusion movement of one element compared to the others. Further investigation needs to be done to find other proofs of the phenomenon. Moreover, continued research is in demand in order to answer technological problems regarding the zircaloy production and metal alloy production in general. (author)

An investigation of deformed microstructure and mechanical properties of Zircaloy-4 processed through multiaxial forging

Energy Technology Data Exchange (ETDEWEB)

Fuloria, Devasri; Nageswararao, P. [Department of Metallurgical and Materials Engineering & Centre of Nanotechnology, IIT Roorkee, Roorkee 247667 (India); Jayaganthan, R., E-mail: rjayafmt@iitr.ernet.in [Department of Metallurgical and Materials Engineering & Centre of Nanotechnology, IIT Roorkee, Roorkee 247667 (India); Department of Engineering Design, Indian Institute of Technology Madras, Chennai 600036 (India); Jha, S. [Nuclear Fuel Complex Limited, Hyderabad 501301 (India); Srivastava, D. [Materials Science Division, Bhabha Atomic Research Centre, Mumbai 40085 (India)

2016-04-15

In the present work, the mechanical behavior of Zircaloy-4 subjected to various deformation strains by multiaxial forging (MAF) at cryogenic temperature (CT) was investigated. The alloy was strained up to different number of cycles, viz., 6 cycles, 9 cycles, and 12 cycles at cumulative strains of 2.96, 4.44, and 5.91, respectively. The mechanical properties of the alloy were investigated by performing the universal tensile test and the Vickers hardness test. Both the test showed improvement in the ultimate tensile strength and hardness value by 51% and 26%, respectively, at the highest cumulative strain of 5.91. The electron backscattered diffraction (EBSD) measurement and transmission electron microscopy (TEM) were used for analyzing the deformed microstructure. The microstructures of the alloy underwent deformation at various cumulative strains/cycles showed grain refinement with the evolution of shear and twin bands that were highest for the alloy deformed at the highest number of cycles. The effective grain refinement was due to twins formation and their intersection, which led to the improvement in mechanical properties of the MAFed alloy, as observed in the present work. - Highlights: • Zircaloy-4 was subjected to MAF at cryogenic temperature. • Microstructural evolution was studied through EBSD and TEM. • Deformed microstructure was marked with various types of twinning and shear banding. • Twins formations are responsible for effective grain refinement and enhanced mechanical properties.

The effects of corrosion conditions and cold work on the nodular corrosion of zircaloy-4

International Nuclear Information System (INIS)

You, Gil Sung

1992-02-01

The nodular corrosion of Zircaloy-4 was investigated on the effects of corrosion conditions and cold work. Variation of steam pressures, heat-up environments and prefilms were considered and cold work effects were also studied. The corrosion rate of Zircaloy-4 was dependent on pressure between 1 and 100 atm and it followed the cubic law as W=16.85 x P 0.31 for plate specimens and W=12.69 x P 0.27 for tube specimens, where W is weight gain (mg/dm 2 ) and P is the steam pressure (atm). The environment variation in autoclave during heat-up period did not affect the early stage of nodular corrosion. The prefilm, which was formed at 500 .deg. C under 1 atm steam for 4 hours, restrained the formation of the initial small nodules. The oxide film formed under 1 atm steam showed no difference of electrical resistivity from the oxides formed under 100 atm steam pressure. Cold work specimens showed the higher resistivity against nodular corrosion than as-received specimens. The corrosion resistance arising from cold work seems to be due to the texture changes by the cold work. The results showed that cold work can affect the later stage of uniform corrosion and the early stage of nodular corrosion, namely, the nodule initiation stage

Effect of hydrogen and hydrides on the viscoplastic behaviour of the recrystallized zircaloy-4; Effet de l'hydrogene et des hydrures sur le comportement viscoplastique du zircaloy-4 recristallise

Energy Technology Data Exchange (ETDEWEB)

Rupa, N

2000-04-15

Zircaloy-4 is the main material of PWR fuel assemblies. In service as during the storage, the integrity of these compounds has to be guaranteed in spite of the presence of hydrogen (in solution in the zirconium matrix) and of hydrides (which precipitate when the amount of hydrogen is higher than the solubility limit). The aim of this work is to characterize the hydrogen and hydrides effect on the viscoplastic behaviour of the non irradiated recrystallized zircaloy-4. The presence of hydrogen in solid solution induces a decrease of the mechanical properties: the creep kinetics are then increased and the tensile stresses decreased. This decrease is particularly visible in conditions of oxygen/dislocations dynamic interactions (revealed on the material without hydrogen). The advanced hypothesis, strengthened by the atomic simulation results, is that the hydrogen facilitates the dislocations movement, in diminishing the effects of anchoring by the interstitials, and/or in increasing the intrinsic mobility of dislocations. The hydrides effect induces a hardening of the material (decrease of the creep kinetics, increase of the tensile stresses and of the relaxed stresses) compensating the decrease by hydrogen. The hardening mechanism is due to an increase of the internal constraints, determined by load-unload tests. For the very weak plastic deformations, the hydrides are an obstacle to the dislocations gliding. They are then passed (that corresponds to a saturation of the internal constraint). The TEM observations as well as the results obtained on the titanium indicate that the precipitates are then submitted to a deformation mechanism. (O.M.)

Effect of Aquo-glycolic Media and Added Anions on the Anodization of Zircaloy-4 in Sulphamic Acid

Directory of Open Access Journals (Sweden)

Viplav Duth Shukla

2011-01-01

Full Text Available Anodization of zircaloy-4 in 0.1 M sulphamic acid has been carried out. Kinetics of anodic oxidation of zircaloy-4 has been studied at a constant current density of 8 mA/cm2 and at room temperature. Thickness estimates were made from capacitance data. The plots of formation voltage vs. time, reciprocal capacitance vs. time, reciprocal capacitance vs. formation voltage and thickness vs. formation voltage were drawn and rate of formation, current efficiency and differential field were calculated. The addition of solvent (ethylene glycol showed better kinetic results. For 25%, 50% and 75% aquo-glycolic media, the dielectric constant values are low leading to a marked improvement in the kinetics. In 80% ethylene glycol, though the dielectric constant value of solution is less, the kinetics was slow which may be attributed to the fact that the electrolyte becomes highly non-polar. Improvement in the kinetics of oxide film formation was observed by the addition of millimolar concentration of anions (CO32-, SO42-, PO43-. The presence of phosphate ions improved the kinetics of anodization to better extent.

Corrosion behavior in high-temperature pressurized water of Zircaloy-4 joints brazed with Zr-Cu-based amorphous filler alloys

Science.gov (United States)

Lee, Jung Gu; Lee, Gyoung-Ja; Park, Jin-Ju; Lee, Min-Ku

2017-05-01

The compositional effects of ternary Zr-Cu-X (X: Al, Fe) amorphous filler alloys on galvanic corrosion susceptibility in high-temperature pressurized water were investigated for Zircaloy-4 brazed joints. Through an Al-induced microgalvanic reaction that deteriorated the overall nobility of the joint, application of the Zr-Cu-Al filler alloy caused galvanic coupling to develop readily between the Al-bearing joint and the Al-free base metal, finally leading to massive localized corrosion of the joint. Contrastingly, joints prepared with a Zr-Cu-Fe filler alloy showed excellent corrosion resistance comparable to that of the Zircaloy-4 base metal, since the Cu and Fe elements forming fine intermetallic particles with Zr did not influence the electrochemical stability of the resultant joints. The present results demonstrate that Fe is a more suitable alloying element than Al for brazing filler alloys subjected to high-temperature corrosive environments.

Investigation of microstructure and mechanical properties of proton irradiated Zircaloy 2

Energy Technology Data Exchange (ETDEWEB)

Sarkar, Apu, E-mail: asarkar@barc.gov.in [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Kumar, Ajay [Nuclear Physics Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India); Mukherjee, S.; Sharma, S.K.; Dutta, D.; Pujari, P.K. [Radiochemistry Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Agarwal, A.; Gupta, S.K.; Singh, P. [Ion Accelerator Development Division, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Reserch Centre, Mumbai, 400 085 (India)

2016-10-15

Samples of Zircaloy 2 have been irradiated with 4 MeV protons to two different doses. Microstructures of the unirradiated and irradiated samples have been characterized by Electron Back Scatter Diffraction (EBSD), X-ray diffraction line profile analysis (XRDLPA), Positron Annihilation Lifetime Spectroscopy (PALS) and Coincident Doppler Broadening (CDB) Spectroscopy. Tensile tests and micro hardness measurements have been carried out at room temperature to assess the changes in mechanical properties of Zircaloy 2 due to proton irradiation. The correlation of dislocation density, grain size and yield stress of the irradiated samples indicated that an increase in dislocation density due to irradiation is responsible for the change in mechanical behavior of irradiated Zircaloy.

An investigation of the microstructures of heat-treated zircaloy-4

International Nuclear Information System (INIS)

Bangaru, N.V.

1985-01-01

A TEM/STEM investigation of the microstructure and microchemistry of commercial Zircaloy-4 samples subjected to three different final heat treatments in the laboratory has been conducted to understand the processing-microstructure-corrosion relationships in these alloys. Pronounced differences in the volume fraction, morphology, and chemistry of the intermetallic particles as well as in the α phase microstructure have been observed among the beta-quenched, as-received (stress-relieved) and alpha-annealed samples. The beta-quenched sample exhibits the most uniform microstructure consisting of acicular α phase with lath boundary Sn enrichment and fine intermetallic particle formation. The as-received sample has the most inhomogeneous microstructure made up of annealed and deformed α phase. The relevance of the observed microstructural features to the nodular corrosion susceptibility is discussed in the light of some existing models of modular corrosion. (orig.)

Finite element analysis of an extended end-plate connection using the T-stub approach

Energy Technology Data Exchange (ETDEWEB)

Muresan, Ioana Cristina; Balc, Roxana [Technical University of Cluj-Napoca, Faculty of Civil Engineering. 15 C Daicoviciu Str., 400020, Cluj-Napoca (Romania)

2015-03-10

Beam-to-column end-plate bolted connections are usually used as moment-resistant connections in steel framed structures. For this joint type, the deformability is governed by the deformation capacity of the column flange and end-plate under tension and elongation of the bolts. All these elements around the beam tension flange form the tension region of the joint, which can be modeled by means of equivalent T-stubs. In this paper a beam-to-column end-plate bolted connection is substituted with a T-stub of appropriate effective length and it is analyzed using the commercially available finite element software ABAQUS. The performance of the model is validated by comparing the behavior of the T-stub from the numerical simulation with the behavior of the connection as a whole. The moment-rotation curve of the T-stub obtained from the numerical simulation is compared with the behavior of the whole extended end-plate connection, obtained by numerical simulation, experimental tests and analytical approach.

Flow stress and dynamic strain-ageing of β-transformed Zircaloy-4

International Nuclear Information System (INIS)

Woo, O.T.; Tseng, D.; Tangri, K.; MacEwen, S.R.

1979-01-01

The 0.2% yield stress of β-transformed Zircaloy-4 was found to be independent of prior-β grain size but varied as the inverse of the transformed β plate width. A dislocation loop expansion model originally proposed by Langford and Cohen (1969) for cold-drawn iron wires is used to explain the inverse plate width dependence. Both air-cooled and water-quenched samples exhibited dynamic strain-ageing effects in approximately the same temperature range of 573 to 673 K: (a) a local minimum in strain-rate sensitivity is associated with a peak or an inflection point in the temperature dependence of the 0.2% yield stress for water-quenched or air-cooled samples respectively, and (b) yield drops were observed in strain rate change tests. (Auth.)

Corrosion Characteristics and Kinetics of Zircaloys and Aluminium Alloys

International Nuclear Information System (INIS)

Sugondo; Chaidir, A

1998-01-01

Corrosion rate characterization of cladding materials has been done by dynamic method. The materials are zircaloy-2,zircaloy-4,AIMg2,and AIMgSi.The zircaloy alloys are characterized in the electrolytes of boric ion,iodide ion,lithium ion and cesium ion with a pH variation.The aluminum alloys are characterized in the cooling water of RSG-GAS reactor in different temperatures and Ph values .The results, show that corrosion product of iodine on zircaloy is not passivated, meanwhile the corrosion product of cesium undergoes passivation. However, the deposited substance in the surface of the specimens as indicated using WDX-SEM shows the same deposition rate.it is concluded therefore that iodine is diffused into the materials without getting resistance from the deposited substances on the surface. The effect of pH to corrosion rate of iodine on the zircaloy fluctuates meanwhile the cesium has the minimum corrosion rate at pH 7.5 At the concentration of 0.1 gram/1,cesium ion is more reactive than iodine but at higher concentration the reactivity becomes competitive . Furthermore , the interaction between zircaloy and boric ion at concentration of 300 ppm and lithium ion at 10 ppm shows an outstanding corrosion rate, i.e. 0.1 mpy. if both substances are mixed then the corrosion rate decreases drastically in the order of 10 -2 mpy.The reason of such a decrease may be due to the formation of complexes of boron lithium on the electrode surface. The arrhenius activation energies for such reaction have been found to be 37629.322 joule/mole 0 K for Al Mg 2 and 41609.822 joule /mole 0 K for AIMgSi ,respectively. This underlies the argument that AI Mg 2 is more reactive than AI Mg Si besides , AI Mg 2 is more reactive under acid condition meanwhile AI Mg Si more reactive under basic condition. Both alloys over come the minimum corrosion rate at the pH in between 4.7 to 7.5 and the level of the corrosion rate in the pH interval was outstanding

Autoclave corrosion of zircaloy-4 cladding samples in LiOH solutions

International Nuclear Information System (INIS)

Hermann, A.

2010-03-01

In reactor operation, pH of the cooling water is adjusted by addition of alkaline hydroxides, and LiOH has been found to be the most suitable one. The addition of LiOH above a certain concentration level (depending on temperature) increases the corrosion rate of zirconium and its alloys. Hydrogen pick-up by the metal is also increased, and this effect is used to produce hydrided specimens for different investigations using the corrosion reaction. At the Paul Scherrer Institute several projects were accomplished to investigate the influence of hydrogen in Zircaloy cladding on its mechanical properties. In order to produce hydrided specimens for comparison and for adjusting new equipment, Zircaloy tubing samples were hydrogen charged by autoclave corrosion in lithiated water. Results of the corrosion experiments are outlined in this publication. Because of the great variety of possible experimental parameters these results are still of interest for the scientific community. Autoclave corrosion was accomplished in 0.2 M or 0.5 M LiOH solution at a constant temperature of 340 o C and a pressure of 160 bar. The corrosion rate increases from 84 mg/(dm 2 d) in 0.2 M LiOH to 153 mg/(dm 2 d) in 0.5 M LiOH. The hydrogen pick-up fraction in 0.5 M LiOH amounts to 80%. In 0.5 M LiOH, Zircaloy tubing samples can be charged with ∼ 500 ppm hydrogen in about 40 hours. In the corrosion experiments described in this report a homogeneous distribution of hydrides should be expected (except very high hydride concentrations) because no temperature gradient exists through the tubing wall. Hydrogen stringers are homogeneously distributed with circumferential orientation (stress-relieved tubing samples). (author)

Some tests of explosion welding in Zry 4 - Zry 4 and other materials

International Nuclear Information System (INIS)

Badino, N.S.; Nashmias, J.L.

1993-01-01

The possibility of welding explosion for nuclear industry materials is studied. In this case, difficulties to join them by fusion welding are presented. The following pairs are selected: Zry 4-Zry 4, Zry 4-Incoloy, Zry 4-stainless steel, Zry 4-Inconel. Good results in Zry 4-Zry 4 and not so clear results are obtained in other cases. Other possible tests are now being studied and thought about. (Author)

A new method of residual stress distribution analysis for corroded Zircaloy-4 cladding

International Nuclear Information System (INIS)

Godlewski, J.; Cadalbert, R.

1992-01-01

An X-ray diffraction method of residual stress measurement is developed to determine the stress level in the metal near the metal/oxide interface of Zircaloy-4 cladding samples oxidized in steam water at 400degC under a pressure of 10.3 MPa. The stress gradient is obtained and the evolution of the average stress is determined as function of the oxidation time. The presence of tetragonal zirconia phase in quite large quantity near the metal/oxide interface could be correlated to the high stress level in the base metal, adjacent to the interface. (author)

A new method for residual stress distribution - analysis of corroded zircaloy-4 cladding

International Nuclear Information System (INIS)

Godlewski, J.; Cadalbert, R.

1992-01-01

An X-ray diffraction method for residual stress measurement is developed to determine the stress level in the metal near the metal/oxide interface of Zircaloy-4 cladding samples oxidized in steam water at 400 deg C under a pressure of 10.3 MPa. The stress gradient is obtained and the evolution of the average stress is determined as a function of the oxidation time. The presence of tetragonal zirconia phase in quite large quantity near the metal/oxide interface could be correlated to the high stress level in the base metal, adjacent to the interface. 12 refs., 5 figs., 1 tab

Corrosion performance of new Zircaloy-2-based alloys

International Nuclear Information System (INIS)

Rudling, P.; Mikes-Lindbaeck, M.; Lethinen, B.; Andren, H.O.; Stiller, K.

1994-01-01

A material development project was initiated to develop a new zirconium alloy, outside the ASTM specifications for Zircaloy-2 and Zircaloy-4, with optimized hydriding and corrosion properties for both boiling water reactors and pressurized water reactors. A number of different alloys were manufactured. These alloys were long-term corrosion tested in autoclaves at 400 C in steam. Also, a 520 C/24 h steam test was carried out. The zirconium metal microstructure and the chemistry of precipitates were characterized by analytical electron microscopy. The metal matrix chemistry was determined by atom probe analysis. The paper describes the correlations between corrosion material performance and zirconium alloy microstructure

Fabrication and use of zircaloy/tantalum-sheathed cladding thermocouples and molybdenum/rhenium-sheathed fuel centerline thermocouples

International Nuclear Information System (INIS)

Wilkins, S.C.; Sepold, L.K.

1985-01-01

The thermocouples described in this report are zircaloy/tantalum-sheathed and molybdenum/rhenium alloy-sheathed instruments intended for fuel rod cladding and fuel centerline temperature measurements, respectively. Both types incorporate beryllium oxide insulation and tungsten/rhenium alloy thermoelements. These thermocouples, operated at temperatures of 2000 0 C and above, were developed for use in the internationally sponsored Severe Fuel Damage test series in the Power Burst Facility. The fabrication steps for both thermocouple types are described in detail. A laser-welding attachment technique for the cladding-type thermocouple is presented, and experience with alternate materials for cladding and fuel therocouples is discussed

Longitudinal study of vertebral type-1 end-plate changes on MR of the lumbar spine

Energy Technology Data Exchange (ETDEWEB)

Mitra, D. [Department of Neuroradiology, Newcastle General Hospital, Westgate Road, NE4 6BE, Newcastle upon Tyne (United Kingdom); Cassar-Pullicino, V.N.; Mccall, I.W. [Department of Radiology, Robert Jones and Agnes Hunt Orthopaedic Hospital, Oswestry (United Kingdom)

2004-09-01

The purpose of this study was to investigate the temporal evolution of type-1 end-plate changes on MRI in patients with degenerative disease of the lumbar spine and to evaluate whether any correlation exists between such evolution and the change in patients' symptoms. Forty-four patients with 48 Modic type-1 end-plate changes (low TI signal and high T2 signal) were studied. All patients had an initial and a follow-up non-contrast lumbar MRI with variable intervals between the studies (12-72 months). Severity of the end-plate changes was assessed by eyeball estimation. Correlation with patients' symptoms was studied with the help of the Visual Analogue Score (VAS), Oswestry Questionnaire Score (OQS) and patients' subjective assessment. Of the 48 disc levels with type-1 changes, 18 (37.5%) converted fully to type 2 (high T1 signal and intermediate to high T2 signal), 7 (14.6%) partially converted to type 2, 19 (39.6%) became worse (i.e. type 1 changes became more extensive) and 4 (8.3%) showed no change. Higher average VAS (5.7) and OQS (42.3) scores were noted in patients where there was worsening type-1 change and lower scores (3.8 and 27, respectively) were seen in those where there was conversion to type-2 change. These trends, however, did not reach statistical significance (P values 0.16 and 0.09 for VAS and OQS, respectively). The statistical relationship was stronger after exclusion of patients with confounding factors (i.e. changes in lumbar MRI other than end-plate changes that could independently explain the evolution of patients' symptoms) with P-values of 0.08 and 0.07 for VAS and OQS, respectively. Type-1 end-plate change represents a dynamic process and in a large majority of cases either converts to type-2 change or becomes more extensive. The evolution of type-1 change relates to change in patient's symptoms, but not to a statistically significant level. (orig.)

Longitudinal study of vertebral type-1 end-plate changes on MR of the lumbar spine

International Nuclear Information System (INIS)

Mitra, D.; Cassar-Pullicino, V.N.; Mccall, I.W.

2004-01-01

The purpose of this study was to investigate the temporal evolution of type-1 end-plate changes on MRI in patients with degenerative disease of the lumbar spine and to evaluate whether any correlation exists between such evolution and the change in patients' symptoms. Forty-four patients with 48 Modic type-1 end-plate changes (low TI signal and high T2 signal) were studied. All patients had an initial and a follow-up non-contrast lumbar MRI with variable intervals between the studies (12-72 months). Severity of the end-plate changes was assessed by eyeball estimation. Correlation with patients' symptoms was studied with the help of the Visual Analogue Score (VAS), Oswestry Questionnaire Score (OQS) and patients' subjective assessment. Of the 48 disc levels with type-1 changes, 18 (37.5%) converted fully to type 2 (high T1 signal and intermediate to high T2 signal), 7 (14.6%) partially converted to type 2, 19 (39.6%) became worse (i.e. type 1 changes became more extensive) and 4 (8.3%) showed no change. Higher average VAS (5.7) and OQS (42.3) scores were noted in patients where there was worsening type-1 change and lower scores (3.8 and 27, respectively) were seen in those where there was conversion to type-2 change. These trends, however, did not reach statistical significance (P values 0.16 and 0.09 for VAS and OQS, respectively). The statistical relationship was stronger after exclusion of patients with confounding factors (i.e. changes in lumbar MRI other than end-plate changes that could independently explain the evolution of patients' symptoms) with P-values of 0.08 and 0.07 for VAS and OQS, respectively. Type-1 end-plate change represents a dynamic process and in a large majority of cases either converts to type-2 change or becomes more extensive. The evolution of type-1 change relates to change in patient's symptoms, but not to a statistically significant level. (orig.)

Microstructural aspects of zircaloy nodular corrosion in steam

International Nuclear Information System (INIS)

Taylor, D.F.

1999-01-01

Zircaloy-2 becomes susceptible to nodular corrosion in high-temperature, high-pressure steam when the total solute concentration of the β-stabilizing alloying elements Fe, Ni and Cr in the α-zirconium matrix falls below a critical value C c that is characteristic of the test conditions. C c for typical commercial Zircaloy-2 in a 24hr/510 C/10.4MPa steam-test is the precipitate-free a-matrix concentration in equilibrium with solute-saturated β phase at about 840 C, the corresponding critical temperature T c .Thus, immunity to nodular corrosion is a metastable condition for α-Zircaloy that requires fast cooling from above T c to achieve adequate solute concentration throughout the matrix. Annealing Zircaloy at any temperature below T c for a sufficiently long time makes it susceptible to nodular corrosion. In the (α+χ) phase field, where χ collectively designates the Fe-, Cr-, and Ni-containing precipitate phases, lowering the solute concentration to less than C c by Ostwald ripening can require many hundreds of hours. Above about 825 C, the temperature of the (α+χ)/(α+β+χ) transus, solute-saturated β phase surrounds each precipitate and a strong inverse activity gradient promotes equilibration with the much lower solute concentration in the α matrix. Sensitization to nodular corrosion occurs most rapidly at about 835 C between the (α+χ)/(α+β+χ) transus and T c . Annealing Zircaloy at temperatures above T c for a sufficiently long time will raise the solute concentration above C c and, with rapid cooling, heal any degree of susceptibility. Annealing within the protective coarsening window between T c and about 850 C, the temperature of the (α+β+χ)/(α+β) transus, achieves rapid precipitate growth in a matrix immune to nodular corrosion

Influence of cortical endplates on ultrasonic properties of trabecular bone

International Nuclear Information System (INIS)

Kim, Yoon Mi; Lee, Kang Il

2015-01-01

The present study investigated the influence of thick cortical endplates on the ultrasonic properties of trabecular bone in a femur with a high fracture risk. Twelve trabecular bone samples were prepared from bovine femurs, and acrylic plates with thicknesses of 1.25, 1.80, and 2.75 mm were manufactured to simulate the cortical endplates using acrylic with a density and a sound speed similar to cortical bone. Although the thickness of the acrylic plates attached to the two sides of the trabecular bone increased, high correlations were observed between the speed of sound and the apparent bone density of the trabecular bone, with Pearson's correlation coefficients of 0.80-0.86. High correlations were also observed between the attenuation coefficient at 0.5 mm and the apparent bone density of the trabecular bone, with Pearson's correlation coefficients of 0.84-0.91. These results suggest that the speed of sound and attenuation coefficient at a specific frequency measured in a femur with relatively thick cortical endplates compared to the calcaneus could be used as indices for predicting the bone mineral density of the femur.

Mitigation of harmful effects of welds in zirconium alloy components

International Nuclear Information System (INIS)

Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.; Webster, R.T.

1995-01-01

Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled Zircaloy-2 tubes containing an axial weld do not reach their full strength because they always fail prematurely in the weld when pressurized to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal and improves the biaxial strength of the tube by 20 to 25% and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 o C and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat treatment to the point where the probability of cracking is very low. (author)

Mitigation of harmful effects of welds in zirconium alloy components

International Nuclear Information System (INIS)

Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.; Webster, R.T.

1993-10-01

Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled zircaloy-2 tubes containing an axial weld do not reach their full strength, because they always fail prematurely in the weld when pressurised to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal, improves the biaxial strength of the tube by 20 to 25%, and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 degrees celsius and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat-treatment to the point where the probability of cracking is very low

Mitigation of harmful effects of welds in zirconium alloy components

International Nuclear Information System (INIS)

Coleman, C.E.; Doubt, G.L.; Fong, R.W.L.; Root, J.H.; Bowden, J.W.; Sagat, S.

1994-01-01

Welding produces local residual tensile stresses and changes in texture in components made from zirconium alloys. In the heat-affected zone in tubes or plates, the basal plane normals are rotated into the plane of the component and perpendicular to the direction of the weld. Thin-walled Zircaloy-2 tubes containing an axial weld do not reach their full strength because they always fail prematurely in the weld when pressurized to failure in a fixed-end burst test. Reinforcing the weld by increasing its thickness by 25% moves the failure to the parent metal and improves the biaxial strength of the tube by 20 to 25% and increases the total elongation by 200 to 450%. In components made from Zr-2.5Nb, the texture in the heat-affected zone promotes delayed hydride cracking (DHC) driven by tensile residual stress. Although the texture is not much affected by heat-treatments below 630 C and large grain interaction stresses remain as a result of mixed textures, macro-residual tensile stresses can be relieved by heat treatment to the point where the probability of cracking is very low

Hydrides blister formation and induced embrittlement on zircaloy-4 cladding tubes in reactivity initiated conditions

International Nuclear Information System (INIS)

Hellouin-De-Menibus, A.

2012-01-01

Our aim is to study the cladding fracture with mechanical tests more representative of RIA conditions, taking into account the hydrides blisters, representative strain rates and stress states. To obtain hydride blisters, we developed a thermodiffusion setup that reproduces blister growth in reactor conditions. By metallography, nano-hardness, XRD and ERDA, we showed that they are constituted by 80% to 100% of δ hydrides in a Zircaloy-4 matrix, and that the zirconium beneath has some radially oriented hydrides. We modeled the blister growth kinetics taking into account the hysteresis of the hydrogen solubility limit and defined the thermal gradient threshold for blister growth. The modeling of the dilatometric behavior of hydrided zirconium indicates the important role of the material crystallographic texture, which could explain differences in the blister shape. Mechanical tests monitored with an infrared camera showed that significant local heating occurred at strain rates higher than 0.1/s. In parallel, the Expansion Due to Compression test was optimized to increase the bi-axiality level from uniaxial stress to plane strain (HB-EDC and VHB-EDC tests). This increase in loading bi-axiality lowers greatly the fracture strain at 25 C and 350 C only in homogeneous material without blister. Eventually, the ductility decrease of unirradiated Zircaloy-4 cladding tube in function of the blister depth was quantified. (author) [fr

Mechanical properties of zircaloy-4 tubes for CAREM 25 fuel rods

International Nuclear Information System (INIS)

Juarez, G; Bianchi, D; Flores, A; Vizcaino, P

2012-01-01

The aim of the present work was giving support to the development of Zircaloy-4 fuel claddings for the CAREM 25 reactor through microstructural and mechanical properties studies along the manufacturing process. The manufacturing route was defined in 4 cold rolling stages and two thermal treatments, one at the middle and one after the last rolling stage. The first two rolling stages were performed in FAESA and the last two in PPFAE-CNEA using the rolling machine HPTR 8-15. The reference values for the evaluation were those indicated in the technical specification CAREM25 F ET-3-B0610. In this context, four tubes were received from FAESA. To these tubes mechanical properties determinations were performed to characterize the material in each step performed in PPFAE. The mechanical properties of the cladding tubes also achieve the standard values (σ 0.2 = 450 MPa, e = 15%), being σ 0.2 = 530 MPa and 18% the elongation (author)

Plating on Zircaloy-2

International Nuclear Information System (INIS)

Dini, J.W.; Johnson, H.R.; Jones, A.

1979-03-01

Zircaloy-2 is a difficult alloy to coat with an adherent electroplate because it easily forms a tenacious oxide film in air and aqueous solutions. Procedures reported in the literature and those developed at SLL for surmounting this problem were investigated. The best results were obtained when specimens were first etched in either an ammonium bifluoride/sulfuric acid or an ammonium bifluoride solution, plated, and then heated at 700 0 C for 1 hour in a constrained condition. Machining threads in the Zircaloy-2 for the purpose of providing sites for mechanical interlocking of the plating also proved satisfactory

A comparative study of pulsed Nd:YAG laser welding and TIG welding of thin Ti6Al4V titanium alloy plate

International Nuclear Information System (INIS)

Gao, Xiao-Long; Zhang, Lin-Jie; Liu, Jing; Zhang, Jian-Xun

2013-01-01

This paper reports on a study aiming at comparing properties of the Ti6Al4V titanium alloy joints between pulsed Nd:YAG laser welding and traditional fusion welding. To achieve the research purpose, Ti6Al4V titanium alloy plates with a thickness of 0.8 mm were welded using pulsed Nd:YAG laser beam welding (LBW) and gas tungsten arc welding (TIG), respectively. Residual distortions, weld geometry, microstructure and mechanical properties of the joints produced with LBW and TIG welding were compared. During the tensile test, with the aid of a high speed infrared camera, evolution of the plastic strain within tensile specimens corresponding to LBW and TIG welding were recorded and analyzed. Compared with the TIG, the welded joint by LBW has the characters of small overall residual distortion, fine microstructure, narrow heat-affected zone (HAZ), high Vickers hardness. LBW welding method can produce joints with higher strength and ductility. It can be concluded that Pulsed Nd:YAG laser welding is much more suitable for welding the thin Ti6Al4V titanium alloy plate than TIG welding.

Interactions of zircaloy cladding with gallium -- 1997 status

International Nuclear Information System (INIS)

Wilson, D.F.; DiStefano, J.R.; King, J.F.; Manneschmidt, E.T.; Strizak, J.P.

1997-11-01

A four phase program has been implemented to evaluate the effect of gallium in mixed oxide (MOX) fuel derived from weapons grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in LWR. This graded, four phase experimental program will evaluate the performance of prototypic Zircaloy cladding materials against: (1) liquid gallium (Phase 1), (2) various concentrations of Ga 2 O 3 (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of an initial series of tests for phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement (LME), and (3) corrosion mechanical. These tests are designed to determine the corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at ≥ 300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (in parts per million) of gallium in the MOX fuel. While continued migration of gallium into the initially formed intermetallic compound results in large stresses that can lead to distortion, this is also highly unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed

The change and differential findings of lumbar vertebral endplate on MRI

International Nuclear Information System (INIS)

Kanamori, Masahiko; Nobukiyo, Masanori; Shigeta, Tetsuya; Ishihara, Hirokazu; Kawaguchi, Yoshiharu; Yoshino, Osamu

2004-01-01

The purpose of this study is to evaluate the differential findings of vertebral endplate of lumbar spine on MRI. Infected disc showed the T1 low/T2 high (Modic type I) to T2 iso-signal intensity on MRI. Post-operated disc showed the Modic type I findings, although the degenerative disc showed the various kinds of types according to the Modic's classification (type I to III). The characteristics findings of the infected disc were T2-high signal intensity of nucleus purposes and the wide affected area of vertebra including endplate on T1 image compared with T2 image. (author)

Fatigue tests and characterization of resulting microstructure by transmission electron microscope on zircaloy 4

International Nuclear Information System (INIS)

Di Toma, S.; Bertolino, G.; Tolley, A.

2012-01-01

This work reports the results of load controlled tension-tension fatigue tests on Zircaloy 4 (Zy-4). The resulting microstructure, particularly the kind and density of dislocations was characterized using a Transmission Electron Microscope (TEM). Specimens were cut from a rolled plate, with tensile axis parallel and perpendicular to the rolling direction. The results show a significant anisotropy of the mechanical properties due to the strong texture developed during rolling. Mainly type dislocations were observed, only in a longitudinal tensile axis specimen, dislocations were observed with a much lower density. The Schmid factors corresponding to the different glide systems were determined for specific grains in both tensile directions (author)

Irradiation effect on fatigue behaviour of zircaloy-4 cladding tubes

International Nuclear Information System (INIS)

Soniak, A.; Lansiart, S.; Royer, J.; Waeckel, N.

1993-01-01

Since nuclear electricity has a predominant share in French generating capacity, PWR's are required to fit grid load following and frequency control operating conditions. Consequently cyclic stresses appear in the fuel element cladding. In order to characterize the possible resulting clad damage, fatigue tests were performed at 350 deg C on unirradiated material or irradiated stress relieved Zircaloy-4 tube portions, using a special device for tube fatigue by repeated pressurization. It appears that, for high stress levels, the material fatigue life is not affected by irradiation. But the endurance fatigue limit undergoes a decrease from the 350 MPa value for unirradiated material to the 210 MPa value for the material irradiated for four cycles in a PWR. However, this effect seems to saturate with irradiation dose: no difference could be detected between the two cycles results and the corresponding four cycles results. The corrosion effect and the load following influence were also investigated: they do not appear to modify the fatigue behaviour in our experimental conditions

High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

International Nuclear Information System (INIS)

Suzuki, Motoye

1982-03-01

In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

Comparison between zircaloy oxidation in steam and air surroundings

International Nuclear Information System (INIS)

Shawkat, M.E.; Hasaneln, H.; Ali, M.; Parlatan, Y.; Albasha, H.

2013-01-01

The available experimental data for Zircaloy oxidation in air were reviewed. The behavior of the oxidation kinetics at different temperature ranges was described. It was shown that maintaining the oxidation kinetics within the oxide pre-breakaway region can prevent elevated sheath temperatures due to the oxidation process during postulated accidents. The available correlations to model the oxidation kinetics for pre-breakaway region were reviewed and assessed. Zircaloy-air oxidation correlation based on Leistikow-Berg data was determined to be the most suitable correlation to model pre-breakaway kinetics and it was compared to Urbanic-Heidrick correlation which is widely used for Zircaloy oxidation in steam environment. The results showed that the energy release due to the Zircaloy-steam oxidation bounds the energy released due to Zircaloy-air oxidation up to a sheath temperature referred as the “crossover temperature”. Below this temperature, the impact of Zircaloy-air oxidation on fuel sheath temperature transient can be predicted conservatively using the Urbanic-Heidrick steam correlation. The crossover temperature was calculated for isothermal sheath heating as well as transient sheath heat-up assuming three linear heating rates of 0.6, 1.0, and 1.3 K/s. (author)

Corrosion behavior of Zircaloy 4 cladding material. Evaluation of the hydriding effect

International Nuclear Information System (INIS)

Blat, M.

1997-04-01

In this work, particular attention has been paid to the hydriding effect in PIE and laboratory test to validate a detrimental hydrogen contribution on Zircaloy 4 corrosion behavior at high burnup. Laboratory corrosion tests results confirm that hydrides have a detrimental role on corrosion kinetics. This effect is particularly significant for cathodic charged samples with a massive hydride outer layer before corrosion test. PIE show that at high burnup a hydride layer is formed underneath the metal/oxide interface. The results of the metallurgical examinations are discussed with respect to the possible mechanisms involved in this detrimental effect of hydrogen. Therefore, according to the laboratory tests results and PIE, hydrogen could be a strong contributor to explain the increase in corrosion rate at high burnup. (author)

Microstructure and Oxidation Behavior of CrAl Laser-Coated Zircaloy-4 Alloy

Directory of Open Access Journals (Sweden)

Jeong-Min Kim

2017-02-01

Full Text Available Laser coating of a CrAl layer on Zircaloy-4 alloy was carried out for the surface protection of the Zr substrate at high temperatures, and its microstructural and thermal stability were investigated. Significant mixing of CrAl coating metal with the Zr substrate occurred during the laser surface treatment, and a rapidly solidified microstructure was obtained. A considerable degree of diffusion of solute atoms and some intermetallic compounds were observed to occur when the coated specimen was heated at a high temperature. Oxidation appears to proceed more preferentially at Zr-rich region than Cr-rich region, and the incorporation of Zr into the CrAl coating layer deteriorates the oxidation resistance because of the formation of thermally unstable Zr oxides.

Study of hydrogen migration produced during the corrosion of PWR reactors fuel cans in zircaloy 4 and zirconia; Etude du transport de l`hydrogene produit lors de la corrosion des gaines d`elements combustibles des reacteurs a eau sous pression dans la zircone et le zircaloy-4

Energy Technology Data Exchange (ETDEWEB)

Aufore, L

1997-12-12

The corrosion of Zircaloy-4-claddings by water from the primary circuit of nuclear power plant goes hand in hand with the release of hydrogen which penetrates the oxide and then the metal. This work focuses on the mechanisms of hydrogen transport in oxide and in metal. Hydrogen transport in oxide is studied on the basis of corrosion tests performed in the autoclave at 360 deg C. These tests are performed on Zircaloy-4 claddings under different chemical conditions (pure water, and pure water with lithium hydroxide). The distribution of hydrogen in oxide film is measured by SIMS. Hydrogen profiles in the oxide are dependent on the oxide microstructure and vary with oxidation time. These observations are confirmed by experiments in which some samples, previously oxidized in the autoclave, are immersed in heavy water. In the oxide layer, two zones are observed: one external porous zone and one internal dense zone. Deuterium diffusion coefficients in dense oxide are determined using SIMS profiles and Fischer diffusion model. Hydrogen transport in metal is also studied by means of gas-phase permeation experiments. These are set up at different temperature (400-500 deg. C) and under different hydrogen pressures and make it possible to determine the hydrogen diffusion coefficients in a Zircaloy-4 cladding under experimental conditions. All these results lead us to discuss of hydrogen transport evolution in cladding during oxidation. A model taking into account hydrogen transport in oxide and in metal, and the hydrides precipitations is proposed. (author) 110 refs.

Stress-relaxation in bending of zircaloy-4 at 673 K, as a function of cold-work

International Nuclear Information System (INIS)

Povolo, F.

1983-01-01

Stress-relaxation data, in bending, in Zircaloy-4 with different degrees of cold-work are presented. The measurements were performed at 673 K, with six different initial stresses and up to times of the order of 1000 h. The stress-relaxation curves are interpreted in terms of a creep model involving jog-drag and cell formation and some dislocation parameters are calculated from the experimental results. The influence of cold-work on these parameters is discussed. (author)

Crack resistance curve determination of zircaloy-4 cladding

International Nuclear Information System (INIS)

Bertsch, J.; Alam, A.; Zubler, R.

2009-03-01

Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 o C and 350 o C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could be

Crack resistance curve determination of zircaloy-4 cladding

Energy Technology Data Exchange (ETDEWEB)

Bertsch, J.; Alam, A.; Zubler, R

2009-03-15

Fracture mechanics properties of fuel claddings are of relevance with respect to fuel rod integrity. The integrity of a fuel rod, in turn, is important for the fuel performance, for the safe handling of fuel rods, for the prevention of leakages and subsequent dissemination of fuel, for the avoidance of unnecessary dose rates, and for safe operation. Different factors can strongly deteriorate the mechanical fuel rod properties: irradiation damage, thermo-mechanical impact, corrosion or hydrogen uptake. To investigate the mechanical properties of fuel rod claddings which are used in Swiss nuclear power plants, PSI has initiated a program for mechanical testing. A major issue was the interaction between specific loading devices and the tested cladding tube, e.g. in the form of bending or friction. Particular for Zircaloy is the hexagonal closed packed structure of the zirconium crystallographic lattice. This structure implies plastic deformation mechanisms with specific, preferred orientations. Further, the manufacturing procedure of Zircaloy claddings induces a specific texture which plays a salient role with respect to the embrittlement by irradiation or integration of hydrogen in the form of hydrides. Both, the induced microstructure as well as the plastic deformation behaviour play a role for the mechanical properties. At PSI, in a first step inactive thin walled Zircaloy tubes and, for comparison reasons, plates were tested. The validity of the mechanical testing of the non standard tube and plate geometries had to be verified. The used Zircaloy-4 cladding tube sections and small plates of the same wall thickness have been notched, fatigue pre-cracked and tensile tested to evaluate the fracture toughness properties at room temperature, 300 {sup o}C and 350 {sup o}C. The crack propagation has been determined optically. The test results of the plates have been further used to validate FEM calculations. For each sample a complete crack resistance (J-R) curve could

Characteristics of Resistance Spot Welded Ti6Al4V Titanium Alloy Sheets

Directory of Open Access Journals (Sweden)

Xinge Zhang

2017-10-01

Full Text Available Ti6Al4V titanium alloy is applied extensively in the aviation, aerospace, jet engine, and marine industries owing to its strength-to-weight ratio, excellent high-temperature properties and corrosion resistance. In order to extend the application range, investigations on welding characteristics of Ti6Al4V alloy using more welding methods are required. In the present study, Ti6Al4V alloy sheets were joined using resistance spot welding, and the weld nugget formation, mechanical properties (including tensile strength and hardness, and microstructure features of the resistance spot-welded joints were analyzed and evaluated. The visible indentations on the weld nugget surfaces caused by the electrode force and the surface expulsion were severe due to the high welding current. The weld nugget width at the sheets’ faying surface was mainly affected by the welding current and welding time, and the welded joint height at weld nugget center was chiefly associated with electrode force. The maximum tensile load of welded joint was up to 14.3 kN in the pullout failure mode. The hardness of the weld nugget was the highest because of the coarse acicular α′ structure, and the hardness of the heat-affected zone increased in comparison to the base metal due to the transformation of the β phase to some fine acicular α′ phase.

Study of the response of Zircaloy- 4 cladding to thermal shock during water quenching after double sided steam oxidation at elevated temperatures

Energy Technology Data Exchange (ETDEWEB)

Sawarn, Tapan K., E-mail: sawarn@barc.gov.in; Banerjee, Suparna; Kumar, Sunil

2016-05-15

This study investigates the failure of embrittled Zircaloy-4 cladding in a simulated loss of coolant accident condition and correlates it with the evolved stratified microstructure. Isothermal steam oxidation of Zircaloy-4 cladding at high temperatures (900–1200 °C) with soaking periods in the range 60–900 s followed by water quenching was carried out. The combined oxide + oxygen stabilized α-Zr layer thickness and the fraction of the load bearing phase (recrystallised α-Zr grains + prior β-Zr or only prior β-Zr) of clad tube specimens were correlated with the %ECR calculated using Baker-Just equation. Average oxygen concentration of the load bearing phase corresponding to different oxidation conditions was calculated from the average microhardness using an empirical correlation. The results of these experiments are presented in this paper. Thermal shock sustainability of the clad was correlated with the %ECR, combined oxide+α-Zr(O) layer thickness, fraction of the load bearing phase and its average oxygen concentration. - Highlights: • Response of the embrittled Zircaloy-4 clad towards thermal shock, simulated under LOCA condition was investigated. • Thermal shock sustainability of the clad was correlated with its evolved stratified microstructure. • Cladding fails at %ECR value ≥ 29. • To resist the thermal shock, clad should have load bearing phase fraction > 0.44 and average oxygen concentration < 0.69 wt%.

Reaction in air and in nitrogen of pre-oxidised Zircaloy-4 and M5™ claddings

Energy Technology Data Exchange (ETDEWEB)

Duriez, C., E-mail: christian.duriez@irsn.fr [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SEREX-LE2M, Centre de Cadarache, St Paul-Lez-Durance 13115 (France); Drouan, D. [Institut de Radioprotection et de Sûreté Nucléaire (IRSN), PSN-RES/SEREX-LE2M, Centre de Cadarache, St Paul-Lez-Durance 13115 (France); Pouzadoux, G. [Université Technologique de Troyes, BP 2060, Troyes 10010 (France)

2013-10-15

High temperature reactivity in air and in nitrogen of pre-oxidised Zircaloy-4 and M5™ claddings has been studied by thermogravimetry. Claddings were pre-oxidised at low temperature with the aim of simulating spent fuel. Different pre-oxidation modes, inducing significant variation in the pre-oxides microstructure, were compared. The behaviour in air, investigated in the 850–1000 °C temperature range, was found to be strongly dependant on the type of pre-oxide: the compact pre-oxide formed in autoclave (at temperature, pressure, and water chemistry representative of PWR conditions) significantly slows down the degradation in air compared to the bare alloys; on the contrary, a pre-oxide formed at 500 °C at ambient pressure, either in oxygen or in steam, favours the initiation of post-breakaway type oxidation, which in air is associated with nitride formation. The behaviour in nitrogen has been investigated in the 800–1200 °C temperature range, with Zircaloy-4 pre-oxidised at 500 °C in O{sub 2}. Reactivity is low up to 1000 °C but becomes very significant at the highest temperatures investigated, 1100 and 1200 °C. Finally, cladding segments first reacted in N{sub 2} at 1100 °C, were exposed to air and show fast oxidation even at the lowest temperature investigated (600 °C)

The hydrogen generated as a gas and storage in Zircaloy during water quenching

International Nuclear Information System (INIS)

Garcia, Eduardo A.

1999-01-01

A simple one-dimensional diffusion model has been developed for the complex process of Zircaloy oxidation during water quenching, calculating the hydrogen liberated as a gas and the hydrogen stored in the metal. The model was developed on the basis of small-scale separate-effects quench experiments performed at Forschungszentrum Karlsruhe. The new oxide surface and the new metallic surface produced by cracking of the oxide during quenching are calculated for each experiment performed at 1200 , 1400 and 1600 C degrees using as-received Zircaloy-4 (no pre oxidation) and with Zircaloy specimens pre oxidised to give oxide thicknesses of 100μm and 300μm. The results are relevant to accident management in light water reactors. (author)

Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

Energy Technology Data Exchange (ETDEWEB)

Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-05-15

Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system.

Development of Mechanical Sealing and Laser Welding Technology to Instrument Thermocouple for Nuclear Fuel Test Rod

International Nuclear Information System (INIS)

Joung, Chang-Young; Ahn, Sung-Ho; Hong, Jin-Tae; Kim, Ka-Hye; Huh, Sung-Ho

2015-01-01

Zircaloy-4 of the nuclear fuel test rod, AISI 316L of the mechanical sealing parts, and the MI (mineral insulated) cable at a thermocouple instrumentation are hetero-metals, and are difficult to weld to dissimilar materials. Therefore, a mechanical sealing method to instrument the thermocouple should be conducted using two kinds of sealing process as follows: One is a mechanical sealing process using Swagelok, which is composed of sealing components that consists of an end-cap, a seal tube, a compression ring and a Swagelok nut. The other is a laser welding process used to join a seal tube, and an MI cable, which are made of the same material. The mechanical sealing process should be sealed up with the mechanical contact compressed by the strength forced between a seal tube and an end-cap, and the laser welding process should be conducted to have no defects on the sealing area between a seal tube and an MI cable. Therefore, the mechanical sealing and laser welding techniques need to be developed to accurately measure the centerline temperature of the nuclear fuel test rod in an experimental reactor. The mechanical sealing and laser welding tests were conducted to develop the thermocouple instrumentation techniques for the nuclear fuel test rod. The optimum torque value of a Swagelok nut to seal the mechanical sealing part between the end-cap and seal tube was established through various torque tests using a torque wrench. The optimum laser welding conditions to seal the welding part between a seal tube and an MI cable were obtained through various welding tests using a laser welding system

Analysis of the texture of zircaloy-4 sheet by crystallite orientation distribution function

International Nuclear Information System (INIS)

Ryoo, Hwei Soo; Hwang Sun Keum

1990-01-01

In order to analyze the texture variation of Zircaloy-4 sheet the Roe's method of calculating the crystallite orientation distribution function(CODF) for hcp system was computer programmed. The coefficients W lmn of CODF were calculated from plane-normal distribution pole figures obtained by X-ray diffraction, and the CODF was computed from a series expansion of spherical harmonics. The Legendre function, which is the basis of the harmonics, was computed up to l=16 to account for the symmetry systems of specimen and hcp crystal. A cross-rolling followed by beta-phase heat treatment and furnace cooling increased the density of basal poles along the sheet normal direction and rotated prism poles around the c axis. (Author)

Zircaloy oxidation studies

International Nuclear Information System (INIS)

Prater, J.T.; Beauchamp, R.H.; Saenz, N.T.

1985-06-01

The oxidation kinetics of Zircaloy-4 in steam have been determined at 1300-2400 0 C. Growth of the ZrO 2 and α-Zr layers display parabolic behavior over the entire temperature range studied. A discontinuity in the oxidation kinetics at 1510 0 C causes rates to increase above those previously established by the Baker-Just relationship. This increase coincides with the tetragonal-to-cubic phase transformation in ZrO/sub 2-x/. No discontinuity in the oxide growth rate is observed upon melting of Zr(0). The effects of temperature gradients have been taken into account and corrected values representative of near-isothermal conditions have been computed

High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments

International Nuclear Information System (INIS)

Chung, H.M.; Garde, A.M.; Kassner, T.F.

1977-01-01

The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350 0 C

Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4

International Nuclear Information System (INIS)

Racine, A.

2005-09-01

In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

Interactions of Zircaloy cladding with gallium: 1998 midyear status

International Nuclear Information System (INIS)

Wilson, D.F.; DiStefano, J.R.; Strizak, J.P.; King, J.F.; Manneschmidt, E.T.

1998-06-01

A program has been implemented to evaluate the effect of gallium in mixed-oxide (MOX) fuel derived from weapons-grade (WG) plutonium on Zircaloy cladding performance. The objective is to demonstrate that low levels of gallium will not compromise the performance of the MOX fuel system in a light-water reactor. The graded, four-phase experimental program was designed to evaluate the performance of prototypic Zircaloy cladding materials against (1) liquid gallium (Phase 1), (2) various concentrations of Ga 2 O 3 (Phase 2), (3) centrally heated surrogate fuel pellets with expected levels of gallium (Phase 3), and (4) centrally heated prototypic MOX fuel pellets (Phase 4). This status report describes the results of a series of tests for Phases 1 and 2. Three types of tests are being performed: (1) corrosion, (2) liquid metal embrittlement, and (3) corrosion-mechanical. These tests will determine corrosion mechanisms, thresholds for temperature and concentration of gallium that may delineate behavioral regimes, and changes in the mechanical properties of Zircaloy. Initial results have generally been favorable for the use of WG-MOX fuel. The MOX fuel cladding, Zircaloy, does react with gallium to form intermetallic compounds at ≥300 C; however, this reaction is limited by the mass of gallium and is therefore not expected to be significant with a low level (parts per million) of gallium in the MOX fuel. Although continued migration of gallium into the initially formed intermetallic compound can result in large stresses that may lead to distortion, this was shown to be extremely unlikely because of the low mass of gallium or gallium oxide present and expected clad temperatures below 400 C. Furthermore, no evidence for grain boundary penetration by gallium has been observed

Effets de la radiolyse de l'air humide et de l'eau sur la corrosion de la couche d'oxyde du Zircaloy-4 oxydé

OpenAIRE

Guipponi , Claire

2009-01-01

Pas de résumé donné.; Les Colis Standards de Déchets Compactés (CSD-C) sont des déchets issus du retraitement des assemblages de combustibles nucléaires. Ils sont en partie constitués des gaines oxydées de Zircaloy-4. Ces pièces métalliques sont cisaillées avant d'être placées dans un étui en acier et compactées sous forme de galettes. Ces galettes contiennent des traces de produits d'activation, de produits de fission et d'actinides présents à la surface du Zircaloy-4 oxydé. Dans l'hypothèse...

Characterization of Zircaloy-4 oxide layers by impedance spectroscopy

International Nuclear Information System (INIS)

Barberis, P.

1999-01-01

Two Zircaloy-4 type alloys with different tin contents (0.5 and 1.2 wt%) have been oxidized in autoclave (400 C in steam) for several durations (1-140 days). The film has been characterized by electrochemical impedance spectroscopy (EIS). Several soaking times have been investigated (up to 40 days). The Cole-Cole representation has been used to display and study the data. A simple electrical model has been derived from the observed spectra: the electrical circuit includes two RC loops in series, whose capacitances are frequency dispersed. It is thoroughly related to the layer structure. It has been shown that even before the kinetic transition, the film is constituted of three parts: an inner layer which is compact, an outer layer subdivided in an external region immediately soaked by the electrolyte, and an internal one in which electrolyte diffusion processes can take place. The kinetic transition is interpreted in terms of an abrupt 'compacity' change, both layers degrading at this point. The alloy with high tin content exhibits higher dispersive properties of the oxide layer formed on it, in correlation with its faster oxidation kinetics. (orig.)

Evaluation on vertebral endplate injury and adjacent intervertebral disk injury of patients with osteoporotic vertebral compression fractures by MRI and its clinical significance

International Nuclear Information System (INIS)

Shen Yu; Shen Huiliang; Fang Xiutong; Zhang Wenbo

2012-01-01

Objective: To investigate the relationship between vertebral endplate injury and adjacent intervertebral disk injury of patients with acute or sub-acute osteoporotic vertebral compression fractures (OVC-F) by MRI, and to provide basis for diagnosis of OVCF. Methods: The clinical data of a total of 66 patients with OVCF underwent vertebroplasty (76 fracture of vertebral bodies) were selected. The vertebral endplate injury and adjacent intervertebral disk injury of OVCF patients were detected by MRI. Results: There were 57 vertebral endplate injury in 76 fracture vertebral bodies (75% ). There were only 27 vertebral bodies with vertebral endplate injury in 57 fracture vertebral bodies with endplate injury (47% ), and 22 vertebral with superior and inferior vertebral endplate injury (39% ), and 8 vertebral bodies with inferior vertebral endplate injury (14% ). There were 48 vertebral bodies with intervertebral disc injury in 76 fracture vertebral bodies (63% ). There were 22 intervertebral disc injury located above the fracture of the lumbar spine in 48 vertebral bodies with intervertebral disc injury (45% ), and 19 fracture vertebral bodies with upper and lower intervertebral disc injury (40% ), and 7 intervertebral injuries located below the fracture of the lumbar spine (15% ). Conclusion: Vertebral endplate injury is frequently associated with the adjacent intervertebral disk injury. The clinical diagnosis and treatment should be emphasized in the fracture vertebral endplate damage and adjacent intervertebral disc injury. (authors)

Stress Corrosion Cracking of Zircaloy-4 in Halide Solutions: Effect of Temperature

Directory of Open Access Journals (Sweden)

Farina S.B.

2002-01-01

Full Text Available Zircaloy-4 was found to be susceptible to stress corrosion cracking in 1 M NaCl, 1 M KBr and 1 M KI aqueous solutions at potentials above the pitting potential. In all the solutions tested crack propagation was initially intergranular and then changed to transgranular. The effect of strain rate and temperature on the SCC propagation was investigated. An increase in the strain rate was found to lead to an increase in the crack propagation rate. The crack propagation rate increases in the three solutions tested as the temperatures increases between 20 and 90 °C. The Surface-Mobility SCC mechanism accounts for the observation made in the present work, and the activation energy predicted in iodide solutions is similar to that found in the literature.

Influence of sintering time on distribution of alloying elements composition in Zircaloy pellet

International Nuclear Information System (INIS)

Sigit; Muchlis B; Widjaksana; Eric, J.; Suryana, RA; Gunawan

1996-01-01

Influence of sintering time on distribution of alloying elements composition in zircaloy pellet has been studied. Zircaloy pellets were obtained by pressing of Zr, Fe, Cr and Sn powders mixture in adequate composition of zircaloy-4, than the green pellets were sintered at 1100 o C for 1 - 3 hours. The alloying elements (Fe, Cr and Sn) composition in zircaloy pellets as sintering product were determined by Scanning Electron Microscope - Energy Dispersive X-Ray Analyser (SEM-EDAX). The experiments showed that there was an accumulation of Sn in a site of the zircaloy green pellet of 17.46 %, but after sintering process, the Sn was distributed everywhere. The influence of sintering time up to 1 hour showed a decreasing Sn composition from 9 % to 2 % which then relatively constant, while for Fe and Cr its decreasing was relatively small, i.e. : 1.86 % to 0.6 % and 1.04 % to 0.17 % respectively. The sintering process revealed no clear grain boundaries and powder homogenization did not complete. Observation on metallographic photos showed that this condition was in initial stage of sintering process where there was a complex phenomenon i.e.: no powder homogenization in green pellet or initial heating rate was extremely quick

Zircaloy behaviour in high temperature irradiated water

International Nuclear Information System (INIS)

Urbanic, V.F.

1982-04-01

The corrosion and hydriding of Zircaloy during irradiation in high temperature water is strongly dependent on the oxygen concentration of the water. Corrosion tests in the NRX and NRU research reactors using small samples have demonstrated the importance of water chemistry in maintaining Zircaloy corrosion and hydriding within acceptable limits. Zircaloy fuel cladding develops non-uniform, patch-type oxides during irradiation in hich temperature water containing dissolved oxygen. Results from examinations of prototype fuel cladding irradiated in the research reactors are presented to show how local variations in coolant flow, fast neutron flux, metallurgical structure and surface condition can influence the onset of non-uniform corrosion under these conditions. Destructive examinations of CANDU-PHW reactor fuel cladding have emphasized the importance of good chemistry control, especially the dissolved oxygen concentration of the water. When reactor coolants are maintained under normal reducing conditions at high pH (5 to 10 cm 3 D 2 /kg D 2 O; 2 /kg D 2 O; pH > 10 with LiOD), Zircaloy cladding develops non-uniform, patch-type oxides. These patch-type oxides tend to coalesce with time to form a thick, uniform oxide layer after extended exposure. Under reducing coolant conditions, Zircaloy cladding absorbs less than 200 mg D/kg Zr (approximately 2.5 mg/dm 2 equivalent hydrogen) in about 500 days. With oxygen in the coolant, deuterium absorption is considerably less despite the significant increase in corrosion under such conditions

The hydrogen generated as a gas and storage in Zircaloy during steam quenching

International Nuclear Information System (INIS)

Garcia, Eduardo A.

2000-01-01

A simple one-dimensional diffusion model has been developed for the complex process of Zircaloy oxidation during steam quenching, calculating the hydrogen liberated as a gas and the hydrogen stored in the metal. The model was developed on the basis of small-scale separate-effects quench experiments performed at Forschungszentrum Karlsruhe. The new oxide surface and the new metallic surface produced by cracking of the oxide during quenching are calculated for each experiment performed at 1200 centigrade, 1400 centigrade and 1600 centigrade using as-received Zircaloy-4 (no pre-oxidation) and with Zircaloy specimens pre-oxidized to give oxide thickness of 100μm and 300μm. The results are relevant to accident management in nuclear power plants. (author)

Zircaloy cladding ID/OD oxidation studies. Final report

International Nuclear Information System (INIS)

Westerman, R.E.; Hesson, G.M.

1977-11-01

The ID/OD oxide ratio that forms on Zircaloy tubing at temperatures relevant to postulated LOCA conditions was measured as a function of time, temperature, and distance from the rupture. The average ratio at the rupture position was less than unity, and decreased with decreasing test time and increasing distance from the point of rupture. The maximum observed ID/OD oxide ratio was 1.4. Ratios in excess of unity were typically found to be a consequence of the OD oxide being thinner than would have been anticipated from the nominal test conditions. Confirmatory data were also obtained on the isothermal oxidation kinetics of Zircaloy. These data are in good agreement with those obtained by other investigators and confirm the conservative nature of the Baker-Just equation that is required for use in licensing calculations

Mechanical analysis of surface-coated zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Lee, You Ho; Lee, Jeong Ik; No, Hee Cheon [Dept. of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2017-08-15

A structural model for stress distributions of coated Zircaloy subjected to realistic incore pressure difference, thermal expansion, irradiation-induced axial growth, and creep has been developed in this study. In normal operation, the structural integrity of coating layers is anticipated to be significantly challenged with increasing burnup. Strain mismatch between the zircaloy and the coated layer, due to their different irradiation-induced axial growth, and creep deformation are found to be the most dominant causes of stress. This study suggests that the compatibility of the high temperature irradiation-induced strains (axial growth and creep) between zircaloy and the coating layer and the capability to undergo plastic strain should be taken as key metrics, along with the traditional focus on chemical protectiveness.

Uniaxial ratcheting behavior of Zircaloy-4 tubes at room temperature

International Nuclear Information System (INIS)

Wen, Mingjian; Li, Hua; Yu, Dunji; Chen, Gang; Chen, Xu

2013-01-01

In this study, a series of uniaxial tensile, strain cycling and uniaxial ratcheting tests were conducted at room temperature on Zircaloy-4 (Zr-4) tubes used as nuclear fuel cladding in Pressurized Water Reactors (PWRs) for the purpose to investigate the uniaxial ratcheting behavior of Zr-4 and the factors which may influence it. The experimental results show that at room temperature this material features cyclic softening remarkably within the strain range of 1.6%, and former cycling under larger strain amplitude cannot retard cyclic softening of later cycling under lower strain amplitude. Uniaxial ratcheting strain accumulates in the direction of mean stress, and the ratcheting stain level is larger under tensile mean stress than that under compressive mean stress. Uniaxial ratcheting strain level increases with the increase of mean stress and stress amplitude, and decreases with the increase of loading rate. The sequence of loading rate appears to have no effects on the final ratcheting strain accumulation. Loading history has great influence on the uniaxial ratcheting behavior. Lower stress level after loading history with higher stress level leads to the shakedown of ratcheting. Higher loading rate after loading history with lower loading rate brings down the ratcheting strain rate. Uniaxial ratcheting behavior is sensitive to compressive pre-strain, and the decay rate of the ratcheting strain rate is slowed down by pre-compression

Zirconium metal-water oxidation kinetics. III. Oxygen diffusion in oxide and alpha Zircaloy phases

International Nuclear Information System (INIS)

Pawel, R.E.

1976-10-01

The reaction of Zircaloy in steam at elevated temperature involves the growth of discrete layers of oxide and oxygen-rich alpha Zircaloy from the parent beta phase. The multiphase, moving boundary diffusion problem involved is encountered in a number of important reaction schemes in addition to that of Zircaloy-oxygen and can be completely (albeitly ideally) characterized through an appropriate model in terms of oxygen diffusion coefficients and equilibrium concentrations for the various phases. Conversely, kinetic data for phase growth and total oxygen consumption rates can be used to compute diffusion coefficients. Equations are developed that express the oxygen diffusion coefficients in the oxide and alpha phases in terms of the reaction rate constants and equilibrium solubility values. These equations were applied to recent experimental kinetic data on the steam oxidation of Zircaloy-4 to determine the effective oxygen diffusion coefficients in these phases over the temperature range 1000--1500 0 C

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

Energy Technology Data Exchange (ETDEWEB)

Isabella J van Rooyen

2012-09-01

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

Effects of rocuronium and vecuronium on initial rundown of endplate potentials in the isolated phrenic nerve diaphragm preparation of rats

OpenAIRE

Li, Jun; Liu, Yong-Qin; Zhang, Han-Ting

2013-01-01

Rocuronium and vecuronium, two non-depolarizing neuromuscular blockers, have been widely used in surgery procedures. However, their electrophysiological properties need to be more widely explored. We examined the effects of rocuronium and vecuronium on initial rundown of endplate potential amplitudes in the non-uniform stretched muscle preparation of the rat isolated phrenic nerve diaphragm. More specifically, the endplate potentials were recorded with one microelectrode from a single endplat...

Interaction between aluminium oxide pellets and Zircaloy tubes in steam atmospheres at temperatures above 12000C

International Nuclear Information System (INIS)

Hagen, S.; Hofmann, P.; Schanz, G.; Sepold, L.

1988-09-01

The burnable poison rods in light water reactors (LWR) consist of Al 2 O 3 /B 4 C pellets surrounded by Zircaloy-4 cladding tubes. In the Al 2 O 3 /B 4 C pellets of a LWR rod alumina is the main constituent (98.6 wt.-%) whereas boron carbide acts as neutron absorber. Failure of the Al 2 O 3 /Zircaloy test rods started at 1350 0 C when first droplets of molten material were observed running down the test bundle forming bundle blockages upon solidification. Post test examinations revealed that the process of liquefaction was initiated by a reduction of alumina by Zircaloy resulting in a (Zr, Al, O) melt which decomposed on cooldown into two metallic phases, a (Zr, Al) alloy and oxygen-stabilized a-Zr(O). The components of an extremely porous ceramic melt were also Zr, Al, and oxygen but with a higher oxygen content compared to the metallic melt. The ceramic melt decomposes on cooldown into an Al 2 O 3 /ZrO 2 eutectic with various amounts of primary constituents. Other types of relocated material were due to melting of essentially unreacted Zircaloy cladding and to debris formation by fracturing of oxidized cladding and Al 2 O 3 pellets stack residues. The interactions between Al 2 O 3 and Zircaloy occurring in a burnable poison rod are furthermore important for the behavior of the entire LWR core because the generated metals are able to attack the UO 2 chemically and dissolve or liquefy the fuel even below the melting point of Zircaloy (1760 0 C). As a result, fuel elements which contain burnable poison rods are expected to fail under severe accident conditions at about 1500 0 C. (orig./HP) [de

Biaxial creep deformation of Zircaloy-4 PWR fuel cladding in the alpha,(alpha + beta) and beta phase temperature ranges

International Nuclear Information System (INIS)

Donaldson, A.T.; Healey, T.; Horwood, R.A.L.

1985-01-01

The biaxial creep behaviour of Zircaloy-4 fuel cladding has been determined at temperatures between 973 - 1073 K in the alpha phase range, in the duplex (alpha + beta) region between 1098 - 1223 K and in the beta phase range between 1323 - 1473 K. This paper presents the creep data together with empirical equations which describe the creep deformation response within each phase region. (author)

Oxidation of zircaloy-2 in high temperature steam

International Nuclear Information System (INIS)

Ikeda, Seiichi; Ito, Goro; Ohashi, Shigeo

1975-01-01

Oxidation tests were conducted for zircaloy-2 in steam at temperature ranging from 900 to 1300 0 C to clarify its oxidation kinetics as a nuclear fuel cladding materials in case of a loss-of-coolant accident. The influence of maximum temperature and heating rate of the specimen on its oxidation rate in steam was investigated. The changes in mechanical properties of the specimens after oxidation tests are also studied. The results obtained were summarized as follows: (1) The weight of the specimen after oxidation in steam increased two times as the time required to reach the maximum temperature increased from 1 to 10 mins. (2) The kinetics of oxidation of zircaloy-2 in steam were not affected by the difference in the surface condition before test such as chemical polishing or pre-oxidation in steam. (3) The dominant growth of oxide film on the surface of zircaloy-2 was observed at the initial stage of oxidation in steam. However, the thickness of oxygen-rich solid solution layer under the film increased gradually with the progress of oxidation and the ratio of oxygen in oxide to that in solid solution has a constant value of 8:2. (4) The breakaway took place only in the specimen subjected to 900 0 C repeated heating. This penomenon was caused by the local growth of the oxide below a crack of the oxide film resulting from the reheating of the specimen. (5) The results of bending tests showed that the deflection until fracture of the specimen was smaller for the one heated at a higher temperature even if the weight increase was of the same order of magnitude for both specimens. (6) It was concluded that the ductility of zircaloy-2 decreased remarkably at a heating temperature in excess of 1100 0 C for more than 5 min. (auth.)

Determination of Boron in Zircaloy by using ICP-AES and Colorimetry

Energy Technology Data Exchange (ETDEWEB)

Kim, Jong-Goo; Pyo, Hyung-Ryul; Choi, Kwang-Soon; Han, Sun-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2007-10-15

Zircaloy has been being widely used in the nuclear industry because of the low cross section of Zirconium against a thermal neutron. Accurate composition data of Zircaloy for Hf, B, and so on having a high cross section against thermal neutron is important to use it as a nuclear material. Accordingly proper determination methods of these elements in Zircaloy are needed. In this study, the application of two methods, ICP-AES and a colorimetry using methylene blue were investigated in order to establish a proper analysis method of Boron in the range from tens to hundreds ug B/g sample of Zircaloy.

Determination of Boron in Zircaloy by using ICP-AES and Colorimetry

International Nuclear Information System (INIS)

Kim, Jong-Goo; Pyo, Hyung-Ryul; Choi, Kwang-Soon; Han, Sun-Ho

2007-01-01

Zircaloy has been being widely used in the nuclear industry because of the low cross section of Zirconium against a thermal neutron. Accurate composition data of Zircaloy for Hf, B, and so on having a high cross section against thermal neutron is important to use it as a nuclear material. Accordingly proper determination methods of these elements in Zircaloy are needed. In this study, the application of two methods, ICP-AES and a colorimetry using methylene blue were investigated in order to establish a proper analysis method of Boron in the range from tens to hundreds ug B/g sample of Zircaloy

Fluorimetric determination of uranium in zirconium and zircaloy alloys; Determinacion fluorimetrica de uranio en aleaciones de zirconio y zircaloy

Energy Technology Data Exchange (ETDEWEB)

Acosta L, E [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

1991-05-15

The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

Research on electron beam welding technology of steel HR-4

International Nuclear Information System (INIS)

Guo Peng; Guan Kai

2001-01-01

The electron beam weldability of HR- 4 steels (J75 and J90) is studied and the welding parameters needed for design and usage are presented. The assessment on the effect of mechanical properties by different processing order of welding and heat-treatment is made

Solvent effects on stress corrosion cracking of zirconium and Zircaloy-4 in iodine

International Nuclear Information System (INIS)

Farina, Silvia B.; Duffo, Gustavo S.; Galvele, Jose R.

2000-01-01

Localized corrosion (pitting, intergranular attack and stress corrosion cracking) of Zircaloy-4 and its principal component, zirconium, was investigated in solutions of iodine in different alcohols (methanol, ethanol, 1-propanol, 1-butanol, 1-pentanol and 1-octanol). Intergranular attack was found in all of the solutions tested, and the attack velocity increases when the size of the alcohol molecule decreases. In some cases it was found that intergranular attack is accompanied by pitting. Slow strain-rate experiments showed that the propagation rate of stress corrosion cracks also depends on the size of the solvent molecule. From these results it may be inferred that the cause of the variation in the velocity is the steric hindrance of the alcohol molecules. The surface mobility SCC mechanism may account for these results. (author)

Investigation of the high temperature steam oxidation of Zircaloy 4 cladding tubes

International Nuclear Information System (INIS)

Leistikow, S.; Berg, H. v.; Kraft, R.; Pott, E.; Schanz, G.

1979-01-01

Also for the ORNL Zircaloy 4 cladding material, an intermediate decrease of the proportion of the ZrO 2 /α-phase layer was found, followed by an drastic increase when the breakaway of the ZrO 2 -scale occurred. Other reasons for small divergencies were evaluated, for instance temperature and time measurements, metallographic evaluation of layer thicknesses, consequences of one-sided (ORNL) and double-sided (KfK) oxidation. The so-called anomalous effect of steam oxidation during temperature transients was reproduced qualitatively and-in case that a reduced gain of oxygen was observed-explained by the predominant existence of the monoclinic oxide phase. The creep-rupture tests below 800 0 C showed a moderate prolongation of time-to-rupture when the tests were performed in steam (or after preoxidation in steam) instead of argon. Also slightly reduced maximum circumferential strain could be measured. (orig./RW) [de

Autophagy Is a Protective Response to the Oxidative Damage to Endplate Chondrocytes in Intervertebral Disc: Implications for the Treatment of Degenerative Lumbar Disc

Directory of Open Access Journals (Sweden)

Ke Chen

2017-01-01

Full Text Available Low back pain (LBP is the leading cause of disability in the elderly. Intervertebral disc degeneration (IDD was considered as the main cause for LBP. Degeneration of cartilaginous endplate was a crucial harmful factor during the initiation and development of IDD. Oxidative stress was implicated in IDD. However, the underlying molecular mechanism for the degeneration of cartilaginous endplate remains elusive. Herein, we found that oxidative stress could induce apoptosis and autophagy in endplate chondrocytes evidenced by western blot analysis, flow cytometry, immunofluorescence staining, GFP-LC3B transfection, and MDC staining. In addition, we also found that the apoptosis of endplate chondrocytes was significantly increased after the inhibition of autophagy by bafilomycin A1 shown by flow cytometry. Furthermore, mTOR pathway upstream autophagy was greatly suppressed suggested by western blot assay. In conclusion, our study strongly revealed that oxidative stress could increase autophagy and apoptosis of endplate chondrocytes in intervertebral disc. The increase of autophagy activity could prevent endplate chondrocytes from apoptosis. The autophagy in endplate chondrocytes induced by oxidative stress was mTOR dependent. These findings might shed some new lights on the mechanism for IDD and provide new strategies for the treatments of IDD.

Hydrogen isotope storage in zircaloy scrap

Energy Technology Data Exchange (ETDEWEB)

Lee, H. S.; Kuk, I. H.; Chung, H.; Paek, S. W.; Kang, H. S

1999-08-01

8 MCi of tritium a year will be produced after wolsong TRF is in operation. The metal hydride form is one of useful tritium storage. The metals in use for metal hydride are uranium, titanium, etc., however uranium is limited to use by regulation, and titanium is relatively costly. Both metals are not produced in country but whole amount is imported. On the other hand 2,000kg of zircaloy scrap is produced by CANDU nuclear fuel fabrication process, which is also useful for hydrogen storage. The purpose of this study is to evaluation of hydrogen absorption capacity for zircaloy scrap that is produced as waste by CANDU nuclear fuel fabrication process. The sample evacuated for an hour at 1000 deg C. The strip showed higher capacity : 0.7 at 25 deg C, 2.0 at 200 deg C, 2.0 at 200 deg C, 2.0 at 400 deg C, respectively. The H/M values for commercial zircaloy sponge were 2.0 at 25 deg C and 2.0 at 400 deg C.

Hydrogen isotope storage in zircaloy scrap

International Nuclear Information System (INIS)

Lee, H. S.; Kuk, I. H.; Chung, H.; Paek, S. W.; Kang, H. S.

1999-08-01

8 MCi of tritium a year will be produced after wolsong TRF is in operation. The metal hydride form is one of useful tritium storage. The metals in use for metal hydride are uranium, titanium, etc., however uranium is limited to use by regulation, and titanium is relatively costly. Both metals are not produced in country but whole amount is imported. On the other hand 2,000kg of zircaloy scrap is produced by CANDU nuclear fuel fabrication process, which is also useful for hydrogen storage. The purpose of this study is to evaluation of hydrogen absorption capacity for zircaloy scrap that is produced as waste by CANDU nuclear fuel fabrication process. The sample evacuated for an hour at 1000 deg C. The strip showed higher capacity : 0.7 at 25 deg C, 2.0 at 200 deg C, 2.0 at 200 deg C, 2.0 at 400 deg C, respectively. The H/M values for commercial zircaloy sponge were 2.0 at 25 deg C and 2.0 at 400 deg C

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.

Behavior of four-bolt extended end-plate connection subjected to lateral loading

Directory of Open Access Journals (Sweden)

Elsayed Mashaly

2011-03-01

Full Text Available A considerable number of literatures have been published on the behavior of end-plate connections in ordinary moment-resisting frames. It was found, experimentally, that this type of connection might act as either a fully-rigid or a semi-rigid connection depending mainly on the thickness of the end plate and the diameter of bolts. In recent years, due to their good ductility and their good ability of energy dissipation, extended end-plate connections are recommended to be widely used in special moment-resisting frames subjected to lateral loads. The purpose of this study is to investigate the effect of both the material and geometric properties of four-bolt extended end-plate connections upon their behavior when subjected to lateral loading. This is done through a parametric study upon a finite element model using the multi-purpose software package ANSYS. The parametric study takes into account 12 parameters which are expected to be effective on the behavior of the studied connection. The results are presented by the relation between the storey drift, which represents the rotation of the connection and the applied lateral load, which simulates the moment on the connection. The results verify that the chosen parameters are considered effective depending on the energy dissipation of the connection.

Weldability of Al4C3-Al composites via diffusion welding technique

International Nuclear Information System (INIS)

Arik, Halil; Aydin, Mustafa; Kurt, Adem; Turker, Mehmet

2005-01-01

In this study, Al-Al 4 C 3 composites, produced by powder metallurgy in situ techniques, were joined by diffusion welding technique at 250 MPa pressure with various welding temperatures and durations. Microstructures and shear strengths of the joined areas were determined. Al powders were mixed with 2% carbon black and milled in a high energy ball mill (mechanical alloying) for up to 20 h. In order to obtain cylindrical blanks with 10 mm in diameter and 15 mm in height, powders were compacted in a single action press at 1000 MPa. Samples were sintered in Ar atmosphere at 650 deg C and metal matrix composite (MMC) containing 8% Al 4 C 3 particles were produced. Products were then joined to each other by using diffusion welding techniques. Scanning electron microscopy examination was carried out on the welded interfaces and shear tests were conducted to the sample interfaces to find out the effect of welding temperatures and duration on the weldability properties. It was found that high welding temperatures resulted in increase of both joined strength and shear properties. However, increase in welding duration did not make any detectable changes. Results indicated that MMC could be joined by diffusion welding technique successfully with the 88% strength of base material

Effect of prosthesis endplate lordosis angles on L5-S1 kinematics after disc arthroplasty.

Science.gov (United States)

Tsitsopoulos, Parmenion P; Wojewnik, Bartosz; Voronov, Leonard I; Havey, Robert M; Renner, Susan M; Zelenakova, Julia; McIntosh, Braden; Carandang, Gerard; Abjornson, Celeste; Patwardhan, Avinash G

2012-06-01

We hypothesized that L5-S1 kinematics will not be affected by the lordosis distribution between the prosthesis endplates. Twelve cadaveric lumbosacral spines (51.3 ± 9.8 years) were implanted with 6° or 11° prostheses (ProDisc-L) with four combinations of superior/inferior lordosis (6°/0°, 3°/3°, 11°/0°, 3°/8°). Specimens were tested intact and after prostheses implantation with different lordosis distributions. Center of rotation (COR) and range of motion (ROM) were quantified. Six-degree lordosis prostheses (n = 7) showed no difference in flexion-extension ROM, regardless of design (6°/0° or 3°/3°) (p > 0.05). In lateral bending (LB), both designs reduced ROM (p lordosis prostheses (n = 5) showed no difference in flexion-extension ROM for either design (p > 0.05). LB ROM decreased with distributed lordosis prostheses (3°/8°) (p lordosis distribution among the two prosthesis endplates. The ProDisc-L prosthesis design where all lordosis is concentrated in the superior endplate yielded COR locations that were anterior and caudal to intact controls. The prosthesis with lordosis distributed between the two endplates yielded a COR that tended to be closer to intact. Further clinical and biomechanical studies are needed to assess the long-term impact of lordosis angle distribution on the fate of the facet joints.

Development of the manufacture and process for DUPIC fuel elements; development of the quality evaluation techniques for end cap welds of DUPIC fuel elements

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Tae; Choi, Myong Seon; Yang, Hyun Tae; Kim, Dong Gyun; Park, Jin Seok; Kim, Jin Ho [Yeungnam University, Kyongsan (Korea)

2002-04-01

The objective of this research is to set up the quality evaluation techniques for end cap welds of DUPIC fuel element. High temperature corrosion test and the SCC test for Zircaloy-4 were performed, and also the possibility of the ultrasonic test technique was verified for the quality evaluation and control of the laser welds in the DUPIC fuel rod end cap. From the evaluation of corrosion properties with measuring the weight gain and observing oxide film of the specimen that had been in the circumstance of steam(400 .deg. C, 1,500 psi) by max. 70 days later, the weight gain of the welded specimens was larger than original tube and the weight increasing rate increased with the exposed days. For the Development of techniques for ultrasonic test, semi-auto ultrasonic test system has been made based on immersion pulse-echo technique using spherically concentrated ultrasonic beam. Subsequently, developed ultrasonic test technique is quite sensible to shape of welds in the inside and outside of tube as well as crack, undercut and expulsion, and also this ultrasonic test, together with metallurgical fracture test, has good reliance as enough to be used for control method of welding process. 43 refs., 47 figs., 8 tabs. (Author)

The effect of texture, heat treatment and elongation rate on stress corrosion cracking in irradiated zircaloy

International Nuclear Information System (INIS)

Pettersson, K.; Stany, W.; Hellstrand, E.

1979-03-01

Irradiated zircaloy samples with different textures and heat treatments have been tested concerning stress corrosion. Irradiated samples of Zr-1Nb, pure Zr and beta quenched zircaloy have also been investigated. Stress-relieve annealled zircaloy is even after irradiation more sensitive to stress corrosion than recrystallized zircaloy. Zr-1Nb and beta quenched zircaloy are much more sinsitive to stress corrosion than the samples with different textures. As a rule irradiated zircaloy is sensitive to stress corrosion at stresses far below the yield point. The breaking stress decreases with the elongation rate. The extension of cracks is much faster in irradiated zircaloy than in unirradiated zircaloy. There is no simple failure criterium for irradiated zircaloy. However for a certain stress and a certain elongation rate the probability for a failure before this stress is reached with a constant elongation rate can be given. (E.R.)

Calorimetric determination of the δ hydride dissolution enthalpy in Zircaloy-4

International Nuclear Information System (INIS)

Vizcaino, Pablo; Banchik, Abraham D.

2003-01-01

In this work, the dissolution enthalpy, ΔH δ→α , of the δ hydride phase in the αZr matrix in Zircaloy-4 has been determined with a differential scanning calorimeter (DSC) in two different ways: by means of a vant Hoff equation, measuring the terminal solubility temperature in dissolution, TSSd, and by direct measurement of the dissolution heat, Q δ→α , as the area between the base line and the calorimetric curve. The application of the DSC technique to the hydride dissolution heat measurements, a transformation which covers an extended temperature range, is completely original and requires a special treatment of the calorimetric curve. These measurements were done on samples, which practically cover the whole solubility range of hydrogen in αZr phase (80-640 ppm). The values obtained, 36.9 kJ/mol H and 39.3 kJ/mol H respectively, are self-consistent and in good agreement with the values of the more recent revisions, but reduces considerably the scatter of the literature data. (author)

Effect of an Excess of Chromium and Hydriding on Zircaloy-4

International Nuclear Information System (INIS)

Soldati, A.; Ghilarducci, A. A; Corso, H.L.; Peretti, H.A.; Bolcich, J.C

2003-01-01

Results of mechanical properties and microstructure morphologies of zircaloy-4 are presented.They were obtained in several laboratory made samples of chemical composition modified with respect to the ASTM B 350 by the addition of alloying elements as well as hydrides.This work is focused mainly on the effect of 900 ppm of additional Cr in excess as compared with the standard composition alloy and with two other laboratory alloys studied before, containing 250 ppm of Ni in excess and 1000 ppm of Fe in excess, respectively.The study is carried out by means of tensile tests at room temperature and at 240C, hardness tests, SEM observations and EDS microanalysis.The neutron irradiation was carried out at Bariloche in the RA6-CNEA reactor.The results indicate that precipitates concentrate along grain boundaries in all cases, and that for higher contents of alloying elements corresponds a higher quantity of precipitates and smaller grain sizes.Except for the hydrided sample, the fracture is ductile with cavities nucleated at precipitates

The effect of welding process and shielding atmosphere on the AlMg4.5Mn weld metal properties

Energy Technology Data Exchange (ETDEWEB)

Prokic Cvetkovic, Radica; Popovic, Olivera [Belgrade Univ. (Serbia). Faculty of Mechanical Engineering; Burzic, Meri; Jovicic, Radomir [Belgrade Univ. (Serbia). Innovation Center; Kastelec Macura, Sandra [Technikum Taurunum, Zemun (Serbia); Buyukyildirim, Galip [IWE, Istanbul (Turkey)

2013-01-15

In this paper, the AlMg4.5Mn alloy has been welded using tungsten inert gas (TIG) and metal inert gas (MIG) processes with different gas shielding atmospheres. Tensile strength, hardness, impact and fracture toughness, fatigue crack growth parameters ({Delta}K{sub th}, da/dN), as well as microstructure were determined. By comparing results for different gas mixtures, the main conclusion for TIG welding was that increased helium content has an important effect on toughness and fatigue crack growth parameters, whereas its effect on other mechanical properties is not significant. On the other hand, in the case of MIG welding, adding helium does not affect mechanical properties, except the fatigue crack growth rate. It was also established that adding nitrogen (TIG welding) and oxygen (MIG welding) reduces toughness and increases crack growth rate, so their use in shielding mixtures is not recommended. (orig.)

Quantitative analysis of nickel in zirconium and zircaloy; Dosage du nickel dans le zirconium et dans le zircaloy

Energy Technology Data Exchange (ETDEWEB)

Rastoix, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1957-07-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [French] On determine colorimetriquenent 10 a 1000 ppm de Ni dans le zirconium et le zircaloy par photo colorimetrie a 440 m{mu} de la dimethylglyoxime nickelique. Le dosage est rapide. Le fer, le cuivre, l'etain, le chrome ne genent pas aux concentrations habituellement rencontrees dans le zirconium et ses alliages. (auteur)

Internal hydriding in irradiated defected Zircaloy fuel rods: A review (LWBR Development Program)

International Nuclear Information System (INIS)

Clayton, J.C.

1987-10-01

Although not a problem in recent commercial power reactors, including the Shippingport Light Water Breeder Reactor, internal hydriding of Zircaloy cladding was a persistent cause of gross cladding failures during the 1960s. It occurred in the fuel rods of water-cooled nuclear power reactors that had a small cladding defect. This report summarizes the experimental findings, causes, mechanisms, and methods of minimizing internal hydriding in defected Zircaloy-clad fuel rods. Irradiation test data on the different types of defected fuel rods, intentionally fabricated defected and in-pile operationally defected rods, are compared. Significant factors affecting internal hydriding in defected Zircaloy-clad fuel rods (defect hole size, internal and external sources of hydrogen, Zircaloy cladding surface properties, nickel alloy contamination of Zircaloy, the effect of heat flux and fluence) are discussed. Pertinent in-pile and out-of-pile test results from Bettis and other laboratories are used as a data base in constructing a qualitative model which explains hydrogen generation and distribution in Zircaloy cladding of defected water-cooled reactor fuel rods. Techniques for minimizing internal hydride failures in Zircaloy-clad fuel rods are evaluated

Tensile behavior of friction stir welded AA 6061-T4 aluminum alloy joints

International Nuclear Information System (INIS)

Heidarzadeh, A.; Khodaverdizadeh, H.; Mahmoudi, A.; Nazari, E.

2012-01-01

Highlights: ► Range of parameters for defect-free friction stir welded AA 6061-T4 was reached. ► A model was developed for predicting UTS and EL of friction stir welded AA 6061-T4. ► The maximum values of UTS and EL of joints were estimated by developed model. ► The optimum values of FSW process parameters were determined. -- Abstract: In this investigation response surface methodology based on a central composite rotatable design with three parameters, five levels and 20 runs, was used to develop a mathematical model predicting the tensile properties of friction stir welded AA 6061-T4 aluminum alloy joints at 95% confidence level. The three welding parameters considered were tool rotational speed, welding speed and axial force. Analysis of variance was applied to validate the predicted model. Microstructural characterization and fractography of joints were examined using optical and scanning electron microscopes. Also, the effects of the welding parameters on tensile properties of friction stir welded joints were analyzed in detail. The results showed that the optimum parameters to get a maximum of tensile strength were 920 rev/min, 78 mm/min and 7.2 kN, where the maximum of tensile elongation was obtained at 1300 rev/min, 60 mm/min and 8 kN.

Deformation and collapse of zircaloy fuel rod cladding into plenum axial gaps

International Nuclear Information System (INIS)

Pfennigwerth, P.L.; Gorscak, D.A.; Selsley, I.A.

1983-01-01

To minimize support structure, blanket and reflector fuel rods of the thoria urania-fueled Light Water Breeder Reactor (LWBR) were designed with non-freestanding Zircaloy-4 cladding. An analytical model was developed to predict deformation of unirradiated cladding into axial gaps of fuel rod plenum regions where it is unsupported. This model uses the ACCEPT finite element computer program to calculate elastic-plastic deformation of cladding due to external pressure. The finite element is 20-node, triquadratic, isoparametric, and 3-dimensional. Its curved surface permits accurate modeling of the tube geometry, including geometric nonuniformities such as circumferential wall thickness variation and initial tube out-of-roundness. Progressive increases in axial gap length due to cladding elongation and fuel stack shrinkage are modeled, as are deformations of fuel pellets and stainless steel support sleeves which bound plenum axial gaps in LWBR type blanket fuel rods. Zircaloy-4 primary and secondary thermal creep representations were developed from uniaxial creep testing of fuel rod tubing. Creep response to multi-axial loading is modeled with a variation of Hill's formulation for anisotropic materials. Coefficients accounting for anisotropic thermal creep in Zircaloy-4 tubes were developed from creep testing of externally pressurized tubes having fixed axial gaps in the range 2.5 cm to 5.7 cm and radial clearances over simulated fuel pellets ranging from zero to 0.089 mm. (orig./RW)

Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Test ESBU-1

International Nuclear Information System (INIS)

Hagen, S.; Malauschek, H.; Peck, S.O.; Wallenfels, K.P.

1983-12-01

This report describes the test conduct and results of the bundle test ESBU-1. The test objective was the investigation of temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of a program of out-of-pile experiments, performed within the framework of the PNS Several Fuel Damage Program. The bundle was composed of a 3x3 array of fuel rod simulators surrounded by a zircaloy shroud which was insulated with a ZrO 2 fiber ceramic wrap. The fuel rod simulators comprised a tungsten heater, UO 2 annular pellets, and zircaloy cladding over a 0.4 m heated length. A steam flow of 1 g/s was inlet to the bundle. The most pronounced temperature escalation was found on the central rod. The initial heatup rate of 2 0 C/s at 1100 0 C increased to approximately 6 0 C/s. The maximum temperature reached was 2250 0 C. The following fast temperature decrease was caused by runoff of molten zircaloy. Molten zircaloy swept down the thin cladding oxide layer formed during heatup. The melt dissolved the surface of the UO 2 pellets and refroze as a coherent lump in the lower part of the bundle. The remaining pellets fragmented during cooldown and formed a powdery layer on the refrozen lump. The lump was sectioned posttest at several elevations: Dissolution of UO 2 by the molten zircaloy, interaction between the melt and previously oxidized zircaloy, and oxidation of the melt had occurred. (orig.) [de

Charpy impact test of oxidized and hydrogenated zircaloy using a thin strip specimen

International Nuclear Information System (INIS)

Otsuka, Teppei; Hashizume, Kenichi; Sugisaki, Masayasu

2004-01-01

The impact properties of an oxidized and a hydrogenated Zircaloy have been studied with an instrumented Charpy machine by using a strip Charpy V-notch specimen (1 mm thick by 4mm wide). Fracture processes such as crack initiation and propagation were examined using load-displacement curves obtained in this study. In the case of the hydrogenated specimen containing preferentially oriented hydrides, an appreciable decrease in the absorbed energy was observed in the crack propagation rather than in the crack initiation. From results of fractographs of the specimen, it was suggested that the reduction of the crack propagation energy of hydrogenated specimen could be attributed to the change of the stress state in the Zircaloy matrix, which was caused by the fracture of hydride in the inner part of specimen. In the case of the specimen oxidized at 973k for 60 min, on which an oxide layer (4 μm in thickness) and oxygen incursion layer (4μm) were formed, the surface layers affected the crack initiation process. The growing oxygen incursion layer, in particular, resulted in the constraint of plastic deformation of the Zircaloy matrix not only in the crack initiation but also in the crack propagation as its thickness increased. (author)

Azimuthally anisotropic hydride lens structures in Zircaloy 4 nuclear fuel cladding: High-resolution neutron radiography imaging and BISON finite element analysis

Science.gov (United States)

Lin, Jun-Li; Zhong, Weicheng; Bilheux, Hassina Z.; Heuser, Brent J.

2017-12-01

High-resolution neutron radiography has been used to image bulk circumferential hydride lens particles in unirradiated Zircaloy 4 tubing cross section specimens. Zircaloy 4 is a common light water nuclear reactor (LWR) fuel cladding; hydrogen pickup, hydride formation, and the concomitant effect on the mechanical response are important for LWR applications. Ring cross section specimens with three hydrogen concentrations (460, 950, and 2830 parts per million by weight) and an as-received reference specimen were imaged. Azimuthally anisotropic hydride lens particles were observed at 950 and 2830 wppm. The BISON finite element analysis nuclear fuel performance code was used to model the system elastic response induced by hydride volumetric dilatation. The compressive hoop stress within the lens structure becomes azimuthally anisotropic at high hydrogen concentrations or high hydride phase fraction. This compressive stress anisotropy matches the observed lens anisotropy, implicating the effect of stress on hydride formation as the cause of the observed lens azimuthal asymmetry. The cause and effect relation between compressive stress and hydride lens anisotropy represents an indirect validation of a key BISON output, the evolved hoop stress associated with hydride formation.

Influence of neutron irradiation on the stability of recipitates in zircaloy: a critical review

International Nuclear Information System (INIS)

Lobo, Raquel M.; Andrade, Arnaldo H. P.

2013-01-01

The realization of RMB enterprise (Brazilian Multipurpose Reactor) will give the country a powerful tool to investigate the behavior materials subjected to irradiation. Among them, zirconium alloys, used as cladding of nuclear fuel in reactors type LWR. It is know that neutron irradiation can affect the stability of precipitates in zircaloys, generating as a result changes in theirs mechanical properties, important application of this alloys. This paper present a critical review of neutron irradiation effects on microstructural stability of zircaloys (2 and 4). (author)

Mechanical properties of irradiated and non-irradiated Zr1%Nb and Zircaloy claddings

International Nuclear Information System (INIS)

Griger, Agnes

2004-01-01

The mechanical properties of irradiated and non-irradiated Zr1%Nb were determined and they were compared with the analogous properties of Zircaloy-4 to establish connections between the evolution of mechanical parameters of Zr1%Nb and Zircaloy-4 cladding materials and the measure of irradiation. Samples were irradiated in the vertical channels of the Budapest Research Reactor for different time periods at 50-65 C temperature. The measure of irradiation (fluent) for different samples was estimated by means of flux measurement and using the effective irradiation time. Post irradiation uniaxial tension tests in transverse direction were carried out on ring specimens. The mechanical parameters of the Zr1%Nb alloy significantly improve due to the effect of irradiation. However, the values of mechanical parameters do not further increase when the fluent increases above 10 20 n/cm 2 . These results are in good accordance with the Russian ones [1]. Contrary to the behaviour of Zr1%Nb alloy, the mechanical parameters of the Zircaloy practically do not change on the effect of irradiation. The originally high values of ultimate tensile strength and yield stress change only slightly with the increasing fluent in the investigated fluent-region. (Author)

Nondestructive characterization of hydrogen concentration in zircaloy cladding tubes with laser ultrasound technique

International Nuclear Information System (INIS)

Yang, Che Hua; Lai, Yu An

2006-01-01

This paper describes a laser ultrasound technique (LUT) for nondestructive characterization of hydrogen concentration (HC) in Zircaloy cladding tubes. With the LUT, guided ultrasonic waves are generated remotely and then propagate in the axial direction of Zircaloy tubes, and finally detected remotely by an optical probe. By measuring the dispersion spectra with the LUT, relations between the dispersion spectra and the HC of the Zircaloy tubes can be established. The LUT is non-contact, capable of remote inspection, and therefore suitable for nondestructive inspection of HC in Zircaloy cladding tubes used in nuclear power plant.

Influence of hydrides orientation on strain, damage and failure of hydrided zircaloy-4; Influence de l'orientation des hydrures sur les modes de deformation, d'endommagement et de rupture du zircaloy-4 hydrure

Energy Technology Data Exchange (ETDEWEB)

Racine, A

2005-09-15

In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300 C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embrittlement depends on many parameters, among which hydrogen content and orientation of hydrides with respect to the applied stress. This investigation is devoted to the influence of the orientation of hydrides with respect to the applied stress on strain, damage and failure mechanisms. Macroscopic and SEM in-situ ring tensile tests are performed on cladding tube material (unirradiated cold worked stress-relieved Zircaloy-4) hydrided with about 200 and 500 wppm hydrogen, and with different main hydrides orientation: either parallel or perpendicular to the circumferential tensile direction. We get the mechanical response of the material as a function of hydride orientation and hydrogen content and we investigate the deformation, damage and failure mechanisms. In both cases, digital image correlation techniques are used to estimate local and global strain distributions. Neither the tensile stress-strain response nor the global and local strain modes are significantly affected by hydrogen content or hydride orientation, but the failure modes are strongly modified. Indeed, only 200 wppm radial hydrides embrittle Zy-4: sample fail in the elastic domain at about 350 MPa before strain bands could develop; whereas in other cases sample reach at least 750 MPa before necking and final failure, in ductile or brittle mode. To model this particular heterogeneous material behavior, a non-coupled damage approach which takes into account the anisotropic distribution of the hydrides is proposed. Its parameters are identified from the macroscopic strain field measurements and a

Thermal isocreep curves obtained during multi-axial creep tests on recrystallized Zircaloy-4 and M5™ alloy

International Nuclear Information System (INIS)

Rautenberg, M.; Poquillon, D.; Pilvin, P.; Grosjean, C.; Cloué, J.M.; Feaugas, X.

2014-01-01

Zirconium alloys are widely used in the nuclear industry. Several components, such as cladding or guide tubes, undergo strong mechanical loading during and after their use inside the pressurized water reactors. The current requirements on higher fuel performances lead to the developing on new Zr based alloys exhibiting better mechanical properties. In this framework, creep behaviors of recrystallized Zircaloy-4 and M5™, have been investigated and then compared. In order to give a better understanding of the thermal creep anisotropy of Zr-based alloys, multi-axial creep tests have been carried out at 673 K. Using a specific device, creep conditions have been set using different values of β = σ zz /σ θθ , σ zz and σ θθ being respectively the axial and hoop creep stresses. Both axial and hoop strains are measured during each test which is carried out until stationary creep is stabilized. The steady-state strain rates are then used to build isocreep curves. Considering the isocreep curves, the M5™ alloy shows a largely improved creep resistance compared to the recrystallized Zircaloy-4, especially for tubes under high hoop loadings (0 < β < 1). The isocreep curves are then compared with simulations performed using two different mechanical models. Model 1 uses a von Mises yield criterion, the model 2 is based on a Hill yield criterion. For both models, a coefficient derived from Norton law is used to assess the stress dependence

Verification of a mechanistic model for the strain rate of zircaloy-4 fuel sheaths during transient heating

International Nuclear Information System (INIS)

Hunt, C.E.L.

1980-10-01

A mechanistic strain rate model for Zircaloy-4, named NIRVANA, was tested against experiments where pressurized fuel sheaths were strained during complex temperature-stress-time histories. The same histories were then examined to determine the spread in calculated strain which may be expected because of variations in dimensions, chemical content and mechanical properties which are allowed in the fuel sheath specifications. It was found that the variations allowed by the specifications could result in a probable spread in the predicted strain of plus or minus a factor of two from the mean value. The experimental results were well within this range. (auth)

Zircaloy nodular corrosion analysis by an image processing technique

International Nuclear Information System (INIS)

Kawashima, Junko; Sato, Kanemitsu; Kuwae, Ryosho; Higashinakagawa, Emiko

1987-01-01

An image processor has been fabricated to examine out-of-pile nodular corrosion for Zircaloy-2 tubings. The covering fraction, which is the percentage of the nodule occupying area on the Zircaloy surface, was measured with the processor. The covering fraction showed a strong correlation with the weight gain at any corrosion time of this experiment. The correlation observed can be explained by a model for the lenticular shape of the nodules. The image processor also gives unfolded pictures of the whole Zircaloy surface. By analyzing the picture, the location of the nodules generated was found to be determined in an early stage of corrosion. New nodules were not produced later, and the nodules only grew larger with time. (orig.)

Hydride reorientation in Zircaloy-4 examined by in situ synchrotron X-ray diffraction

Energy Technology Data Exchange (ETDEWEB)

Weekes, H.E. [Department of Materials, Royal School of Mines, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom); Jones, N.G. [Department of Materials Science and Metallurgy, University of Cambridge, 27 Charles Babbage Road, Cambridge CB3 0FS (United Kingdom); Lindley, T.C. [Department of Materials, Royal School of Mines, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom); Dye, D., E-mail: david.dye@imperial.ac.uk [Department of Materials, Royal School of Mines, Imperial College London, Prince Consort Road, London SW7 2BP (United Kingdom)

2016-09-15

The phenomenon of stress-reorientation has been investigated using in situ X-ray diffraction during the thermomechanical cycling of hydrided Zircaloy-4 tensile specimens. Results have shown that loading along a sample’s transverse direction (TD) leads to a greater degree of hydride reorientation when compared to rolling direction (RD)-aligned samples. The elastic lattice micro-strains associated with radially oriented hydrides have been revealed to be greater than those oriented circumferentially, a consequence of strain accommodation. Evidence of hydride redistribution after cycling, to α-Zr grains oriented in a more favourable orientation when under an applied stress, has also been observed and its behaviour has been found to be highly dependent on the loading axis. Finally, thermomechanical loading across multiple cycles has been shown to reduce the difference in terminal solid solubility of hydrogen during dissolution (TSS{sub D,H}) and precipitation (TSS{sub P,H}).

Refusion of zircaloy scraps by VAR (vacuum arc remelting): preliminary results; Fusao de cavacos de zircaloy por VAR: resultados preliminares

Energy Technology Data Exchange (ETDEWEB)

Pereira, L.A.T.; Mucsi, C.S.; Sato, I.M.; Rossi, J.L.; Martinez, L.G., E-mail: lgallego@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Correa, H.P.S. [Universidade Federal do Mato Grosso do Sul (UFMS), Campo Grande, MS (Brazil); Orlando, M.T.D. [Universidade Federal do Espirito Santo (UFES), Vitoria, ES (Brazil)

2010-07-01

Fuel elements and structural components of the core of PWR nuclear reactors are made in zirconium alloys known as Zircaloy. Machining chips and shavings resulting from the manufacturing of these components can not be discarded as scrap, once these alloys are strategic materials for the nuclear area, have high costs and are not produced in Brazil on an industrial bases and, consequently, are imported for the manufacture of nuclear fuel. The reuse of Zircaloy chips has economic, strategic and environmental aspects. In this work is proposed a process for recycling Zircaloy scraps using a VAR (vacuum arc remelting) furnace in order to obtain ingots suitable for the manufacture of components of the reactors. The ingots obtained are being studied in order to verify the influence of processing on composition and microstructure of the remelted material. In this work are presented preliminary results of the composition of obtained ingots compared to start material and the resulting microstructure. (author)

Elucidating the iodine stress corrosion cracking (SCC) process for zircaloy tubing

International Nuclear Information System (INIS)

Nagai, M.; Shimada, S.; Nishimura, S.; Amano, K.

1984-01-01

Several experimental investigations were made to enhance understanding of the iodine stress corrosion cracking (SCC) process for Zircaloy: (1) oxide penetration process, (2) crack initiation process, and (3) crack propagation process. Concerning the effect of the oxide layer produced by conventional steam-autoclaving, no significant difference was found between results for autoclaved and as-pickled samples. Tests with 15 species of metal iodides revealed that only those metal iodides which react thermodynamically with zirconium to produce zirconium tetraiodide (ZrI 4 ) caused SCC of Zircaloy. Detailed SEM examinations were made on the SCC fracture surface of irradiated specimens. The crack propagation rate was expressed with a da/dt=C Ksup(n) type equation by combining results of tests and calculations with a finite element method. (author)

The influence of hydrogen on the deformation behavior of zircaloy 4

International Nuclear Information System (INIS)

Flanagan, M. E.; Koss, D. A.; Motta, A. T.

2008-01-01

The deformation behavior of Zr based cladding forms a basis for fuel behavior codes and affects failure criteria; as such, it is critical to reactor safety. The present study examines the influence of hydrogen on the uniaxial deformation behavior of hydrided cold worked and stress relieved Zircaloy 4 plate material. Specimens of various orientations (i.e., stress axis aligned with the rolling direction, the transverse direction, or normal to the plate surface direction) were tested in compression at a range of temperatures (25 .deg. , 300 .deg. , and 400 .deg. C), and strain rates (from 10-4/s to 10-1/s). Contrasting the deformation behavior of the material containing ∼45 wt ppm H with that of the material containing ∼420 wt. ppm H shows that increasing H content (a) causes a small decrease in the 0.2% yield stress that is eliminated at 1.0% flow stress, (b) increases the strain hardening in the rolling direction but not in the other orientations, (c) has no effect on the temperature dependence of the strain hardening, and (d) does not affect the strain rate hardening behavior. Increasing H content also has no observable effect on the high degree of plastic anisotropy of this plate material which is manifested in difficult through thickness deformation, resulting in high flow stresses for specimens oriented in the normal to plate surface direction

Superior corrosion resistance properties of TiN-based coatings on Zircaloy tubes in supercritical water

Energy Technology Data Exchange (ETDEWEB)

Khatkhatay, Fauzia [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Jiao, Liang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jian, Jie [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Zhang, Wenrui [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jiao, Zhijie [Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109-2104 (United States); Gan, Jian; Zhang, Hongbin [Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Zhang, Xinghang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Department of Mechanical Engineering, Texas A and M University, College Station, TX 77843-3123 (United States); Wang, Haiyan, E-mail: wangh@ece.tamu.edu [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States)

2014-08-01

Thin films of TiN and Ti{sub 0.35}Al{sub 0.65}N nanocomposite were deposited on polished Zircaloy-4 tubes. After exposure to supercritical water for 48 h, the coated tubes are remarkably intact, while the bare uncoated tube shows severe oxidation and breakaway corrosion. X-ray diffraction patterns, secondary electron images, backscattered electron images, and energy dispersive X-ray spectroscopy data from the tube surfaces and cross-sections show that a protective oxide, formed on the film surface, effectively prevents further oxidation and corrosion to the Zircaloy-4 tubes. This result demonstrates the effectiveness of thin film ceramics as protective coatings under extreme environments.

The influences of deformation velocity and temperature on localized deformation of zircaloy-4 in tensile tests

International Nuclear Information System (INIS)

Boratto, F.J.M.

1973-01-01

A new parameter to describe the necking stability in zircaloy-4 during tensile tests is introduced. The parameter is defined as: s = ∂Ln (dσ/dε)/∂Ln ((1/L)dL/dt) for constant temperature, deformation and history. Measures of stress strain rate sensitivity n, reduction of the area at fracture, and deformation profiles of tensile fracture, are done. A complete description of the curve of non-uniform deformation variation with the temperature, is presented. The results are compared with existing data for pure commercially titanium. The influence of strain rate and history on s and n parameters, in the temperature range from 100-700 0 C). (author) [pt

Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction ESSI-4 ESSI-11

International Nuclear Information System (INIS)

Hagen, S.; Kapulla, H.; Malauscheck, H.; Wallenfels, K.P.; Buescher, B.J.

1985-03-01

The tests had the initial heatup rate as main parameter. The experimental arrangement consisted of a fuel rod simulator (central tungsten heater, UO 2 ring pellets and zircaloy cladding), a zircaloy shroud and the fiber ceramic insulation. A steam flow of ca. 20 g/min was introduced at the lower end of the bundle. A temperature escalation was observed in every test. The maximum cladding surface temperature in the single rod tests never exceeded 2200 0 C. The escalation began in the upper region of the rods and moved down the rods, opposite to the direction of steam flow. For fast initial heatup rates, the runoff of molten zircaloy was a limiting process for the escalation. For slow heatup rates, the formation of a protective oxide layer reduced the reaction rate. The test with less insulation thickness showed a reduction of the escalation. A stronger influence was found for the gap between shroud and insulation. This is caused by convection heat losses to the steam circulating in this gap by natural convection. Removal of the gap between shroud and insulation in essentially the same experimental arrangement produced a faster escalation. The posttest appearance of the fuel rod simulators showed that, at slow heatup rates oxidation of the cladding was complete, and the fuel rod was relatively intact. Conversely, at fast heatup rates, relatively little cladding oxidation with extensive dissolution of the UO 2 pellets and runoff of molten cladding was observed. (orig./HP) [de

Fluorimetric determination of uranium in zirconium and zircaloy alloys

International Nuclear Information System (INIS)

Acosta L, E.

1991-05-01

The objective of this procedure is to determine microquantities of uranium in zirconium and zircaloy alloys. The report also covers the determination of uranium in zirconium alloys and zircaloy in the range from 0.25 to 20 ppm on 1 g of base sample of radioactive material. These limit its can be variable if the size of the used aliquot one is changed for the final determination of uranium. (Author)

Oxidation of Zircaloy-4 in steam-nitrogen mixtures at 600-1200 °C

Science.gov (United States)

Steinbrueck, Martin; da Silva, Fabio Oliveira; Grosse, Mirco

2017-07-01

High-temperature oxidation of zirconium alloys in steam-nitrogen atmospheres may be relevant during various nuclear accident scenarios. Therefore, isothermal oxidation tests with Zircaloy-4 in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200 °C using thermogravimetry. The gas compositions were varied between 0 and 100 vol% nitrogen including 0.1 and 90 vol%. The strong effect of nitrogen on the oxidation kinetics of zirconium alloys was confirmed in these tests in mixed steam-nitrogen atmospheres. Even very low concentrations of nitrogen (starting from less than 1 vol%) strongly increase reaction kinetics. Nitrogen reduces transition time from protective to non-protective oxide scale (breakaway). The formation of zirconium nitride, ZrN, and its re-oxidation is the main reason for the highly porous oxide scales after transition. The results of this study have shown the safety relevant role of nitrogen during severe accidents and, more generally, suggest the need of using well controlled gas atmospheres for experiments on oxidation of zirconium alloys.

Growth and microstructure formation of isothermally-solidified Zircaloy-4 joints brazed by a Zr-Ti-Cu-Ni amorphous alloy ribbon

Science.gov (United States)

Kim, K. H.; Lim, C. H.; Lee, J. G.; Lee, M. K.; Rhee, C. K.

2013-10-01

The microstructure and growth characteristics of Zircaloy-4 joints brazed by a Zr48Ti16Cu17Ni19 (at.%) amorphous filler metal have been investigated with regard to the controlled isothermal solidification and intermetallic formation. Two typical joints were produced depending on the isothermal brazing temperature: (1) a dendritic growth structure including bulky segregation in the central zone (at 850 °C), and (2) a homogeneous dendritic structure throughout the joint without segregation (at 890 °C). The primary α-Zr phase was solidified isothermally, nucleating to grow into a joint with a cellular or dendritic structure. Also, the continuous Zr2Ni and particulate Zr2Cu phases were formed in the segregated center zone and at the intercellular region, respectively, owing to the different solubility and atomic mobility of the solute elements (Ti, Cu, and Ni) in the α-Zr matrix. A disappearance of the central Zr2Ni phase was also rate-controlled by the outward diffusion of the Cu and Ni elements. When the detrimental Zr2Ni intermetallic phase was eliminated by a complete isothermal solidification at 890 °C, the strengths of the joints were high enough to cause yielding and fracture in the base metal, exceeding those of the bulk Zircaloy-4, at room temperature as well as at elevated temperatures (up to 400 °C).

Influence of impurities on the ignition, combustion and explosion properties of Zircaloy filings

International Nuclear Information System (INIS)

Muenzel, H.; Praetorius, R.

1990-11-01

The influence of solid substances (e.g. UO 2 , MoO 3 , KNO 3 ) and liquids (e.g. water, nitric acid) on the behavior of Zircaloy filings was investigated. The addition of solid substances as well as liquids increases the ignition temperature. Samples with more than 50% water cannot be ignited (except with KCl solutions). With solid impurities added two modes of combustion are observed with propagation velocities of about 1 and >4 cm/s, respectively. The velocity depends on the heat capacity of the sample. In the presence of water the velocity increases by about two orders of magnitude. The maximum pressure observed in dust explosions in the presence of solid impurities depends on the heat capacity and the amount of Zircaloy burnt but not on the chemical properties of the added substances. The maximum pressure can be higher than 20 bar if water or nitric acid are added. With the proposed models and few additional experimental measurements it is possible to predict the behavior of other Zircaloy filings. (orig.) With 32 refs., 20 tabs., 91 figs [de

Effect of hydrides on the ductile-brittle transition in stress-relieved, recrystallized and beta-treated zircaloy-4

International Nuclear Information System (INIS)

Pelchat, J.; Barcelo, F.

1991-01-01

This paper is concerned with the influence of δ-hydrides on the mechanical properties of three heat treated cold-rolled Zircaloy-4 sheets (stress-relieved, recrystallized and β treated), tested at room temperature and 350 0 C. Smooth tensile specimens of two thicknesses: 0.5 and 3.1 mm, containing different hydride volume fractions, up to 18% (about 1400 ppm H), have been tested. Metallographic and fractographic analysis were carried out in order to examine the fracture morphology near and on the fracture surface, and to determine the evolution of the fracture mechanism of hydrides as a function of temperature, hydride orientation and volume fraction

Variation in the strain anisotropy of Zircaloy with temperature and strain

International Nuclear Information System (INIS)

Hindle, E.D.; Worswick, D.

1984-04-01

Strain anisotropy was investigated at temperatures in the range 293 to 1117K in circular tensile specimens prepared from rolled Zircaloy-2 plate so that their tensile axes were parallel to and transverse to the rolling direction. The strain anisotropy factor for both types of specimen increased markedly in the high alpha phase region above 923K reaching a maximum at circa 1070K. Above this temperature in the alpha-plus-beta phase region the strain anisotropy decreased rapidly as the proportion of beta phase increased and was almost non-existent at 1173K. The strain anisotropy was markedly strain dependent, particularly in the high alpha phase region. The study was extended to Zircaloy-4 pressurized water reactor (PWR) 17 x 17 type fuel rod tubing specimens which were strained under biaxial conditions using cooling conditions which promoted uniform diametral strain over most of their lengths (circa 250 mm). In these circumstances the strain anisotropy is manifest by a reduction in length. Measurement of this change along with that in diameter and wall thickness produced data from which the strain anisotropy factor was calculated. The results, although influenced by additional factors discussed in the paper, were similar to those observed in the uniaxial Zircaloy-2 tensile tests. (author)

Variation in the strain anisotropy of Zircaloy with temperature and strain

International Nuclear Information System (INIS)

Hindle, E.D.; Worswick, D.

1984-01-01

The strong crystallographic texture which is developed during the fabrication of zirconium-based alloys causes pronounced anisotropy in their mechanical properties, particularly deformation. The tendency for circular-section tension specimens with a high concentration of basal poles in one direction to become elliptical when deformed in tension has been used in this study to provide quantitative data on the effects of both strain and temperature on strain anisotropy. Tension tests were carried out over a temperature range of 293 to 1193 K on specimens machined from Zircaloy-2 plate. The strain anisotropy was found to increase markedly at temperatures over 923 K, reaching a maximum in the region of 1070 K. The strain anisotropy increased with increasing strain in this temperature region. The study was extended to Zircaloy-4 pressurized-water reactor fuel cladding by carrying out tube swelling tests and evaluating the axial deformation produced. Although scatter in the test results was higher than that exhibited in the tension tests, the general trend in the data was similar. The effects of the strain anisotropy observed are discussed in relation to the effects of temperature on the ductility of Zircaloy fuel cladding tubes during postulated largebreak loss-of-coolant accidents

The significance of calcified fibrocartilage on the cortical endplate of the translational sheep spine model.

Science.gov (United States)

Sinclair, Sarina K; Bell, Spencer; Epperson, Richard Tyler; Bloebaum, Roy D

2013-05-01

To gain an understanding of the vertebral cortical endplate and factors that may affect the ability to achieve skeletal attachment to intervertebral implants and fusion, this study aimed to characterize the hypermineralized tissue on the cortical endplate of the vertebral body on a commonly used animal model. Skeletally mature sheep were injected with tetracycline prior to euthanasia and the C2-C3, T5-T6, and L2-L3 spinal motion segments were excised and prepared. Vertebral tissues were imaged using backscatter electron (BSE) imaging, histology, and tetracycline labeling was used to assess bone remodeling within different tissue layers. It was determined that the hypermineralized tissue layer was calcified fibrocartilage (CFC). No tetracycline labels were identified in the CFC layer, in contrast to single and double labels that were present in the underlying bone, indicating the CFC present on the cortical endplate was not being actively remodeled. The average thickness of the CFC layer was 146.3 ± 70.53 µm in the cervical region, 98.2 ± 40.29 µm in the thoracic region, and 150.89 ± 69.25 µm in the lumbar region. This difference in thickness may be attributed to the regional biomechanical properties of the spine. Results from this investigation indicate the presence of a nonremodeling tissue on the cortical endplate of the vertebral body in sheep spines, which attaches the intervertebral disc to the vertebrae. This tissue, if not removed, would likely prevent successful bony attachment to an intervertebral device in spinal fusion studies and total disc replacement surgeries. Copyright © 2013 Wiley Periodicals, Inc.

Microstructural Evolution in Friction Stir Welding of Ti-6Al-4V

Science.gov (United States)

Rubisoff, H.; Querin, J.; Magee, D.; Schneider, J.

2008-01-01

Friction stir welding (FSW) is a thermo-mechanical process that utilizes a nonconsumable rotating pin tool to consolidate a weld joint. In the conventional FSW process, the pin tool is responsible for generating both the heat required to soften the material and the forces necessary to deform and combine the weld seam. As such, the geometry of the pin tool is important to the quality of the weld and the process parameters required to produce the weld. Because the geometry of the pin tool is limitless, a reduced set of pin tools was formed to systematically study their effect on the weldment with respect to mechanical properties and resultant microstructure. In this study 0deg, 15deg, 30deg, 45deg, and 60deg tapered, microwave sintered, tungsten carbide (WC) pin tools were used to FSW Ti-6Al-4V. Transverse sections of the weld were used to test for mechanical properties and to document the microstructure using optical microscopy. X-ray diffraction (XRD) was also used to characterize the microstructure in the welds. FSW results for the 45deg and 60deg pin tools are reported in this paper.

Oxidation-induced deformation of zircaloy-4 tubing in steam in the temperature range 600-1000 degree C

International Nuclear Information System (INIS)

Aly, A.E.; Hussein, A.G.; EL-Raghy, S.M.; EL-Sayed, A.A.; EL-Banna, O.A.

1992-01-01

The oxidation-induced deformation of zircaloy-4 (zry-4) tubing in steam has been studied in the temperature range 600 to 1000 degree C. The induced deformation has been measured in both radial and axial directions of the tube. The effect of hydrogen addition to steam was also investigated. The oxidation-induced deformation has been characterized by uniform and non-uniform (distortion) strain period. During the uniform strain period the radial strain kinetics were found in general, to be parallel to the oxidation kinetics. The axial strain (δA) induced by oxidation was found to be always lower than the radial strain (εR). The addition of 5% by volume hydrogen to steam leads to an increase in the oxidation rate and to a decrease in the degree of anisotropy between radial and axial strains

Correlation of Weld Appearance with Microstructure and Mechanical Properties of 2024-T4 Aluminum Alloy Welded by Fiber Laser with Filler Wire

Directory of Open Access Journals (Sweden)

XU Fei

2017-11-01

Full Text Available Two typical cross-section of welds, including nail shape and near X shape, are obtained in the process of fiber laser welding 2024-T4 Al alloy with filler wire. The correlations of the two weld appearances and other elements (such as microstructure, microhardness, and joint's tensile properties were analyzed. The results show that the weld with near X shape cross-section during the welding process is more stable than that with nail shape cross-section, and the welding spatter of the former is smaller than that of the latter. The microstructure of the weld zone is columnar grains and equiaxed grains, the columnar grains are formed near the fusion line and growing along the vertical direction of the fusion line, the equiaxed grains are distributed in the center of the weld zone. The secondary dendrite of the grains in the center of the weld with nail shape cross-section grows better, and gradually forms to equiaxed dendrite, while the grains size of the weld with near X shape cross-section is relatively finer, exhibiting equiaxed cellular grain. Compared with the joint with nail shape cross-section of the weld, the joint with near X shape cross-section of the weld have some different characteristics, the precipitation strengthening phase θ(Al2Cu content in weld zone of the latter is more than that of the former, the average microhardness value of the weld zone of the latter is higher than that of the former, the softening phenomenon of heat affect zone (HAZ of the latter is weaker than that of the former, and the joint's tensile strength and plasticity of the latter are lower than that of the former slightly.

Hydride phase dissolution enthalpy in neutron irradiated Zircaloy-4

International Nuclear Information System (INIS)

Vizcaino, Pablo; Banchik, Abraham D.

2003-01-01

The differential calorimetric technique has been applied to measure the dissolution enthalpy, ΔH irrad δ→α , of zirconium hydrides precipitated in structural components removed from the Argentine Atucha 1 PHWR nuclear power plant after 10.3 EFPY. An average value of ΔH irrad δ→α = 5 kJ/mol H was obtained after the first calorimetric run. That value is seven times lower than the value of ΔH δ→α = 37.7 kJ/mol H recently determined in Zircaloy-4 specimens taken from similar unirradiated structural components using the same calorimetric technique, [1]. Post-irradiation thermal treatments gradually increase that low value towards the unirradiated value with increasing annealing temperature similar to that observed for TSSd irrad . Therefore the same H atom trapping mechanism during reactor operation already proposed to explain the evolution of TSSd irrad is also valid for Q irrad δ→α . As the ratio Q/ΔH is proportional to the number N H of H atoms precipitated as hydrides, the increment of Q irrad δ→α with the thermal treatment indicates that the value of N H also grows with the annealing reaching the value corresponding to the bulk H concentration when ΔH irrad δ→α ≅ 37 kJ/mol H. That is a direct indication that the post-irradiation thermal treatment releases the H atoms from their traps increasing the number of H atoms available to precipitate at the end of each calorimetric run and/or isothermal treatment. (author)

Superficial characterization by XP S of silver nanoparticles and their hydrothermal deposit over zircaloy; Caracterizacion superficial por XPS de nanoparticulas de plata y su deposito hidrotermal sobre zircaloy

Energy Technology Data Exchange (ETDEWEB)

Contreras R, A.; Gutierrez W, C.; Martinez M, I.; Medina A, A. L., E-mail: aida.contreras@inin.gob.mx [ININ, Departamento de Tecnologia de Materiales, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2012-10-15

The analysis technique of X-ray photoelectron spectroscopy (XP S) is sensitive exclusively to the first layers of the solids surface, which allows obtaining information about the chemical, physical and electronic properties of them. The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy pipes that contain in their interior pellets or uranium dioxide. In this work is studied the zircaloy surface, oxidized zircaloy under similar conditions to those of a reactor BWR type and oxidized zircaloy with a hydrothermal deposit of silver nanoparticles and zinc. The silver deposit is a proposal of the Materials Technology Department of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, which has the same objective that the noble metals deposit (Pt, Pd, and Rh) that is practiced in some of the reactors BWR, in order to mitigating the speed of crack growth for IGSCC in stainless steels 304 Ss. (Author)

Development of various welding techniques for refractory and reactive metals and alloys

International Nuclear Information System (INIS)

Tonpe, Sunil; Saibaba, N.

2016-01-01

Nuclear Fuel Complex (NFC), Hyderabad, India with its excellent manufacturing facilities, produces nuclear fuel and structural components for nuclear reactors. NFC has taken up the challenging job of production of various critical components made out of refractory and reactive metals and alloys for nuclear and aerospace applications as an indigenization import substitute program. Refractory metals are prime candidates for many high temperature aerospace components because of refractory metal's high melting points and inherent creep resistance. The use of refractory metals is often limited because of their poor room temperature properties, inadequate oxidation resistance at elevated temperatures, difficulties associated with joining or welding etc. These advanced materials demand stringent requirement with respect to chemistry, dimensional tolerances, mechanical and metallurgical properties. This paper discusses in detail various welding techniques adopted in NFC for refractory and reactive metals and alloys such as Nb, Zr, Ti, Ta, Zircaloy, Titanium-half alloy etc. to manufacture various components and assemblies required for nuclear and aerospace applications

Effect of hydriding temperature and strain rate on the ductile-brittle transition in β treated Zircaloy-4

International Nuclear Information System (INIS)

Bai, J.B.

1996-01-01

In this paper, the effect of hydriding temperature and strain rate on the ductile-brittle transition in β treated Zircaloy-4 has been investigated. The hydriding temperature used is 700degC, strain rates being 4x10 -4 s -1 and 4x10 -3 s -1 . The results show that at same conditions the ductility of hydrides decreases as the hydriding temperature decreases. There exists a critical temperature (transition temperature) of 250degC for hydriding at 700degC, below which the hydrided specimens (and so for the hydrides) are brittle, while above it they are ductile. This transition temperature is lower than the one mentioned by various authors obtained for hydriding at 400degC. For the same hydriding temperature of 700degC, the specimens tested at 4x10 -3 s -1 are less ductile than those tested at 4x10 -4 s -1 . Furthermore, unlike at a strain rate of 4x10 -4 s -1 , there is no more a clear ductile-brittle transition behaviour. (author)

A tem investigation on intermetallic particles in zircaloy-2

International Nuclear Information System (INIS)

Sudarminto, Harini Sosiati; Kuwano, Noriyuki; Oki, Kensuke

1996-01-01

Tem investigation were conducted on the heat treated zircaloy-2 having the composition of Zr containing 1.6% Sn, 0.2% Fe, 0.1% Cr and 0.05% Ni (%wt) in order tostudy the characteristics of intermetallic particles related to the microstructural basis on the corrosion effect. Forged zircaloy-2 was annealed in the β-phase at 1050 C degrees for various isothermally in the α-phase region at 650 and 750 C degrees, followed by water quenching. The size precipates, the lower became their number. By increasing the annealing temperature, the growth of precipitates formed in this zircaloy-2 were of the Zr(Cr,Fe) 2 and Zr 2 (Fe,Cr,Ni) types. These kinds of precipitates and the ratios of Fe/Cr were independent of size and shape of precipitates and annealing time and temperature. (author), 16 refs, 2 tabs, 5 figs

Intra- and interobserver reproducibility of vertebral endplate signal (modic) changes in the lumbar spine

DEFF Research Database (Denmark)

Jensen, Tue Secher; Sorensen, Joan Solgaard; Kjær, Per

2007-01-01

. PURPOSE: To assess the intra- and interobserver reliability of the "Nordic Modic classification" protocol. MATERIAL AND METHODS: MRI scans of 50 individuals representative of the general Danish population aged 40 were evaluated by two observers. Criteria for grading the changes were developed...... by the Nordic Modic Consensus Group. After consensus was established, all 50 MRI examinations were evaluated independently by each observer. Intraobserver reliability was assessed by re-evaluation of the 50 examinations by one of the observers. Kappa statistics were used to calculate agreement. RESULTS: Intra....... CONCLUSION: In this study, we found convincing reproducibility of a detailed evaluation protocol of vertebral endplate signal changes, the "Nordic Modic Classification." The authors recommend that the evaluation protocol should be used in future studies investigating vertebral endplate signal changes....

Association of endothelin-1 expression and cartilaginous endplate degeneration in humans.

Directory of Open Access Journals (Sweden)

Wei Yuan

Full Text Available BACKGROUND: Inflammatory cytokines are involved in intervertebral disc (IVD degeneration. Endothelin-1 (ET-1, a 21-amino-acid cytokine implicated with cartilage degradation, is secreted by vascular endothelial cells and also by many other cell types. The expression of ET-1 in human IVD cartilage endplate (CEP and its role in disc degeneration have not been explored. METHODS AND FINDINGS: The expression of ET-1 in degenerated CEP was analyzed by immunohistochemical staining and Western blotting; ET-1 was demonstrated in cartilaginous endplate cells (CECs by immunofluorescent staining. The ET-1 mRNA expression and protein production by CECs stimulated by tumor necrosis factor alpha (TNF-α, a pro-inflammatory cytokine, were determined by real-time PCR analysis and Western blotting, respectively. The matrix metalloprotease-1 (MMP-1, MMP-13 and tissue inhibitor of metalloproteases-1 (TIMP-1 levels in the supernatant of cultured CECs treated with ET-1 were determined using enzyme-linked immunosorbent assays. Nitric oxide (NO release and nitric oxide synthase (NOS activity were measured using a spectrophotometric assay. The apoptosis of CECs by ET-1 was measured by an Annexin V-FITC detection assay. The production of ET-1 in degenerated cartilage endplate was significantly higher than normal CEP. The results showed that ET-1 was expressed by CECs and modulated by TNF-α in a dose-dependent manner. ET-1 increased production of MMP-1 and MMP-13, decreased TIMP-1 production, and induced NO and NOS release by cultured CECs. The direct stimulation of CECs by ET-1 did not promote cell apoptosis. CONCLUSION: The study results suggest that ET-1 played a pivotal role in human CEP degeneration, and may be a new target for development of therapies for this condition.

Influence of Welding Process and Post Weld Heat Treatment on Microstructure and Pitting Corrosion Behavior of Dissimilar Aluminium Alloy Welds

Science.gov (United States)

Venkata Ramana, V. S. N.; Mohammed, Raffi; Madhusudhan Reddy, G.; Srinivasa Rao, K.

2018-03-01

Welding of dissimilar Aluminum alloy welds is becoming important in aerospace, shipbuilding and defence applications. In the present work, an attempt has been made to weld dissimilar aluminium alloys using conventional gas tungsten arc welding (GTAW) and friction stir welding (FSW) processes. An attempt was also made to study the effect of post weld heat treatment (T4 condition) on microstructure and pitting corrosion behaviour of these welds. Results of the present investigation established the differences in microstructures of the base metals in T4 condition and in annealed conditions. It is evident that the thickness of the PMZ is relatively more on AA2014 side than that of AA6061 side. In FS welds, lamellar like shear bands are well noticed on the top of the stir zone. The concentration profile of dissimilar friction stir weld in T4 condition revealed that no diffusion has taken place at the interface. Poor Hardness is observed in all regions of FS welds compared to that of GTA welds. Pitting corrosion resistance of the dissimilar FS welds in all regions was improved by post weld heat treatment.

Amorphization of Laves-Phase Precipitates in Zircaloy-4 by Neutron Irradiation

Energy Technology Data Exchange (ETDEWEB)

Peters, H.R.; Taylor, D.F.; Yang, Walter J.S.

1999-04-23

Examination of corrosion coupons by transmission electron microscopy after their exposure in the Idaho Advanced Test Reactor (ATR) has broadened the Zircaloy-4 precipitate-amorphization database and validated a new kinetic model for previously unavailable values of temperature and fast-neutron flux. The model describes the amorphization of Zr(Fe,Cr){sub 2} intermetallic precipitates in zirconium alloys as a dynamic competition between radiation damage and thermal annealing that leaves some iron atoms available for flux-assisted diffusion to the zirconium matrix. It predicts the width of the amorphous zone as a function of neutron flux (E>1 MeV), temperature, and time. In its simplest form, the model treats the crystalline/amorphous and precipitate/matrix interfaces as parallel planes, and its accuracy decreases for small precipitates and high fluence as the amorphous-zone width approaches precipitate dimensions. The simplest form of the model also considers diffusion to be rate-determining. This is an accurate approximation for steady-state conditions or slow changes in flux and temperature, but inappropriate for the analysis of faster transients. The paper addresses several difficulties inherent in measuring amorphous-zone width, and utilizes the expanded database to evaluate the improvements in predictive accuracy available through both conversion of the model to spherical coordinates and extension of its time dependency.

Predictors of new vertebral endplate signal (Modic) changes in the general population

DEFF Research Database (Denmark)

Jensen, Tue Secher; Kjaer, Per; Korsholm, Lars

2010-01-01

predictors of new VESC were female gender, disc-related MRI findings (disc degeneration, disc bulges, disc herniation, and other endplate changes) and lifestyle factors [high physical work or leisure activity, high body mass index (BMI), and heavy smoking]. Bivariate and multivariate logistic regressions...... with disc degeneration, bulges or herniations at 40 were the only predictors of new VESC at age 44. Therefore, the development of new VESC at the age of 44 appears to be based on the status and dynamics of the disc, rather than being the result of gender or lifestyle factors such as smoking and physical...... were used to identify predictors of new VESC. New VESC at the age of 44 appeared in 67 of the 344. The majority (84%) of these new signal changes were type 1 VESC and almost half (45%) were only in the endplate and did not extend into the vertebral body. In the multivariate analysis, lumbar disc levels...

Zircaloy oxidation and cladding deformation in PWR-specific CORA experiments

International Nuclear Information System (INIS)

Minato, K.; Hering, W.; Hagen, S.

1991-07-01

Out-of-pile bundle experiments (zircaloy 4) are performed in the CORA facility to investigate the behavior of PWR fuel elements during severe fuel damage (SFD) accidents. Within the international cooperation the most significant phenomena such as cladding deformation, oxidation (especially the zirconium/steam reaction), melt formation, melt release, and relocation which were found in all tests have been analyzed. (orig./MM) [de

Effect of an Excess of Iron and Hydriding on the Metallurgical Properties of Zircaloy-4

International Nuclear Information System (INIS)

Ghilarducci, Ada; Ramos, R; Martin, E; Peretti, Hernan; Corso Hugo

2000-01-01

Results are presented of mechanical properties and microstructure morphologies obtained in samples of zircaloy-4 of modified composition by an excess in the content of alloying elements as well as hydrides.This work is focused mainly on the effect of 1000 wt. ppm additional Fe as compared to the standard composition alloy.The study is carried out by means of tensile tests at room temperature and at 240 0 C, by hardness tests, by SEM observations and EDS microanalysis.The results indicate that precipitates concentrate along grain boundaries in all cases, and that for higher contents of alloying elements corresponds a higher quantity of precipitates and smaller grain sizes.Except for the hydrided sample, the fracture is ductile with cavities nucleated at precipitates. Finally it is concluded that an increase in Fe content affect the mechanical properties

Growth and microstructure formation of isothermally-solidified Zircaloy-4 joints brazed by a Zr–Ti–Cu–Ni amorphous alloy ribbon

Energy Technology Data Exchange (ETDEWEB)

Kim, K.H. [University of Science and Technology, Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of); Lim, C.H. [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of); Lee, J.G., E-mail: jglee88@kaeri.re.kr [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of); Lee, M.K.; Rhee, C.K. [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of)

2013-10-15

The microstructure and growth characteristics of Zircaloy-4 joints brazed by a Zr{sub 48}Ti{sub 16}Cu{sub 17}Ni{sub 19} (at.%) amorphous filler metal have been investigated with regard to the controlled isothermal solidification and intermetallic formation. Two typical joints were produced depending on the isothermal brazing temperature: (1) a dendritic growth structure including bulky segregation in the central zone (at 850 °C), and (2) a homogeneous dendritic structure throughout the joint without segregation (at 890 °C). The primary α-Zr phase was solidified isothermally, nucleating to grow into a joint with a cellular or dendritic structure. Also, the continuous Zr{sub 2}Ni and particulate Zr{sub 2}Cu phases were formed in the segregated center zone and at the intercellular region, respectively, owing to the different solubility and atomic mobility of the solute elements (Ti, Cu, and Ni) in the α-Zr matrix. A disappearance of the central Zr{sub 2}Ni phase was also rate-controlled by the outward diffusion of the Cu and Ni elements. When the detrimental Zr{sub 2}Ni intermetallic phase was eliminated by a complete isothermal solidification at 890 °C, the strengths of the joints were high enough to cause yielding and fracture in the base metal, exceeding those of the bulk Zircaloy-4, at room temperature as well as at elevated temperatures (up to 400 °C)

The effect of heat treatments on the corrosion behavior of Zircaloy-4

International Nuclear Information System (INIS)

Zhou Bangxin; Zhao Wenjin; Miao Zhi; Pan Shufang; Li Cong; Jiang Yourong

1996-06-01

The effect of penultimate annealing temperature and cooling rate on the corrosion behavior of Zircaloy-4 cladding tube has been investigated. Both nodular corrosion and uniform corrosion resistance can be improved obviously after changing the heat treatment from the original annealing at 650 degree C to quenching from 830 degree C (upper temperature of alpha phase region or lower temperature of beta phase region). Although the nodular corrosion resistance can be improved obviously after quenching from beta phase, there was a second transition in the variation between weight gain and exposure time, which shows a poor uniform corrosion resistance after a long exposure time during the autoclave tests. The main factor of affecting corrosion behavior is the solid solution contents of Fe and Cr in alpha zirconium rather than the size of second phase particles. About 200 μg/g Fe and Cr super saturated solid solution in alpha zirconium could get good uniform and nodular corrosion resistance, but much more solid solution contents of Fe and Cr in alpha zirconium could bring about a trend toward poor uniform corrosion resistance for long-term exposure time. (14 refs., 10 figs., 1 tab.)

Dynamic thermo-chemo-mechanical strain of Zircaloy-4 slotted rings for evaluating strategies that mitigate stress corrosion cracking

Energy Technology Data Exchange (ETDEWEB)

Ferrier, G.A.; Metzler, J.; Farahani, M.; Chan, P.K.; Corcoran, E.C. [Royal Military College of Canada, Kingston, ON (Canada)

2014-07-01

Stress corrosion cracking (SCC) in Zircaloy-4 fuel sheaths has been investigated by static loading of slotted ring samples under hot and corrosive conditions. However, in nuclear reactors, power ramps can have short (e.g., 10-20 minutes) and recurring time frames due to dynamic processes such as on-power refuelling, adjuster rod manoeuvres, and load following. Therefore, to enable out-reactor dynamic testing, an apparatus was designed to dynamically strain slotted ring samples under SCC conditions. This apparatus can additionally be used to test fatigue properties. Unique capabilities of this apparatus and preliminary results obtained from static and dynamic tests are presented. (author)

Oxidation of Zircaloy-4 in steam-nitrogen mixtures at 600–1200 °C

Energy Technology Data Exchange (ETDEWEB)

Steinbrueck, Martin, E-mail: martin.steinbrueck@kit.edu; Oliveira da Silva, Fabio; Grosse, Mirco

2017-07-15

High-temperature oxidation of zirconium alloys in steam-nitrogen atmospheres may be relevant during various nuclear accident scenarios. Therefore, isothermal oxidation tests with Zircaloy-4 in steam-nitrogen mixtures have been performed at 600, 800, 1000, and 1200 °C using thermogravimetry. The gas compositions were varied between 0 and 100 vol% nitrogen including 0.1 and 90 vol%. The strong effect of nitrogen on the oxidation kinetics of zirconium alloys was confirmed in these tests in mixed steam-nitrogen atmospheres. Even very low concentrations of nitrogen (starting from less than 1 vol%) strongly increase reaction kinetics. Nitrogen reduces transition time from protective to non-protective oxide scale (breakaway). The formation of zirconium nitride, ZrN, and its re-oxidation is the main reason for the highly porous oxide scales after transition. The results of this study have shown the safety relevant role of nitrogen during severe accidents and, more generally, suggest the need of using well controlled gas atmospheres for experiments on oxidation of zirconium alloys.

Influence of hydride microstructure on through-thickness crack growth in zircaloy-4 sheet

International Nuclear Information System (INIS)

Raynaud, P.A.; Meholic, M.J.; Koss, D.A.; Motta, A.T.; Chan, K.S.

2007-01-01

The fracture toughness of cold-worked and stress-relieved Zircaloy-4 sheet subject to through-thickness crack growth within a 'sunburst' hydride microstructure was determined at 25 o C. The results were obtained utilizing a novel testing procedure in which a narrow linear strip of hydride blister was fractured at small loads under bending to create a well-defined sharp pre-crack that arrested at the blister-substrate interface. The hydriding procedure also forms 'sunburst' hydrides emanating from the blister that were aligned both in the plane of the crack and in the crack growth direction. Subsequent tensile loading caused crack growth initiation into the field of 'sunburst' hydrides. Specimen failure occurred under near-linear elastic behavior, and the fracture toughness for crack growth initiation into sunburst hydrides was in the range K Q ∼10-15 MPa√m. These results, when combined with those of a previous study, indicate that the through-thickness crack growth initiation toughness at 25 o C is very sensitive to the hydride microstructure. (author)

Ti-6Al-4V electron beam weld qualification using laser scanning confocal microscopy

International Nuclear Information System (INIS)

Wanjara, P.; Brochu, M.; Jahazi, M.

2005-01-01

Processing conditions for manufacturing Ti-6Al-4V components by welding using an electron beam source are known to influence the transformation microstructure in the narrow fusion and heat-affected zones of the weld region. This work examined the effect of multiple-sequence welding on the characteristics of the transformed beta microstructure, using laser scanning confocal microscopy to resolve the Widmanstaetten alpha-beta structure in the fusion zone. The evolution in the alpha interlamellar spacing and plate thickness with processing was then related to microhardness measurements in the weld region

Implications of Y-fluting microstructures in zircaloy stress-corrosion fracture and analogous systems

International Nuclear Information System (INIS)

Banks, T.M.; Garlick, A.

1982-01-01

Transgranular cleavage is an important mode of crack propagation during stress-corrosion cracking (SCC) of Zircaloy in iodine vapour; and another characteristic feature is the presence of parallel closely spaced ridges. These are often referred to as Y-flutings because each ridge takes the form of an inverted Y when viewed along the direction of crack growth. The flutings are shown here to be formed by localised ductile parting of the Zircaloy near the tips of cleavage cracks; high mechanical constraints in those regions and the limited number of available slip systems result in the formation of a planar array of parallel tunnels. Upon final separation these appear as a pattern of parallel ridges on each fracture face. Striking similarities in morphology have been noted here between Y-flutings in Zircaloy and those produced during tests on unstable fluid interfaces: the direction of motion of the fluid interface can be determined from the Y-morphology and is in agreement with observations from Zircaloy SCC tests. It is further demonstrated that equations governing thermodynamic and kinetic instability of fluid interfaces can be adapted to relate the fluting spacing in Zircaloy to standard fracture mechanics parameters. (author)

Plasma Processes of Cutting and Welding

Science.gov (United States)

1976-02-01

TIG process. 2.2.2 Keyhole Welding In plasma arc welding , the term...Cutting 3 3 4 4 4 2.2 Plasma Arc Welding 5 2.2.1 Needle Arc Welding 2.2.2 Keyhole Welding 5 6 3. Applications 8 93.1 Economics 4. Environmental Aspects of...Arc Lengths III. Needle Arc Welding Conditions IV. Keyhole Welding Conditions v. Chemical Analyses of Plates Used - vii - 1. 2. 3. 4. 5. 6. 7. 8.

Development of processes for zircaloy chips recycling by electric arc furnace remelting and powder metallurgy

International Nuclear Information System (INIS)

Pereira, Luiz Alberto Tavares

2014-01-01

PWR reactors employ, as nuclear fuel, UO 2 pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopy. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering. (author)

The post irradiation examination of fuel in support of Bruce A nuclear division fueling with flow program

International Nuclear Information System (INIS)

Montin, J.; Sagat, S.; Day, R.; Novak, J.; Bromfield, H.

1995-01-01

Bruce A Nuclear Division (BAND) units are operating at ∼ 75% of full power, because of the potential of a power pulse in the event of an inlet header break. As a result, BAND is converting to fueling with flow, to eliminate the potential of a power pulse and to allow for full-power operation. Concerns regarding the integrity of the end-of-life (EOL) bundles interacting with the latch at the downstream end of the fuel channel were raised. BAND carried out a test program in which EOL bundles in the upstream position 13 of Unit 2 were cascaded into the downstream latch position 1 of another channel. Six of twelve cascaded bundles and two typical EOL position 13 (benchmark) bundles were selected for post-irradiation examination (PIE). Incipient cracks were found in the assembly welds (endplateto-endcap welds) of all six cascaded bundles. No incipient cracks were found in the benchmark bundles. Metallographic and fractographic examination, along with crack dating, and hydrogen and deuterium analyses, indicated that the incipient cracks were the result of delayed-hydride assisted cracking at the EOL. Consequently, Ontario Hydro changed the design of the outlet shield plug to support all three rings of the fuel bundle, to minimize stress and prevent endplate cracking. Also, an ultrasonic endplate inspection tool (UT) was developed and located in the fuel bay. to inspect fuelbundle endplates for cracks. A second test was done involving a series of four bundle cascades in BAND Unit 4 channels that had new outlet shield plugs. The latch bundles were discharged after a hot shutdown. The cascaded Unit 2 and Unit 4 latch bundles were checked for cracks using the UT. The PIE found incipient cracks or less-than-ideal welds in the assembly welds of fuel elements from Unit 2 (latch-supported fuel bundles) that had been identified by the UT as having incipient cracks. No incipient cracks were found in the assembly welds of fuel elements from Unit 4 (new outlet shield

Influence of foreign matter on the flammability of Zircaloy

International Nuclear Information System (INIS)

Praetorius, R.; Muenzel, H.

1990-01-01

When cutting Zircaloy cladding in the head end of a reprocessing plant, fine particles with a high chemical reactivity are produced. Spontaneous ignition may cause fire or dust explosion. Therefore their ignition and fire behaviour was studied. As a result it can be stated that sugar or a concentrated sugar solution (syrup) poured over a Zircaloy fire is particularly suited as a fire-extinguishing agent. The developing caramel melt prevents air access and sparking. In addition, the sugar can be washed out easily before cementing, and so additional waste arising can be avoided. (DG) [de

Development of processes for zircaloy chips recycling by electric arc furnace remelting and powder metallurgy; Desenvolvimento de processos de reciclagem de cavacos de zircaloy via refusao em forno eletrico a arco e metalurgia do po

Energy Technology Data Exchange (ETDEWEB)

Pereira, Luiz Alberto Tavares

2014-09-01

PWR reactors employ, as nuclear fuel, UO{sub 2} pellets with Zircaloy clad. In the fabrication of fuel element parts, machining chips from the alloys are generated. As the Zircaloy chips cannot be discarded as ordinary metallic waste, the recycling of this material is important for the Brazilian Nuclear Policy, which targets the reprocess of Zircaloy residues for economic and environmental aspects. This work presents two methods developed in order to recycle Zircaloy chips. In one of the methods, Zircaloy machining chips were refused using an electric-arc furnace to obtain small laboratory ingots. The second one uses powder metallurgy techniques, where the chips were submitted to hydriding process and the resulting material was milled, isostatically pressed and vacuum sintered. The ingots were heat-treated by vacuum annealing. The microstructures resulting from both processing methods were characterized using optical and scanning electron microscopy. Chemical composition, crystal phases and hardness were also determined. The results showed that the composition of recycled Zircaloy comply with the chemical specifications and presented adequate microstructure for nuclear use. The good results of the powder metallurgy method suggest the possibility of producing small parts, like cladding end-caps, using near net shape sintering. (author)

Zircaloy-oxidation and hydrogen-generation rates in degraded-core accident situations

International Nuclear Information System (INIS)

Chung, H.M.; Thomas, G.R.

1983-02-01

Oxidation of Zircaloy cladding is the primary source of hydrogen generated during a degraded-core accident. In this paper, reported Zircaloy oxidation rates, either measured at 1500 to 1850 0 C or extrapolated from the low-temperature data obtained at 0 C, are critically reviewed with respect to their applicability to a degraded-core accident situation in which the high-temperature fuel cladding is likely to be exposed to and oxidized in mixtures of hydrogen and depleted steam, rather than in an unlimited flux of pure steam. New results of Zircaloy oxidation measurements in various mixtures of hydrogen and steam are reported for >1500 0 C. The results show significantly smaller oxidation and, hence, hydrogen-generation rates in the mixture, compared with those obtained in pure steam. It is also shown that a significant fraction of hydrogen, generated as a result of Zircaloy oxidation, is dissolved in the cladding material itself, which prevents that portion of hydrogen from reaching the containment building space. Implications of these findings are discussed in relation to a more realistic method of quantifying the hydrogen source term for a degraded-core accident analysis

Development of resistance welding process. 4. Preparation of pressuring enclosed creep test specimen of 7A material

International Nuclear Information System (INIS)

Endo, Hideo; Seki, Masayuki; Ishibashi, Fujio; Hirako, Kazuhito; Tsukada, Tatsuya

2001-02-01

Mechanical strength in the position welded by resistance welding system was examined in 1999. The test specimens were destroyed in the welding position in a shorter time than expected in the creep test. Therefore, test specimens were prepared to evaluate the cause of destruction. Inner-pressure enclosed creep test specimens were prepared by resistance welding method. Cladding material with low deviation of thickness and high re-crystallization rate was used. Heat treatment after resistance welding was performed to remove the influence of residual stress and the precipitation of carbides. (1) Before preparation of specimens, the welding condition was fixed. Three test specimens were prepared. Two specimens without heat treatment were transported to MMS in Oarai Engineering Center on Aug. 4, 2000. One specimen with heat treatment was transported to MMS after evaluating the residual stress to get optimum heat treatment condition. (2) Specimens were prepared with welding end plugs to both ends of ferritic ODS cladding. Enclosing sides were welded with highly strong Ferritic/Martensitic steel end plugs. The other sides were welded with ferritic ODS end plugs. (3) Some kinds of electrical wave data were obtained during performing welding. Welding position was evaluated with supersonic detector after performing welding. (4) Mechanical strength of welding position in high temperature 800degC was confirmed to be equal to or larger than that of cladding material. The highly qualified specimens in the present were successfully prepared. (author)

Morphologic and radiologic data of isolated vertebral end-plates under particular consideration of radiodensitometry

International Nuclear Information System (INIS)

Unflath, U.

1979-01-01

The intention was to find examination techniques for macroscopico-morphological diagnosis, for quantification of the concavity and methods for roentgenologico-densitometric analysis of vertebral end-plates. The following factors were used for macromorphologic diagnosis: osteophytes, ridge constitution, ridge transition, cribriform plate and concavity. The radiodensitometric analysis comprised investigation of bone density, anisotropy and ridge mineralisation. The specimen were radiographed on a mammograph together with calcium chloride solutions of increasing concentration. Statistically significant differences in the behaviour of basal and upper plates were not observed. With reference to the factors osteophytes and ridge constitution the basal plates seem to be morphologically more variable than the upper plates. It resulted that the features absolute concavity and concavity quotient are highly appropriate factors to document the constitution of the isolated vertebral end-plates. Also the features density, anisotropy and ridge mineralisation were found to be useful for documentation. The result of data correlation was that high concavity quotients are accompanied by macroscopically medium- to small-sized pores. Density and anisotropy of the overall profiles do not show correlation, but voluminous cribriform laminae tend to develop more inhomogenic structures. As to be expected, bone density increases with macroscopic fine porosity. The higher the concavity quotient is, the smaller are the values for density, ridge mineralisation and anisotropy. Thus a high degree of vertebral end-plate concavity indicates reduced bone density and lower mineral content. (orig./MG) [de

Thermal diffusion of hydrogen in zircaloy-2 containing hydrogen beyond terminal solid solubility

International Nuclear Information System (INIS)

Maki, Hideo; Sato, Masao.

1975-01-01

The thermal diffusion of hydrogen is one of causes of uneven hydride precipitation in zircaloy fuel cladding tubes that are used in water reactors. In the diffusion model of hydrogen in zircaloy, the effects of the hydride on the diffusibility of hydrogen has been regarded as negligibly small in comparison with that of hydrogen dissolved in the matrix. Contrary to the indications given by this model, phenomena are often encountered that cannot be explained unless hydride platelets have considerable ostensible diffusibility in zircaloy. In order to determine quantitatively the diffusion characteristics of hydrogen in zircaloy, a thermal diffusion experiment was performed with zircaloy-2 fuel cladding tubes containing hydrogen beyond the terminal solid solubility. In this experiment, a temperature difference of 20 0 --30 0 C was applied between the inside and outside surfaces of the specimen in a thermal simulator. To explain the experimental results, a modified diffusion model is presented, in which the effects of stress are introduced into Markowitz's model with the diffusion of hydrogen in the hydride taken into account. The diffusion equation derived from this model can be written in a form that ostensibly represents direct diffusion of hydride in zircaloy. The apparent diffusion characteristics of the hydride at around 300 0 C are Dsub(p)=2.3x10 5 exp(-32,000/RT), (where R:gas constant, T:temperature) and the apparent heat of transport Qsub(p) =-60,000 cal/mol. The modified diffusion model well explains the experimental results in such respects as reaches a steady state after several hours. (auth.)

The mechanisms of inhibition of frog endplate currents with homologous derivatives of the 1,1-dimethyl-3-oxybutyl phosphonic acid

Czech Academy of Sciences Publication Activity Database

Pryazhnikov, E.; Ostroumov, A.; Druginina, O.; Vyskočil, František; Skorinkin, A.

2012-01-01

Roč. 61, č. 4 (2012), s. 395-404 ISSN 0862-8408 R&D Projects: GA ČR(CZ) GA202/09/0806; GA AV ČR(CZ) IAA500110905 Institutional research plan: CEZ:AV0Z50110509 Keywords : endplate potential * nicotinic cholinoreceptor * allosteric modulation * open channel block * lipophilicity and cholinergic effect Subject RIV: ED - Physiology Impact factor: 1.531, year: 2012

Corrosion behaviour of zircaloy 4 fuel rod cladding in EDF power plants

Energy Technology Data Exchange (ETDEWEB)

Romary, H; Deydier, D [EDF, Direction de l` Equipment SEPTEN, Villeurbanne (France)

1997-02-01

Since the beginning of the French nuclear program, a surveillance of fuel has been carried out in order to evaluate the fuel behaviour under irradiation. Until now, nuclear fuels provided by suppliers have met EDF requirements concerning fuel behaviour and reliability. But, the need to minimize the costs and to increase the flexibility of the power plants led EDF to the definition of new targets: optimization of the core management and fuel cycle economy. The fuel behaviour experience shows that some of these new requirements cannot be fully fulfilled by the present standard fuel due to some technological limits. Particularly, burnup enhancement is limited by the oxidation and the hydriding of the Zircaloy 4 fuel rod cladding. Also, fuel suppliers and EDF need to have a better knowledge of the Zy-4 cladding behaviour in order to define the existing margins and the limiting factors. For this reason, in-reactor fuel characterization programs have been set up by fuel suppliers and EDF for a few years. This paper presents the main results and conclusions of EDF experience on Zy-4 in-reactor corrosion behaviour. Data obtained from oxide layer or zirconia thickness measurements show that corrosion performance of Zy-4 fuel rod cladding, as irradiated until now in EDF reactors, is satisfactory but not sufficient to meet the future needs. The fuel suppliers propose in order to improve the corrosion resistance of fuel rod cladding, low tin Zy-4 cladding and then optimized Zy-4 cladding. Irradiation of these claddings are ongoing. The available corrosion data show the better in-reactor corrosion resistance of optimized Zy-4 fuel rod cladding compared to the standard Zy-4 cladding. The scheduled fuel surveillance program will confirm if the optimized Zy-4 fuel rod cladding will meet the requirements for the future high burnup and high flexibility fuel. (author). 10 refs, 19 figs, 4 tabs.

Modification of hydrogen determinator for total hydrogen analysis in irradiated zircaloy cladding tube

International Nuclear Information System (INIS)

Park, Soon Dal; Choi, Kwnag Soon; Kim, Jong Goo; Joe, Kih Soo; Kim, Won Ho

1999-01-01

A hydrogen determinator was modified and installed in the glove box to analyse total hydrogen content in irradiated zircaloy tube. The analysis method of hydrogen is Inert Gas Fusion(IGF)-Thermal Conductivity Detection(TCD). The hydrogen recoveries of no tin method using Ti and Zr matrix standards, respectively, were available within 3 μg of hydrogen. Also the smaller size of sample showed the better hydrogen recovery. It was found that the hydrogen standard of Ti matrix is available to hydrogen analysis in zircaloy sample. The mean radioactivity of irradiated zircaloy sample was 10 mR/hr and hydrogen concentration was 130 ppm

Study of radiation effects on zircaloy 4 microstructure (Impact on susceptibility to fuel pellet-cladding interaction in PWR)

International Nuclear Information System (INIS)

Lefebvre, F.

1989-01-01

In PWR the fast neutron flux is an important parameter for fuel can aging by modification of zircaloy-4 microstructure: amorphisation and dissolution of intermetallic precipitates. These phenomena are both analysed and their influence on fuel-cladding interaction is discussed. Irradiations by 1 MeV electrons, Ar ions, Kr ions and fast neutrons are realized for comparison of damages with different defect creation kinetics. Amorphisation is explained as the crystal amorphous state transformation allowing precipitate dissolution by creation of a chemical potential gradient between matrix and amorphous phase. Progressive dissolution of precipitates produced by irradiation decrease the number of potential sites for stress corrosion cracking, improving rupture resistance of the alloy by fuel-cladding interaction [fr

Embrittlement of zircaloy cladding due to oxygen uptake (CBRTTL)

International Nuclear Information System (INIS)

Reymann, G.A.

1979-02-01

A model for embrittlement of zircaloy due to oxygen uptake at high temperatures is described. The model defines limits for oxygen content and temperature which, if exceeded, give rise to zircaloy cladding which is sufficiently embrittled to cause failure either on quenching or normal handling following a transient. A significant feature of this model is that the onset of embrittlement is dependent on the cooling rate. A distinction is made between slow and fast cooling, with the boundary at 100 K/s. The material property correlations and computer subcodes described in MATPRO are developed for use in Light Water Reactor (LWR) codes

Superficial characterization and zircaloy-2 electrochemistry with hydrothermal deposit of platinum; Caracterizacion superficial y electroquimica de zircaloy-2 con deposito hidrotermal de platino

Energy Technology Data Exchange (ETDEWEB)

Contreras R, A.; Arganis J, C. R.; Medina A, A. L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Gris C, M. M., E-mail: aida.contreras@inin.gob.mx [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

2011-11-15

The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy-2 tubes that contain in their interior UO{sub 2} pellets. With the objective of mitigating the speed of crack growth by IGSCC to a minimum negative impact on the BWR operation, General Electric developed the noble metals chemical addition (NMCA), in where noble metals particles as Pt, Pd, and Rh, are deposited on the surface of the metal to catalyze the recombination of H{sub 2} and O{sub 2}. Hydrogen is also injected to have it in excess and to favor this recombination (HWC) and zinc to reduce dose. In this work was oxidized zircaloy-2 low similar conditions to the HWC, platinum was deposited starting from a solution of Na{sub 2}Pt(OH){sub 6} with 30 ppm of Pt, in refined samples and without polishing, they were characterized by scanning electron microscopy, energy dispersed spectroscopy, XPS and electrochemistry, by means of Tafel curves and cyclical polarization. On the zircaloy surface was found a ZrO{sub 2} layer that remains under the different study conditions. Under HWC conditions is the oxides formation, possibly complex oxides of zirconium, iron and tin. After the platinum deposit these oxides decrease forming the sub-oxides: Zr{sub 2}O, Zr O, Zr{sub 2}O{sub 3}. The Tafel curves indicates the reduction of the oxygen of the sample with platinum and the cyclical polarization curves show that the reactions that happen on the zircaloy electrodes are not dur to located corrosion. (Author)

Conversion of zircaloy to a massive chemically inert form

International Nuclear Information System (INIS)

Atkinson, A.; Kearsey, H.A.; Knibbs, R.H.; Mercer, A.C.; Nickerson, A.K.; Pearson, D.; Sambell, R.A.J.; Taylor, R.I.

1985-01-01

The report covers work carried out in the period July 1980 - December 1982 on the development and assessment of an aqueous route for the conversion of Zircaloy fuel element cladding to a stable oxide form and on alternative methods for incorporating the oxide into monolithic waste forms suitable for long-term storage and disposal. The work included two aspects, preliminary process development studies aimed at demonstrating the key steps in the process, and studies on the alternative immobilization techniques and the properties of the resulting waste forms. Experimental studies have shown that the ''hydrous zirconium oxide'' (with a residual fluoride content), following calcination at about 500 0 C, can be hot-pressed at 800-1000 0 C and 22.5 MPa to a high density ceramic waste form with good capacity for the incorporation of active species, such as U 4+ and Sr 2+ , and high leach resistance. Parallel studies have been carried out on the incorporation of the washed ''hydrous zirconium oxide'' in a range of cement matrices. A preliminary chemical engineering assessment of the overall process has been made and flowsheets for a plant to convert 250 kg Zircaloy/day have been prepared

Prediction of water droplet evaporation on zircaloy surface

International Nuclear Information System (INIS)

Lee, Chi Young; In, Wang Kee

2014-01-01

In the present experimental study, the prediction of water droplet evaporation on a zircaloy surface was investigated using various initial droplet sizes. To the best of our knowledge, this may be the first valuable effort for understanding the details of water droplet evaporation on a zircaloy surface. The initial contact diameters of the water droplets tested ranged from 1.76 to 3.41 mm. The behavior (i.e., time-dependent droplet volume, contact angle, droplet height, and contact diameter) and mode-transition time of the water droplet evaporation were strongly influenced by the initial droplet size. Using the normalized contact angle (θ*) and contact diameter (d*), the transitions between evaporation modes were successfully expressed by a single curve, and their criteria were proposed. To predict the temporal droplet volume change and evaporation rate, the range of θ* > 0.25 and d* > 0.9, which mostly covered the whole evaporation period and the initial contact diameter remained almost constant during evaporation, was targeted. In this range, the previous contact angle functions for the evaporation model underpredicted the experimental data. A new contact angle function of a zircaloy surface was empirically proposed, which represented the present experimental data within a reasonable degree of accuracy. (author)

Evaluation of the DSC technique in the determination of the terminal temperature solubility of hydrogen in Zircaloy 4

International Nuclear Information System (INIS)

Vizcaino, Pablo; Banchik, Abraham D.

1999-01-01

The determination of the terminal temperature solubility of hydrogen in the zirconium alloys is a matter of our interest, since the hydride precipitation in these materials make them susceptible to the delay hydrogen cracking (DHC). Nowadays efforts are directed to the determination of this parameter on irradiated materials. In this frame we have begun a program to determine these curves on irradiated zirconium alloys like Zircaloys, Zr 2.5% Nb, Zr 1% Nb, zirconium sponge and high oxygen zirconium with different microstructures using the differential scanning calorimeter technique. The earlier task that we have imposed to us is the evaluation the reproducibility of this technique and the influence of the heating and cooling rates on the results. With these objectives we have initiated this work measuring the terminal temperature solubility (TTS) in unirradiated recrystallized and cold worked Zircaloy-4. We have found that the reproducibility of the technique is good; the average values (TSSD) and (TSSP) agree within the experimental error and the dispersions are low. Variance analysis indicates that the values of TSSD and TSSP do not depend on the heating and cooling rates. Within the rank analyzed the differences in the measurements are only possible to be attributed to statistical and/or experimental errors. The obtained curves of solubility show to a good agreement with the values obtained by others authors until 200 ppm (weight) of hydrogen. Similar comparisons between precipitation curves have not been made in regard to which these values depend on the maximum temperature reached in the experience. (author)

A new strain gage method for measuring the contractile strain ratio of Zircaloy tubing

International Nuclear Information System (INIS)

Hwang, S.K.; Sabol, G.P.

1988-01-01

An improved strain gage method for determining the contractile strain ratio (CSR) of Zircaloy tubing was developed. The new method consists of a number of load-unload cyclings at approximately 0.2% plastic strain interval. With this method the CSR of Zircaloy-4 tubing could be determined accurately because it was possible to separate the plastic strains from the elastic strain involvement. The CSR values determined by use of the new method were in good agreement with those calculated from conventional post-test manual measurements. The CSR of the tubing was found to decrease with the amount of deformation during testing because of uneven plastic flow in the gage section. A new technique of inscribing gage marks by use of a YAG laser is discussed. (orig.)

Internal-bore-welding of 2 1/4 Cr--1 Mo steel tube-to-tubesheet joints

International Nuclear Information System (INIS)

Moorhead, A.J.; Slaughter, G.M.

1976-01-01

In order to avoid the disadvantages of the conventional face-side tube-to-tubesheet weld, the steam generators for the Clinch River Breeder Reactor Plant (a power-producing demonstration LMFBR) will be built using a relatively new technique known as internal-bore-welding (IBW). In IBW the tube does not pass through the tubesheet but rather is welded to a short stub machined on the tube side of the tubesheet. This joint has the important advantages of being inspectable by radiography and eliminating the crevice; however, it is much more difficult to weld than is the face-side design. Because of the close proximity of the tubes, there is not room for an orbiting-arc welding head on the outside of the tube. Consequently, this weld must be made by welding from the inside- or bore-side of the tube. The results are presented of the initial phases of a program undertaken at ORNL to develop improved bore-side welding equipment, to gain further understanding of this technique, and to develop mechanical property data for autogeneous welds in 2 1/4 Cr-1 Mo steel tube and tubesheet materials

Influence de l'orientation des hydrures sur les modes de déformation, d'endommagement et de rupture du Zircaloy-4 hydruré.

OpenAIRE

Racine , Aude

2005-01-01

In pressurized water reactors of nuclear power plants, fuel pellets are contained in cladding tubes, made of Zirconium alloy, for instance Zircaloy-4. During their life in the primary water of the reactor (155 bars, 300°C), cladding tubes are oxidized and consequently hydrided. A part of the hydrogen given off precipitates as Zirconium hydrides in the bulk material and embrittles the material. This embitterment depends on many parameters, among which hydrogen content and orientation of hydrid...

Correlation of fatigue properties and microstructure in investment cast Ti-6Al-4V welds

International Nuclear Information System (INIS)

Oh, Jinkeun; Kim, Nack J.; Lee, Sunghak; Lee, Eui W.

2003-01-01

The effect of microstructural characteristics on high-cycle fatigue properties and fatigue crack propagation behavior of welded regions of an investment cast Ti-6Al-4V were investigated. High-cycle fatigue and fatigue crack propagation tests were conducted on the welded regions, which were processed by two different welding methods: tungsten inert gas (TIG) and electron beam (EB) welding. Test data were analyzed in relation to microstructure, tensile properties, and fatigue fracture mode. The base metal was composed of an alpha plate colony structure transformed to a basket-weave structure with thin α platelets after welding and annealing. High-cycle fatigue results indicated that fatigue strength of the EB weld was lower than that of the base metal or the TIG weld because of the existence of large micropores formed during welding, although it had the highest yield strength. In the case of the fatigue crack propagation, the EB weld composed of thinner α platelets had a faster crack propagation rate than the base metal or the TIG weld. The effective microstructural feature determining the fatigue crack propagation rate was found to be the width of α platelets because it was well matched with the reversed cyclic plastic zone size calculated in the threshold ΔK regime

Joint properties of dissimilar Al6061-T6 aluminum alloy/Ti–6%Al–4%V titanium alloy by gas tungsten arc welding assisted hybrid friction stir welding

International Nuclear Information System (INIS)

Bang, HanSur; Bang, HeeSeon; Song, HyunJong; Joo, SungMin

2013-01-01

Highlights: • Hybrid friction stir welding for Al alloy and Ti alloy joint has been carried out. • Mechanical strength of dissimilar joint by HFSW and FSW has been compared. • Microstructure of dissimilar joint by HFSW and FSW has been compared. - Abstract: Hybrid friction stir butt welding of Al6061-T6 aluminum alloy plate to Ti–6%Al–4%V titanium alloy plate with satisfactory acceptable joint strength was successfully achieved using preceding gas tungsten arc welding (GTAW) preheating heat source of the Ti alloy plate surface. Hybrid friction stir welding (HFSW) joints were welded completely without any unwelded zone resulting from smooth material flow by equally distributed temperature both in Al alloy side and Ti alloy side using GTAW assistance for preheating the Ti alloy plate unlike friction stir welding (FSW) joints. The ultimate tensile strength was approximately 91% in HFSW welds by that of the Al alloy base metal, which was 24% higher than that of FSW welds without GTAW under same welding condition. Notably, it was found that elongation in HFSW welds increased significantly compared with that of FSW welds, which resulted in improved joint strength. The ductile fracture was the main fracture mode in tensile test of HFSW welds

Fatigue properties of Zircaloy-2 in a PWR water environment

International Nuclear Information System (INIS)

Anon.

1984-01-01

The continuing trend of operation of light water reactors is towards power cycling as a means of operating the systems more efficiently. Depending upon the reactor design and mode of power cycling this could lead to significant fatigue usage in Zircaloy structural components. In order to design against the possibility of gross yielding or fast fracture of such components as a result of this it is obviously necessary to be able to predict conservatively the fatigue properties of Zircaloy under the reactor operating conditions

Ti–6Al–4V welded joints via electron beam welding: Microstructure, fatigue properties, and fracture behavior

Energy Technology Data Exchange (ETDEWEB)

Yang, Xiaoguang [School of Energy and Power Engineering, Beihang University, Beijing 100191 (China); Co-Innovation Center for Advanced Aero-Engine, Beijing 100191 (China); Li, Shaolin [School of Energy and Power Engineering, Beihang University, Beijing 100191 (China); Qi, Hongyu, E-mail: qhy@buaa.edu.cn [School of Energy and Power Engineering, Beihang University, Beijing 100191 (China); Co-Innovation Center for Advanced Aero-Engine, Beijing 100191 (China)

2014-03-01

The effect of microstructural characteristics on the fatigue properties of electron beam-welded joints of forged Ti–6Al–4V and its fracture behavior were investigated. Tensile tests and fatigue tests were conducted at room temperature in air atmosphere. The test data were analyzed in relation to microstructure, high-cycle fatigue properties, low-cycle fatigue properties, and fatigue crack propagation properties. The high-cycle fatigue test results indicated that the fatigue strength of the joint welded via electron beam welding was higher than that of the base metal because the former had a high yield strength and all high-cycle fatigue specimens were fractured in the base metal. Although the joint specimens had a lower low-cycle fatigue life than the base metal, they mainly ruptured at the fusion zone of the joint specimen and their crack initiation mechanism is load-dependent. The fatigue crack propagation test results show that the joint had a slower crack propagation rate than the base metal, which can be attributed to the larger grain in the fusion zone.

Cumulative fatigue and creep-fatigue damage at 3500C on recrystallized zircaloy 4

International Nuclear Information System (INIS)

Brun, G.; Pelchat, J.; Floze, J.C.; Galimberti, M.

1985-06-01

An experimental programme undertaken by C.E.A., E.D.F. and FRAGEMA with the aim of characterizing the fatigue and creep fatigue behaviour of zircaloy-4 following annealing treatments (recrystallized, stress-delived) is in progress. The results given below concern only recrystallized material. Cyclic properties, low-cycle fatigue curves and creep behaviour laws under stresses have been established. Sequential tests of pure fatigue and creep-fatigue were performed. The cumulative life fractions at fracture depend on the sequence of leading, stress history and number of cycles of prestressing. The MINER's rule appears to be conservative with regard to a low-high loading sequence whereas it is not for the reverse high-low loading sequences. Fatigue and creep damage are not interchangeable. Pre-creep improves the fatigue resistance. Pre-fatigue improves the creep strength as long as the beneficial effect of cyclic hardening overcomes the damaging effect of surface cracking. The introduction of a tension hold time into the fatigue cycle slightly increases cyclic hardening and reduces the number of cycles to failure. For hold times of less than one hour, the sum of fatigue and creep life fractions is closed to one

Electrochemical behavior of thin anodic oxide films on Zircaloy-4: Role of the mobile defects

International Nuclear Information System (INIS)

Salot, R.; Lefebvre-Joud, F.; Baroux, B.

1996-01-01

The first stages of the electrochemical oxidation of Zircaloy-4 are investigated using simple electrochemical tests and modeling the passive film modifications occurring as a result of contact with the electrolyte. Variations in electrode potential (open-circuit conditions) or current density (potentiodynamic scans) can be simply explained by a high field (F ∼ 10 6 V/cm) assisted passive film growth. Under open-circuit conditions, this field does not vary with exposure time (in the 2 h to 48 h range). The minimum electric field for the onset of high-field behavior is also evaluated and found smaller than the theoretical value which can be explained by a variation in the concentration of mobile defects throughout the film. Measurements of the electrode potential decay after a potentiodynamic scan confirm this model, allowing interpretation of the film modification as a combination of two separate phenomena: film growth under a high electric field and point defect annihilation

The Determination of Composite Elements in Zircaloy-2 by X-Ray Fluorescence and Emission Spectrometry Method

International Nuclear Information System (INIS)

Dian Anggraini; Rosika Kriswarini; Yusuf N

2007-01-01

Analysis of composing elements in zircaloy-2 has been done by Emission Spectrometry method and X-Ray Fluorescence (XRF). The aim of the analysis is to verify conformity between composing elements in zircaloy-2 and the material certificate. Spectrometry Emission method has higher sensitivity in element determination of a material than that of XRF method, so can be estimated that emission spectrometry method has higher accuracy than that of XRF method. The result of qualitative analysis by Emission Spectrometry indicate that the composing elements in zircaloy-2 were Sn, Cr and Ni. However, the qualitative analysis result by XRF method indicated that the composing elements in zircaloy 2 were Sn, Cr, Ni and Fe. Fe element can not be analysed by Emission Spectrometry method because Emission Spectrometer did not equipped with Fe detector. The quantitative analysis result of the composing elements in the material with both methods showed that Sn, Cr and Ni concentration of zircaloy 2 existed in concentration ranges of the material certificate. Result of statistical test (F and t-test) of analysis result of both methods can be used for analyzing composing elements in zircaloy 2. Emission Spectrometry method was more sensitive and accurate for determining Cr and Ni element in zircaloy 2 than that of emission Spectrometry method but both methods had same accuracy. The precision of measurement of Sn, Cr and Ni element using XRF method was better than that of Emission spectrometry method. (author)

Influence of laser treatment of beta phase on texture, mechanical properties and water corrosion of zircaloy 4

International Nuclear Information System (INIS)

Darchis, L.; Brun, G.; Baron, J.L.

1987-06-01

Two heat treatments by laser of zircaloy 4 cladding tubes are compared: one is superficial (1/1Oth of the thickness) and the other full thickness. In the full thickness treatment a global attenuation of preferential orientations is induced without marked gradient with the same texture found in a classical heat treatment. A peculiar texture is observed when beta transformation affects only 1/10. Mechanical properties measured by elongation and burst tests at 20 and 400 0 C are decreased by 40 to 60% for full thickness treatment and only 1 to 5% for superficial treatment. Water corrosion resistance at 360 0 C in conditions found for PWR is slightly increased by treatment on full thickness. No decrease of behavior is observed after 6 months on the martensitic structure obtained by superficial treatment [fr

Oxidation of Zircaloy Fuel Cladding in Water-Cooled Nuclear Reactors

Energy Technology Data Exchange (ETDEWEB)

Macdonald, Digby; Urquidi-Macdonald, Mirna; Chen, Yingzi; Ai, Jiahe; Park, Pilyeon; Kim, Han-Sang

2006-12-12

effects of second phase particles (SPPs) on the electrochemistry of passive zirconium in the hydrogenated, high temperature aqueous solutions were examined by using different heat-treated Zircaloy-4 samples; i.e., as-received, -quenched, and -annealed. The average size of the second phase particles in the Zircaloy-4 samples was in the sequence of -quenched < -annealed < as-received, with the reverse sequence being observed in the areal density. Electrochemical studies show that the size and density of the second phase particles are the determining factors of the electrochemical properties of the passive films. The second phase particles may cause short circuits in the electrical path across the passive film, which would explain the effect of the size and the density of the SPPs, and hence heat treatment, on the corrosion properties of passive Zircaloy-4.

Iodine stress corrosion cracking in Zircaloy

International Nuclear Information System (INIS)

Andrade, A.H.P. de; Pelloux, R.M.N.

1983-01-01

The subcritical growth of iodine-induced cracks in unirradiated Zircaloy plates is investigated as a function of the stress intensity factor K. The testing variables are: crystallographic texture (f-Number), microstructure (grain directionaly), heat treatment (stress relieved vs recrystallized plate), and temperature. The iodine partial pressure is 40Pa. (author) [pt

Mechanisms of the inhibition of endplate acetylcholine receptors by antiseptic chlorhexidine (experiments and models)

Czech Academy of Sciences Publication Activity Database

Shaihutdinova, A.R.; Nikolsky, E. E.; Vyskočil, František; Skorinkin, A.I.

2009-01-01

Roč. 380, č. 6 (2009), s. 551-560 ISSN 0028-1298 R&D Projects: GA AV ČR(CZ) IAA500110905 Institutional research plan: CEZ:AV0Z50110509 Keywords : acetylcholine * endplate currents Subject RIV: ED - Physiology Impact factor: 2.631, year: 2009

The effect of post-weld heat treatment on the microstructure and notched tensile fracture of Ti–15V–3Cr–3Al–3Sn to Ti–6Al–4V dissimilar laser welds

Energy Technology Data Exchange (ETDEWEB)

Hsieh, C.T.; Shiue, R.K. [Department of Materials Science and Engineering, National Taiwan University, Taipei 106, Taiwan, ROC (China); Huang, R.-T. [Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan, ROC (China); Tsay, L.W., E-mail: b0186@mail.ntou.edu.tw [Institute of Materials Engineering, National Taiwan Ocean University, Keelung 202, Taiwan, ROC (China)

2016-01-20

A CO{sub 2} laser was applied for dissimilar welding of Ti–15V–3Cr–3Al–3Sn (Ti–15–3) to Ti–6Al–4V (Ti–6–4) alloys. The microstructures and notched tensile strength (NTS) of the dissimilar welds were investigated in the as-welded and post-weld heat treatment (PWHT) conditions, and the results were compared with Ti–6–4 and Ti–15–3 homogeneous laser welds with the same PWHT. The results indicated that predominant α″ with a few α and β phases was formed in the as-welded fusion zone (FZ). Furthermore, the FZ hardness was susceptible to the PWHT and showed a plateau for the specimens aged in the temperature range from 426 to 482 °C/4 h. In comparison with the homogeneous Ti–15–3 weld under the same PWHT conditions, the dilution of Ti–15–3 with Ti–6–4 caused a slight increase in the Al equivalent (Al{sub EQ}) of the FZ, resulting in a further rise in FZ hardness. With the PWHT at/below 538 °C, the dissimilar welds were associated with low NTS or high notch brittleness.

Superficial characterization by XP S of silver nanoparticles and their hydrothermal deposit over zircaloy

International Nuclear Information System (INIS)

Contreras R, A.; Gutierrez W, C.; Martinez M, I.; Medina A, A. L.

2012-10-01

The analysis technique of X-ray photoelectron spectroscopy (XP S) is sensitive exclusively to the first layers of the solids surface, which allows obtaining information about the chemical, physical and electronic properties of them. The combustible elements of the boiling water nuclear reactors (BWR) are formed by zircaloy pipes that contain in their interior pellets or uranium dioxide. In this work is studied the zircaloy surface, oxidized zircaloy under similar conditions to those of a reactor BWR type and oxidized zircaloy with a hydrothermal deposit of silver nanoparticles and zinc. The silver deposit is a proposal of the Materials Technology Department of the Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico, which has the same objective that the noble metals deposit (Pt, Pd, and Rh) that is practiced in some of the reactors BWR, in order to mitigating the speed of crack growth for IGSCC in stainless steels 304 Ss. (Author)

Microstructure and orientation of hydrides in zircaloy-4 tubes for CAREM 25 fuel rods

International Nuclear Information System (INIS)

Juarez, G; Flores, A; Bianchi, D; Vizcaino, P

2012-01-01

The aim of the present work was giving support to the development of Zircaloy-4 fuel claddings for the CAREM 25 reactor through microstructural and mechanical properties studies along the manufacturing process. The manufacturing route was defined in 4 cold rolling stages and two thermal treatments, one at the middle and one after the last rolling stage. The first two rolling stages were performed in FAESA and the last two in PPFAECNEA using the rolling machine HPTR 8-15. The reference values for the evaluation were those indicated in the technical specification CAREM25 F ET-3-B0610. In this context, four tubes were received from FAESA. To these tubes microstructural and hydride orientation studies were performed to characterize the material in each step performed in PPFAE. The material received from FAESA has a grain size of 6 um, which fulfills the specification requirements for the final product (<11.2 um), so that, the final rolling stages (3rd and 4th) refine the grain until 3 um finished in the microstructure. Hydride morphology evolved from a random orientation in the recrystallized material or radial in the stress relieved material to the circumferential orientation requested in the specification. The final thermal treatment did not modify this morphology. The PD. (Marshall parameter) produced a value of 48, indicating that only 4% of the hydrides are oriented between 48 o and 90 o , that is, far from the circumferential direction (author)

Iodine stress-corrosion cracking in irradiated Zircaloy cladding

International Nuclear Information System (INIS)

Mattas, R.F.; Yaggee, F.L.; Neimark, L.A.

1979-01-01

Irradiated Zircaloy cladding specimens, which had experienced fluences from 0.1 to 6 x 10 21 n/cm 2 (E>0.1 MeV), were gas-pressure tested in an iodine environment to investigate their stress-corrosion cracking (SCC) susceptibility. The test temperatures and hoop stresses ranged from 320 to 360 0 C and 150 to 500 MPa, respectively. The results indicate that irradiation, in general, increases the susceptibility of Zircaloy to iodine SCC. For specimens that experienced fluences >2 x 10 21 n/cm 2 (E>0.1 MeV), the 24-h failure stress was 177+-18 MPa, regardless of the preirradiation metallurgical condition. An analytical model for iodine SCC has been developed which agrees reasonably well with the test results

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)

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

International Nuclear Information System (INIS)

Chapman, R.H.

1976-01-01

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

Testing of dissimilar metal welds according KTA 3201.4

International Nuclear Information System (INIS)

Giersbeck, Kai; Huenies, Gordon

2014-01-01

The amended German standard KTA 3201.4 from 2010 has triggered intensified requirements of the mechanized ultrasonic testing methodology. The report discusses the most important changes of KTA 3201.4 concerning the issues reference block, qualification of the testing methodology, testing of cladded surfaces, dissimilar metal joints, thermal conduits. The demonstration of dissimilar metal weld testing using intelligent NDT is demonstrated for the nozzles in German nuclear power plants.

Structure of Ti-6Al-4V nanostructured titanium alloy joint obtained by resistance spot welding

Energy Technology Data Exchange (ETDEWEB)

Klimenov, V. A., E-mail: klimenov@tpu.ru [Tomsk State University of Architecture and Building, 2 Solyanaya Sq, Tomsk, 634003 (Russian Federation); National Research Tomsk Polytechnic University, 30 Lenin Av., Tomsk, 634050 (Russian Federation); Kurgan, K. A., E-mail: kirill-k2.777@mail.ru [Tomsk State University of Architecture and Building, 2 Solyanaya Sq, Tomsk, 634003 (Russian Federation); Chumaevskii, A. V., E-mail: tch7av@gmail.com [Institute of Strength Physics and Materials Science, Siberian Branch of the Russian Academy of Sciences, 2/4 Akademicheskii pr., Tomsk, 634021 (Russian Federation); Klopotov, A. A., E-mail: klopotovaa@tsuab.ru [Tomsk State University of Architecture and Building, 2 Solyanaya Sq, Tomsk, 634003 (Russian Federation); National Research Tomsk State University, 36 Lenin Ave., Tomsk, 634050 (Russian Federation); Gnyusov, S. F., E-mail: gnusov@rambler.ru [National Research Tomsk Polytechnic University, 30 Lenin Av., Tomsk, 634050 (Russian Federation)

2016-01-15

The structure of weld joints of the titanium alloy Ti-6Al-4V in the initial ultrafine-grained state, obtained by resistance spot welding, is studied using the optical and scanning electron microscopy method and the X-ray structure analysis. The carried out studies show the relationship of the metal structure in the weld zone with main joint zones. The structure in the core zone and the heat affected zone is represented by finely dispersed grains of needle-shaped martensite, differently oriented in these zones. The change in the microhardness in the longitudinal section of the weld joint clearly correlates with structural changes during welding.

Effect of the aluminum flow pattern on the bonding of aluminum to oxidized Zircaloy-2

International Nuclear Information System (INIS)

Watson, R.D.; Lambert, J.P.

1965-04-01

The bonds produced when hot aluminum is allowed to flow smoothly from an extrusion die to the oxidized surface of a heated tube of Zircaloy-2 are consistently inferior to those produced with back-extruded flow. The difference is believed to be due to the reduction in, or elimination of, the oxide layer on the aluminum that comes in contact with the surface of the Zircaloy-2. This method of bonding aluminum to Zircaloy-2 is covered by Canadian patent 702,438 January 1965. (author)

Microstructure evolution of recrystallized Zircaloy-4 under charged particles irradiation

Science.gov (United States)

Gaumé, M.; Onimus, F.; Dupuy, L.; Tissot, O.; Bachelet, C.; Mompiou, F.

2017-11-01

Recrystallized zirconium alloys are used as nuclear fuel cladding tubes of Pressurized Water Reactors. During operation, these alloys are submitted to fast neutron irradiation which leads to their in-reactor deformation and to a change of their mechanical properties. These phenomena are directly related to the microstructure evolution under irradiation and especially to the formation of -type dislocation loops. In the present work, the radiation damage evolution in recrystallized Zircaloy-4 has been studied using charged particles irradiation. The loop nucleation and growth kinetics, and also the helical climb of linear dislocations, were observed in-situ using a High Voltage Electron Microscope (HVEM) under 1 MeV electron irradiation at 673 and 723 K. In addition, 600 keV Zr+ ion irradiations were conducted at the same temperature. Transmission Electron Microscopy (TEM) characterizations have been performed after both types of irradiations, and show dislocation loops with a Burgers vector belonging to planes close to { 10 1 bar 0 } first order prismatic planes. The nature of the loops has been characterized. Only interstitial dislocation loops have been observed after ion irradiation at 723 K. However, after electron irradiation conducted at 673 and 723 K, both interstitial and vacancy loops were observed, the proportion of interstitial loops increasing as the temperature is increased. The loop growth kinetics analysis shows that as the temperature increases, the loop number density decreases and the loop growth rate tends to increase. An increase of the flux leads to an increase of the loop number density and a decrease of the loop growth rate. The results are compared to previous works and discussed in the light of point defects diffusion.

Closing the weld gap with laser/mig hybrid welding process

DEFF Research Database (Denmark)

Bagger, Claus; Olsen, Flemming Ove; Wiwe, Bjarne David

2003-01-01

In this article, laboratory tests are demonstrated that systematically accesses the critical gap distance when welding CMn 2.13 mm steel with a 2.6 kW CO2 laser, combined with a MIG energy source. In the work, the welding speed is varied at gap distances from 0 to 0.8 mm such that the limits...... for obtaining sound welds are identified. The welds are quality assessed according to ISO 13.919-1 and EN25817, transversal hardness measurements are made and the heat input to the workpiece is calculated. The results show that the critical gap is 0.1 mm for a laser weld alone. With hybrid welding, this can...... be increased to 0.6 mm, even at a welding speed of 3.5 m/min. The maximum welding speed with the hybrid process is comparable to laser welding alone, 4.5 m/min. The measured hardness is comparable to MIG welding, and this corresponds to a 33 percent reduction compared to laser welding alone. The heat input...

Treatment of zircaloy cladding hulls by isostatic pressing

International Nuclear Information System (INIS)

Tegman, R.; Burstroem, M.

1984-12-01

A method for the treatment of Zircaloy fuel hulls is proposed. It involves hot isostatic pressing (HIP) for making large, completely densified metallic bodies of the waste. The hulls are packed into a bellows-shaped container of steel. On packing the fuel hulls give a filling factor of only 14%, which is too low for non-deformable compaction in a normal container, but by using a belloped container, a non-deformable compaction can be obtained without any pretreatment of the hulls. Fully dense and mechanically strong blocks of Zircaloy can be fabricated by holding them at temperatures of around 1000 degrees C for three hours. It is also feasible to incorporate the other metallic parts of the fuel bundle, such as top and bottom tie plates and spacers, in the pressing. The HIP-densified hulls provide an effective means of self-containment of radioactive waste due to the excellent corrosion resistance of Zircaloy. A waste loading factor of close to 100% can be realized. Futher, a volume reduction factor of 7 and a surface reduction factor of aout 250 for a 1-ton canister can be achieved. Equilibrium calculations have shown that tritium present in the hulls can quantitatively be contained in the HIPed block. A study has been made of a possible process for industrilscale use. (Author)

Effects of Friction Stir Welding on Corrosion Behaviors of AA2024-T4 Aluminum Alloy

Directory of Open Access Journals (Sweden)

Ales Steve Korakan

2017-01-01

Full Text Available In this work, the corrosion behavior of welded joints of AA2024-T4 Al alloy produced by friction stir welding process has been investigated. Tests were performed in an aerated 3.5% NaCl aqueous solution with pH = 7 at 20±2°C. Corrosion rate and corrosion morphology of weld regions were evaluated and compared to those of the parent metal. The microstructure of weld nugget, thermomechanical affected zone, heated affected zone, and parent metal were analyzed using scanning electron microscopy and energy dispersive spectroscopy. It was observed that corrosion initiated at FSW related spots and the sizes of local corrosion increased with time.

The Effect of Friction Stir Welding on Corrosion Behavior of Ti-6Al-4V

Science.gov (United States)

Nasresfahani, Ali Reza; Soltanipur, Abdol Reza; Farmanesh, Khosro; Ghasemi, Ali

2017-09-01

Fusion welding can deteriorate corrosion behavior of Ti-6Al-4V alloy. However, the use of friction stir welding leads to a more appropriate corrosion resistance. In this study, the corrosion resistance of welded zones of Ti-6Al-4V alloy using friction stir welding technique is evaluated. For these purposes, the study of structural characteristics using SEM and FESEM equipped with EDS micro-analyses was conducted. Micro-hardness test was also employed to estimate the hardness of welded zones. Corrosion behavior was investigated by a potentiostat instrument. SEM micrographs, EDS and XRD analyses confirmed non-uniformity of chemical composition within the welded zones. The results reveal that the stir zone contains typical alpha and prior beta phases. Nevertheless, thermomechanical zone included equiaxed and bimodal lamellae structure. Furthermore, the presence of different types of phases and microstructure in the thermomechanical zone led to reduced corrosion resistance. The corresponding values of corrosion current density in the stir zone, thermomechanical zone and base metal were 0.048, 0.55 and 0.032 µA, respectively. Corresponding corrosion potential for these zones was estimated as -207, -110 and -157 mV. Evidently, the results show that corrosion resistance of thermomechanical zone is less than that of the stir zone and both zones have lower value than the base metal.

TEM examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

2005-09-01

In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

Fracture toughness of austenitic stainless steel weld metal at 4 K

International Nuclear Information System (INIS)

Goodwin, G.M.

1984-08-01

Selection of the welding processess and weld filler metals for fabrication of a large toroidal superconducting magnet is described. Data available in the literature are collected and compared with data generated in this study for three welding processes, shielded metal arc (SMA), gas tungsten arc (GTA), and flux cored arc (FCA) welds had the highest fracture toughness as measured by K/sub Ic/ estimated from J/sub Ic/. The SMA and FCA welds had about the same toughness, below the GTA values but above the average from the literature. The fracture mode for all three processes was typified by ductile dimples. The fracture morphology of the FCA weld specimens was influenced by the solidification substructure, and small particles were found to be nucleation sites for void formation, especially for the GTA welds. All three welding processes were deemed adequate for the intended service and were used to fabricate the large magnet. A trunnion-type turning fixture eliminated the need for welding in the vertical and overhead positions. The GTA process was used for all root passes, and the horizontal welds were filled by the SMA process. Over 80% of the welds were done in the flat position with the FCA process, and its high deposition rate and ease of operation are credited with contributing greatly to the success of the effort

On use of weld zone temperatures for online monitoring of weld quality in friction stir welding of naturally aged aluminium alloys

International Nuclear Information System (INIS)

Imam, Murshid; Biswas, Kajal; Racherla, Vikranth

2013-01-01

Highlights: • FSWs for 6063-T4 AA are done at different process parameters and sheet thicknesses. • Weld nugget zone and heat affected zone temperatures are monitored for each case. • Microstructural and mechanical characterisation of welds is done in all cases. • Weld ductility is found to be particularly sensitive to weld zone temperatures. • Strong correlation is found between WNZ and HAZ temperatures and weld properties. - Abstract: 6063-T4 aluminium alloy sheets of 3 and 6 mm thicknesses were friction stir butt welded using a square tool pin at a wide range of tool rotational speeds. Properties of obtained welds were characterised using tensile tests, optical micrographs, X-ray diffraction, and transmission electron microscopy. Shape, size, and distribution of precipitates in weld zones, and strength and ductility of welds were seen to directly correlate with peak temperatures in weld nugget and heat affected zones, independent of sheet thickness. In addition, fluctuations in measured temperature profiles, for 3 mm sheets, were seen to correlate with an increase in scatter of weld nugget zone properties for 3 mm sheets. Optimal weld strength and ductility were obtained for peak weld nugget zone temperatures of around 450 °C and corresponding peak heat affected zone temperatures of around 360–380 °C. Results obtained suggest that, at least for naturally aged aluminium alloys, nature of temperature evolution and magnitudes of peak temperatures in weld nugget and heat affected zones provide information on uniformity of properties in weld zones, overaging of heat affected zones, and formation of tunnel defects from improper material mixing at low weld zone temperatures

Coating of Zircaloy sheaths with silica glass using the Sol-Gel technique for protection against oxidation

International Nuclear Information System (INIS)

De Sanctis, O.; Pellegri, N.; Gomez, L.

1990-01-01

With the aim of improving corrosion resistance of Zircaloy, a few Zircaloy sheaths were covered with vitreous silica. Deposition was made by dip coating in tetraetilortosilicate (TEOS) solutions and later densification treatment at 500 degrees C. Oxidation tests were performed and compared with sheaths not covered with silica. As a result, an effective increase in the resistance to dry oxidation was found in sheaths which had been protected. The coating-Zircaloy interface was studied using XPS (scanner). (Author). 6 refs., 3 figs

Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

Science.gov (United States)

Huang, F. H.

1993-12-01

Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

Stress-relaxation of zirconium and Zircaloy-4 near to 673 K

International Nuclear Information System (INIS)

Rubiolo, G.H.

1989-01-01

Stress-relaxation data in polycrystalline -zirconium and Zircaloy-4, between 645 K and 695 K, are reported. The study has been performed at different initial conditions of the material: recrystallized, cold-worked 64% by rolling and stress relieved at 813 K, for 1 h in high vacuum. The results are interpreted in terms of a constitutive equation of plastic deformation based on diffusion controlled motion of jogged screw dislocations and cell-formation. The internal stress field, in the recrystallized material, is assumed to be composed of two terms. A component is generated by the cell walls and is stable during a relaxation run. The other one is generated by the impurities segregated to the mobile dislocations and is strain rate dependent. From fitting of the experimental data to the model, it was possible to estimate: a) the activation energy for self-diffusion; b) the binding energy between the impurity and the dislocation; c) the activation energy for the diffusion of the impurity; d) the concentration of jogs and, e) the concentration of impurities in the crystal. The results obtained seem to indicate that oxygen is responsible for dynamic strain-ageing. It is concluded that, in the temperature region where strain-ageing is active, the mobile dislocations will form cell walls with jogs saturated with oxygen. This can inhibit climb and stop the recovery process in the walls. Furthermore, the strain-rate sensitivity parameter, derived from the proposed model, can explain the changes in curvature found on-the stress-relaxation curves between 298 and 723 K. (Author) [es

Ex-reactor determination of thermal gap and contact conductance between uranium dioxide: zircaloy-4 interfaces. Stage I: low gas pressure

International Nuclear Information System (INIS)

Garnier, J.E.; Begej, S.

1979-04-01

A study of thermal gap and contact conductance between depleted uranium dioxide (UO 2 ) and Zircaloy-4 (Zr4) has been made utilizing two measurement apparatuses developed as part of this program. The Modified Pulse Design (MPD) apparatus is a transient technique employing a heat pulse (laser) and a signal detector to monitor the thermal energy transmitted through a UO 2 /Zr4 sample pair which are either physically separated or in contact. The Modified Longitudinal Design (MLD) apparatus is a steady-state technique based on a modified cylindrical column design with a self-guarding sample geometry. Description of the MPD and MLD apparatus, data acquisition, reduction and error analysis is presented along with information on specimen preparation, thermal property and surface characterization. A technique using an optical height gauge to determine the average mean-plane of separation between the simple pairs is also presented

Air oxidation of Zircaloy-4, M5 (registered) and ZIRLOTM cladding alloys at high temperatures

International Nuclear Information System (INIS)

Steinbrueck, M.; Boettcher, M.

2011-01-01

The paper presents the results of isothermal and transient oxidation experiments of the advanced cladding alloys M5 (registered) and ZIRLO TM in comparison to Zircaloy-4 in air at temperatures from 973 to 1853 K. Generally, oxidation in air leads to a strong degradation of the cladding material. The main mechanism of this process is the formation of zirconium nitride and its re-oxidation. From the point of view of safety, the barrier effect of the fuel cladding is lost much earlier than during accident transients with a steam atmosphere only. Comparison of the three alloys investigated reveals a qualitatively similar, but quantitatively varying oxidation behavior in air. The mainly parabolic oxidation kinetics, where applicable, is comparable for the three alloys. Strong differences of up to 500% in oxidation rates were observed after transition to linear kinetics at temperatures below 1300 K. The paper presents kinetic rate constants as well as critical times and oxide scale thicknesses at the point of transition from parabolic to linear kinetics.

Temperature escalation in PWR fuel rod simulator bundles due to the Zircaloy/steam reaction: Test ESBU-2A

International Nuclear Information System (INIS)

Hagen, S.; Kapulla, H.; Malauschek, H.; Wallenfels, K.P.; Peck, S.O.

1984-07-01

This report describes the test conduct and results of the bundle test ESBU-2A, which was run to investigate the temperature escalation of zircaloy clad fuel rods. This investigation of temperature escalation is part of a series of out-of-pile experiments, performed within the framework of the PNS Severe Fuel Damage Program. The test bundle was of a 3 x 3 array of fuel rod simulators with a 0.4 m heated length. The fuel rod simulators were electrically heated and consisted of tungsten heaters, UO 2 annular pellets, and zircaloy cladding. A nominal steam flow of 0.7 g/s was inlet to the bundle. The bundle was surrounded by a zircaloy shroud which was insulated with ZrO 2 fiber ceramic wrap. The initial heatup rate of the bundle was 0.4 0 C/s. The temperature escalation began at the 255 mm elevation after 1200 0 C had been reached. At this elevation, the measured peak temperature was limited to 1500 0 C. It was concluded from different thermocouple results, that induced by this first escalation melt was formed in the lower part of the bundle. Consequently, the escalation in the lower part must be much higher, at least up to the melting temperature of zircaloy. Due to the failure in the steam production system, steam starvation in the upper region may explain the beginning of the escalation at the 255 mm elevation. The maximum temperature reached was 2175 0 C on the center rod at the end of the test. The unregularities in the steam supply may be the reason for less oxidation than expected. (orig./GL) [de

Characteristics and natural course of vertebral endplate signal (Modic) changes in the Danish general population

DEFF Research Database (Denmark)

Jensen, Tue S; Bendix, Tom; Sorensen, Joan S

2009-01-01

BACKGROUND: Vertebral endplate signal changes (VESC) are more common among patients with low back pain (LBP) and/or sciatica than in people who are not seeking care for back pain. The distribution and characteristics of VESC have been described in people from clinical and non-clinical populations...

Stacking faults in Zr(Fe, Cr)2 Laves structured secondary phase particle in Zircaloy-4 alloy.

Science.gov (United States)

Liu, Chengze; Li, Geping; Yuan, Fusen; Han, Fuzhou; Zhang, Yingdong; Gu, Hengfei

2018-02-01

Stacking faults (SFs) in secondary phase particles (SPPs), which generally crystallize in the Laves phase in Zircaloy-4 (Zr-4) alloy, have been frequently observed by researchers. However, few investigations on the nano-scale structure of SFs have been carried out. In the present study, an SF containing C14 structured SPP, which located at grain boundaries (GBs) in the α-Zr matrix, was chosen to be investigated, for its particular substructure as well as location, aiming to reveal the nature of the SFs in the SPPs in Zr-4 alloy. It was indicated that the SFs in the C14 structured SPP actually existed in the local C36 structured Laves phase, for their similarities in crystallography. The C14 → C36 phase transformation, which was driven by synchroshearing among the (0001) basal planes, was the formation mechanism of the SFs in the SPPs. By analyzing the strained regions near the SPP, a model for understanding the driving force of the synchroshear was proposed: the interaction between SPP and GB resulted in the Zener pinning effect, leading to the shearing parallel to the (0001) basal planes of the C14 structured SPP, and the synchroshear was therefore activated.

Prediction of residual stresses in electron beam welded Ti-6Al-4V plates

Energy Technology Data Exchange (ETDEWEB)

Xu, Lianyong; Ge, Keke; Jing, Hongyang; Zhao, Lei; Lv, Xiaoqing [Tianjin Univ. (China); Han, Yongdian [Tianjin Univ. (China). Key Lab. of Advanced Joining Technology

2017-05-01

A thermo-metallurgical procedure based on the SYSWELD code was developed to predict welding temperature field, microstructure and residual stress in butt-welded Ti-6Al-4V plate taking into account phase transformation. The formation of martensite was confirmed by the CCT diagram and microstructure in the weld joint, which significantly affects the magnitude of residual stress. The hole drilling procedure was utilized to measure the values of residual stress at the top surface of the specimen, which are in well agreement with the numerical results. Both simulated and test results show that the magnitude and distribution of residual stress on the surface of the plate present a large gradient feature from the weld joint to the base metal. Moreover, the distribution law of residual stresses in the plate thickness was further analyzed for better understanding of its generation and evolution.

The oxidation kinetics and the structure of the oxide film on Zircaloy before and after the kinetic transition

International Nuclear Information System (INIS)

Arima, T.; Masuzumi, T.; Furuya, H.; Idemitsu, K.; Inagaki, Y.

2001-01-01

Oxidation kinetics of Zircaloy-4 have been measured using a micro-balance technique in CO-CO 2 gas mixtures between 450 deg. C and 600 deg. C. Oxidation kinetics of Zircaloy-4 obeyed a cubic rate law with time at 450-600 deg. C up to 24 h. At 600 deg. C, the kinetic transition occurred after about 36 h. After the transition, oxidation kinetics obeyed a linear rate law. X-ray diffraction patterns for the samples oxidized at 600 deg. C showed that the volume fraction of tetragonal phase of zirconia decreased with time until the kinetic transition occurred and was almost constant after that. In addition, stresses in the oxide films were found to be larger for the pre-transition samples than for the post-transition ones. (authors)

Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

International Nuclear Information System (INIS)

Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E.

2000-01-01

We used 10-MeV electrons to rapidly produce radiation damage in zirconium alloys, investigated whether electrons produced the same microstructural changes as neutrons, then performed post-irradiation corrosion tests to determine whether electron-irradiated materials displayed similar corrosion behavior to neutron-irradiated materials. Two irradiations were completed using 10-MeV electrons with the beam normal to thin disks of material of 4 diameter slightly larger than the beam. The beam distribution. and disk cooling were designed to produce radial temperature and dose distributions having maxima at the disk center. A high-temperature irradiation was performed on annealed Zr-2.5Nb disks, achieving a central dose of 1.3 dpa and at a central temperature of ∼450 deg C. After irradiation, the samples contained needle-like β-Nb precipitates in the α-Zr matrix similar to those produced by neutrons. A low-temperature irradiation was performed on half-moon disks of Zr-2.5Nb and Zircaloy-2 pressure tube materials at 310 deg C central temperature and 1.3-dpa central dose. Dislocation loops were observed, again similar to those produced in neutron-irradiated materials. Some of the high-temperature electron-irradiated disks were exposed to 300 deg C moist air (saturated with D 2 O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D 2 0 (11.0 pD at room temperature) in an autoclave. Measurements of oxide thickness by Fourier Transform Infrared Reflectance (FTIR) spectroscopy showed that electron irradiation reduced the corrosion rate of Zr-2.5Nb compared with that of unirradiated material, as observed for neutron irradiation. For exposures to moist air and to D 2 O, the theoretical deuterium uptakes for the electron-irradiated materials were, respectively, about 4 times and 1.5 to 2 times those for the unirradiated materials. This is also in good agreement with results for neutron-irradiated pressure tube materials. Thus, 10-Me

A Comparative Study on the Laser Welding of Ti6Al4V Alloy Sheets in Flat and Horizontal Positions

Directory of Open Access Journals (Sweden)

Baohua Chang

2017-04-01

Full Text Available Laser welding has been increasingly utilized to manufacture a variety of components thanks to its high quality and speed. For components with complex shapes, the welding position needs be continuously adjusted during laser welding, which makes it necessary to know the effects of the welding position on the quality of the laser welds. In this paper, the weld quality under two (flat and horizontal welding positions were studied comparatively in the laser welding of Ti6Al4V titanium alloy, in terms of weld profiles, process porosity, and static tensile strengths. Results show that the flat welding position led to better weld profiles, less process porosity than that of the horizontal welding position, which resulted from the different actions of gravity on the molten weld metals and the different escape routes for pores under different welding positions. Although undercuts showed no association with the fracture positions and tensile strengths of the welds, too much porosity in horizontal laser welds led to significant decreases in the strengths and specific elongations of welds. Higher laser powers and travel speeds were recommended, for both flat and horizontal welding positions, to reduce weld porosity and improve mechanical properties.

Modeling of the cold work stress relieved Zircaloy-4 cladding tubes mechanical behavior under PWR operating conditions

International Nuclear Information System (INIS)

Richard, F.; Delobelle, P.; Leclercq, S.; Bouffioux, P.; Rousselier, G.

2003-01-01

This paper proposes a damaged viscoplastic model to simulate, for different isotherms (320, 350, 380, 400 and 420 degC), the out-of-flux anisotropic mechanical behavior of cold work stress relieved Zircaloy-4 cladding tubes over the fluence range 0-85.1024 nm -2 (E > 1 MeV). The model, identified from uni and biaxial tests conducted at 350 and 400 degC, is validated from tests performed at 320, 380 and 420 degC. This model is able to simulate strain hardening under internal pressure followed by a stress relaxation period (thermal creep), which is representative of a pellet cladding mechanical interaction occurring during a power transient (class 2 incidental condition). Both the integration of a scalar state variable, characterizing the damage caused by a bombardment with neutrons, and the modification of the static recovery law allowed us to simulate the fast neutron flux effect (irradiation creep). (author)

Oxidation-induced embrittlement and structural changes of Zircaloy-4 tubing in steam at 700-1000 deg. C

Energy Technology Data Exchange (ETDEWEB)

Ali, A E; Huessein, A G; El-Sayed, A A; El Banna, O A [Atomic Energy Authority, Cairo (Egypt); El Raghy, S M [Cairo Univ. (Egypt). Faculty of Engineering

1997-02-01

The oxidation-induced embrittlement and structural changes of Zircaloy-4 (KWU-Type) tubing was investigated under light water reactors (LWR) Loss-of-Coolant. Accident conditions (LOCA) in temperature range 700-1000 deg. C. The effect of hydrogen addition to steam was also investigated in the temperature range 800-1000 deg. C. The oxidation-induced embrittlement was found to be a function of both temperature and time. Fractography investigation of oxidized tubing showed a typical brittle fracture in the stabilized-alpha zone. The microhardness measurements revealed that the alpha-Zr is harder than that near the mid-wall position. The oxidation-induced embrittlement at 900 deg. C was found to be higher than at 1000 deg. C. The results also indicated that the addition of 5% by volume hydrogen to steam resulted in an increase in the degree of embrittlement. (author). 22 refs, 9 figs, 3 tabs.

Quantitative analysis of nickel in zirconium and zircaloy

International Nuclear Information System (INIS)

Rastoix, M.

1957-01-01

A rapid spectrophotometric has been developed for determination of 10 to 1000 ppm of Ni in zirconium and zircaloy using dimethylglyoxime. Iron, copper, tin and chromium, do not interfere at the concentration usually present in zirconium and its alloys. (author) [fr

Latest MIG, TIG arc-YAG laser hybrid welding systems for various welding products

Science.gov (United States)

Ishide, Takashi; Tsubota, Shuho; Watanabe, Masao

2003-03-01

Laser welding is capable of high-efficiency low-strain welding, and so its applications are started to various products. We have also put the high-power YAG laser of up to 10 kW to practical welding use for various products. On the other hand the weakest point of this laser welding is considered to be strict in the welding gap aiming allowance. In order to solve this problem, we have developed hybrid welding of TIG, MIG arc and YAG laser, taking the most advantages of both the laser and arc welding. Since the electrode is coaxial to the optical axis of the YAG laser in this process, it can be applied to welding of various objects. In the coaxial MIG, TIG-YAG welding, in order to make irradiation positions of the YAG laser beams having been guided in a wire or an electrode focused to the same position, the beam transmitted in fibers is separated to form a space between the separated beams, in which the laser is guided. With this method the beam-irradiating area can be brought near or to the arc-generating point. This enables welding of all directions even for the member of a three-dimensional shape. This time we carried out welding for various materials and have made their welding of up to 1 mm or more in welding groove gap possible. We have realized high-speed 1-pass butt welding of 4m/min in welding speed with the laser power of 3 kW for an aluminum alloy plate of approximately 4 mm thick. For a mild steel plate also we have realized butt welding of 1m/min with 5 kW for 6 mm thick. Further, in welding of stainless steel we have shown its welding possibility, by stabilizing the arc with the YAG laser in the welding atmosphere of pure argon, and shown that this welding is effective in high-efficiency welding of various materials. Here we will report the fundamental welding performances and applications to various objects for the coaxial MIG, TIG-YAG welding we have developed.

Research on the Effects of Technical Parameters on the Molding of the Weld by A-TIG Welding

OpenAIRE

Shi, Kai; Pan, Wu

2012-01-01

The effects of welding parameters on the molding of weld by A-TIG welding of a 4mm thickness mild steel plate is studied in the present paper. The results obtained show that: as welding current increases A-TIG welding penetration gets deeper than TIG welding; size and shape of HAZ has remarkable change; A-TIG welding has the narrower weld pool width than TIG welding.

Influence of the Overlapping Factor and Welding Speed on T-Joint Welding of Ti6Al4V and Inconel 600 Using Low-Power Fiber Laser

Directory of Open Access Journals (Sweden)

Shamini Janasekaran

2016-06-01

Full Text Available Double-sided laser beam welding of skin-stringer joints is an established method for many applications. However, in certain cases with limited accessibility, single-sided laser beam joining is considered. In the present study, single-sided welding of titanium alloy Ti6Al4V and nickel-based alloy Inconel 600 in a T-joint configuration was carried out using continuous-wave (CW, low-power Ytterbium (Yb-fiber laser. The influence of the overlapping factor and welding speed of the laser beam on weld morphology and properties was investigated using scanning electron microscopy (SEM and X-ray diffraction (XRD, respectively. XRD analysis revealed the presence of intermetallic layers containing NiTi and NiTi2 at the skin-stringer joint. The strength of the joints was evaluated using pull testing, while the hardness of the joints was analyzed using Vickers hardness measurement at the base metal (BM, fusion zone (FZ and heat-affected zone (HAZ. The results showed that the highest force needed to break the samples apart was approximately 150 N at a laser welding power of 250 W, welding speed of 40 mm/s and overlapping factor of 50%. During low-power single-sided laser welding, the properties of the T-joints were affected by the overlapping factor and laser welding speed.

Fundamental studies on electron-beam welding of heat-resistant superalloys for nuclear plants: Report 4. Mechanical properties of welded joints

International Nuclear Information System (INIS)

Susei, S.; Shimizu, S.; Aota, T.

1982-04-01

In this report, electron-beam (EB) welded joints and TIG welded joints of various superalloys to be used for nuclear plants, such as Hastelloy-type, Inconel-type and Incoloy-type, are systematically evaluated in terms of tensile properties, low-cycle fatigue properties at elevated temperatures, creep and creep-rupture properties. It was fully confirmed as conclusion that the EB welded joints are superior to the TIG welded ones in mechanical properties, especially at high temperature. In the evaluation of creep properties, ductility is one of the most important criteria to represent the resistance against fracture due to creep deformation, and this criterion is very useful in evaluating the properties of welded joints. Therefore, the more comparable to the base metal the electron beam welded joint becomes in terms of ductility, the more resistant is it against fracture. From this point of view, the electron beam welded joint is considerably superior to the TIG welded joint [fr

Neutron irradiation effects on intermetallic precipitates in Zircaloy as a function of fluence

International Nuclear Information System (INIS)

Etoh, Y.; Shimada, S.

1993-01-01

Intermetallic precipitates in Zircaloy-2 and -4, recrystallized at the α-phase temperature, have been examined using analytical electron microscopy. The specimens were irradiated in BWRs up to a fast neutron fluence of 1.4x10 26 n/m 2 (E>1 MeV). Neutron irradiation induces a crystalline-to-amorphous transition, depleting Fe in the amorphous phase of Zr(Fe, Cr) 2 precipitates in the alloys. Amorphization starts from the periphery of the precipitates and all of them are totally amorphized at higher fluences than 1.2x10 26 n/m 2 . The width of the Fe-depleted zone increases in proportion to the 0.45 power of fluence. This result indicates that diffusion of Fe is the rate-controlling process for Fe depletion in Zr(Fe, Cr) 2 precipitates. Dissolution of Zr 2 (Fe, Ni) precipitates in Zircaloy-2 occurs during neutron irradiation. At a high fluence, such as 1.2x10 26 n/m 2 , Zr 2 (Fe, Ni) precipitates are almost completely dissolved into the matrix and the dissolution rate of Fe is faster than that of Ni. (orig.)

A comprehensive model for in-plane and out-of-plane vibration of CANDU fuel endplate rings

Energy Technology Data Exchange (ETDEWEB)

Yu, S.D., E-mail: syu@ryerson.ca; Fadaee, M.

2016-08-01

Highlights: • Proposed an effective method for modelling bending and torsional vibration of CANDU fuel endplate rings. • Applied successfully the thick plate theory to curved structural members by accounting for the transverse shear effect. • The proposed method is computationally more efficient compared to the 3D finite element. - Abstract: In this paper, a comprehensive vibration model is developed for analysing in-plane and out-of-plane vibration of CANDU fuel endplate rings by taking into consideration the effects of in-plane extension in the circumferential and radial directions, shear, and rotatory inertia. The model is based on Reddy’s thick plate theory and the nine-node isoparametric Lagrangian plate finite elements. Natural frequencies of various modes of vibration of circular rings obtained using the proposed method are compared with 3D finite element results, experimental data and results available in the literature. Excellent agreement was achieved.

Operating envelope to minimize probability of fractures in Zircaloy-2 pressure tubes

International Nuclear Information System (INIS)

Azer, N.; Wong, H.

1994-01-01

The failure mode of primary concern with Candu pressure tubes is fast fracture of a through-wall axial crack, resulting from delayed hydride crack growth. The application of operating envelopes is demonstrated to minimize the probability of fracture in Zircaloy-2 pressure tubes based on Zr-2.5%Nb pressure tube experience. The technical basis for the development of the operating envelopes is also summarized. The operating envelope represents an area on the pressure versus temperature diagram within which the reactor may be operated without undue concern for pressure tube fracture. The envelopes presented address both normal operating conditions and the condition where a pressure tube leak has been detected. The examples in this paper are prepared to illustrate the methodology, and are not intended to be directly applicable to the operation of any specific reactor. The application of operating envelopes to minimized the probability of fracture in 80 mm diameter Zircaloy-2 pressure tubes has been discussed. Both normal operating and leaking pressure tube conditions have been considered. 3 refs., 4 figs

Hydrides formation In Zircaloy-4 irradiated with neutrons

International Nuclear Information System (INIS)

Vizcaino, P; Flores, A V; Vicente Alvarez, M A; Banchik, A.D; Tolley, A; Condo, A; Santisteban, J R

2012-01-01

Under reactor operating conditions zirconium components go through transformations which affect their original properties. Two phenomena of significant consequences for the integrity of the components are hydrogen uptake and radiation damage, since both contribute to the material fragilization. In the case of the Atucha I nuclear power reactor, the cooling channels, Zircaloy-4 tubular structural components about 6 meters long, were designed to withstand the entire lifetime of the reactor. Inside them, fuel elements 5.3 meters long are located. The fuel elements are cooled by a heavy water flow which circulates from the bottom (250 o ) to the top of the reactor (305 o C). The channels are affected by a fast neutron flux (En>1 Mev), increasing from a nominal value of 1.35 x 10 13 neutrons/cm 2 sec at the bottom to 1.69 x 10 13 neutrons/cm 2 sec at the top, reaching a maximum value of 3.76 x 10 13 neutrons/cm 2 sec at the center of the channels. However, due to the reactor operating conditions, they are replaced after about 10 effective full power years, time at which they reach 10 22 neutrons/cm 2 at the most neutronically active regions of the reactor. Studies on cooling channels are meaningful from many points of view. The channels are structural components which do not work under internal pressure or any other type of structural stress. The typical temperature of the cladding tubes in the reactor is about 350 o C, at which many types of irradiation defects are annealed [1]. The temperature range of the cooling channels lies between 200 o C-235 o C (outer foil of the channels) and 260 o C-300 o C (internal tube), a difference which makes the defect recovery kinetics slower. In the present context, following the program developed in the research contract 15810, we continue with the work started on the effects of the radiation on the hydride formation focusing on the dislocation loops in the zirconium matrix and its possible role as preferential sites for hydride

The effect of oxide microstructure on kinetic transition in out-of-pile steam corrosion test for Zircaloy-2 and Nb-added Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Nanikawa, Shuichi [Japan Nuclear Fuel Co. Ltd., Yokosuka, Kanagawa (Japan); Etoh, Yoshinori [Japan Nuclear Fuel Co. Ltd., Yokohama, Kanagawa (Japan)

2001-06-01

In order to study the mechanism of kinetic transition of corrosion rate for zirconium alloys, oxide films formed on Zircaloy-2 (Zry-2) and Nb-added Zircaloy-2 (0.5Nb/Zry-2) in steam at 673 K and 10.3 MPa were examined with TEM and SIMS. Kinetic transition occurred at almost the same oxide thicknesses for both Zry-2 and 0.5Nb/Zry-2, but the corrosion rate after the transitions were quite different for the two alloys. Zircaloy-2 showed cyclical oxidation, while the weight gain of 0.5Nb/Zry-2 increased linearly. The morphology and crystal structure were similar for the oxides of the two alloys and both the oxide films still mainly consisted of columnar grains even after the transition. Interface layers which mainly consisted of {alpha}-Zr crystallites were observed for both alloys and the oxygen content in the interface layers increased after the transition. The solute concentrations of Fe, Cr and Ni became higher, accompanying the increase of oxygen concentrations at columnar grain boundaries in the oxide films after the transition for 0.5Nb/Zry-2. It was thought that the properties of grain boundaries of the 0.5Nb/Zry-2 oxide films changed after the transition, and the increase in oxygen diffusivity at grain boundaries caused the linear increase in weight gain. (author)

Tensile properties in zircaloy-II after 590 MeV proton irradiation

Energy Technology Data Exchange (ETDEWEB)

Dai, Y. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Victoria, M. [Ecole Polytechnique Federale, Lausanne (Switzerland)

1997-09-01

In order to investigate radiation potential damage effects on the SINQ Zircaloy-rod target, four Zircaloy-II tensile specimens were irradiated at the PIREX facility in 1995 to a proton fluence about 3x10{sup 20} p/cm{sup 2}, which produced a radiation damage of about 1.35 displacements per atom (dpa). Tensile test results show that, although there is some reduction in tensile elongation, substantial ductility still exists after such irradiation dose which corresponds to the peak value obtained in the SINQ target for 23 days operation at 1 mA. (author) 1 fig., 2 refs.

Observations on deformation systems in zircaloy-2 deformed at room temperature

International Nuclear Information System (INIS)

Pettersson, K.; Bergqvist, H.

1975-08-01

Different polycrystalline samples of Zircaloy-2 with textures such that the c-axis of most of the grains are oriented near the sheet normal were subjected to loading conditions such that sheet thinning was accomplished. Metallography showed that no twinning was involved. Electron microscopy showed the presence of dislocations which were usually confined to deformation bands. With the help of stereo micrographs the most likely plane of slip was determined to be (1011). The possibility of slip as a means of breaking the oxide film in iodine induced stress corrosion cracking of Zircaloy-2 is briefly discussed. (author)

Effect of water-cooling treatment times on properties of friction stir welded joints of 7N01-T4 aluminum alloy

Science.gov (United States)

Zhang, T. H.; Wang, Y.; Fang, X. F.; Liang, P.; Zhao, Y.; Li, Y. H.; Liu, X. M.

2018-02-01

Due to the deformation caused by residual stress in the welding process, welded components need treatment to reduce welding distortion. In this paper, several different times of flame-heating and water-cooling treatment were subjected to the friction stir welding joints of 15mm thick 7N01P-T4 aluminum alloy sheets to study the microstructure variation of friction stir welding joints of 7N01P-T4 aluminum alloy, and to analyze the effect on micro-hardness, tensile and fracture mechanical properties. This investigation will be helpful to optimize treatment methods and provide instruction on industrial production.

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

Zircaloy PWR fuel cladding deformation tests under mainly convective cooling conditions

International Nuclear Information System (INIS)

Hindle, E.D.; Mann, C.A.

1980-01-01

In a loss-of-coolant accident the temperature of the cladding of the fuel rods may rise to levels (650-810 0 C) where the ductility of Zircaloy is high (approximately 80%). The net outward pressure which will obtain if the coolant pressure falls to a small fraction of its normal working value produces stresses in the cladding which can result in large strain through secondary creep. An earlier study of the deformation of specimens of PWR Zircaloy cladding tubing 450 mm long under internal pressure had shown that strains of over 50% could be produced over considerable lengths (greater than twenty tube diameters). Extended deformation of this sort might be unacceptable if it occurred in a fuel element. The previous tests had been carried out under conditions of uniform radiative heat loss, and the work reported here extends the study to conditions of mainly convective heat loss believed to be more representative of a fuel element following a loss of coolant. Zircaloy-4 cladding specimens 450 mm long were filled with alumina pellets and tested at temperatures between 630 and 845 0 C in flowing steam at atmospheric pressure. Internal test pressures were in the range 2.9-11.0 MPa (400-1600 1b/in 2 ). Maximum strains were observed of the same magnitude as those seen in the previous tests, but the shape of the deformation differed; in these tests the deformation progressively increased in the direction of the steam flow. These results are compared with those from multi-rod tests elsewhere, and it is suggested that heat transfer has a dominant effect in determining deformation. The implications for the behaviour of fuel elements in a loss-of-coolant accident are outlined. (author)

Lentiviral-mediated RNAi targeting caspase-3 inhibits apoptosis induced by serum deprivation in rat endplate chondrocytes in vitro

International Nuclear Information System (INIS)

Ding, L.; Wu, J.P.; Xu, G.; Zhu, B.; Zeng, Q.M.; Li, D.F.; Lu, W.

2014-01-01

Current studies find that degenerated cartilage endplates (CEP) of vertebrae, with fewer diffusion areas, decrease nutrient supply and accelerate intervertebral disc degeneration. Many more apoptotic cells have been identified in degenerated than in normal endplates, and may be responsible for the degenerated grade. Previous findings suggest that inhibition of apoptosis is one possible approach to improve disc regeneration. It is postulated that inhibition of CEP cell apoptosis may be responsible for the regeneration of endplates. Caspase-3, involved in the execution phase of apoptosis, is a candidate for regulating the apoptotic process. In the present study, CEP cells were incubated in 1% fetal bovine serum. Activated caspases were detected to identify the apoptotic pathway, and apoptosis was quantified by flow cytometry. Lentiviral caspase-3 short hairpin RNA (shRNA) was employed to study its protective effects against serum deprivation. Silencing of caspase-3 expression was quantified by reverse transcription-polymerase chain reaction and Western blots, and inhibition of apoptosis was quantified by flow cytometry. Serum deprivation increased apoptosis of rat CEP cells through activation of a caspase cascade. Lentiviral caspase-3 shRNA was successfully transduced into CEP cells, and specifically silenced endogenous caspase-3 expression. Surviving cells were protected by the downregulation of caspase-3 expression and activation. Thus, lentiviral caspase-3 shRNA-mediated RNAi successfully silenced endogenous caspase-3 expression, preventing inappropriate or premature apoptosis

Lentiviral-mediated RNAi targeting caspase-3 inhibits apoptosis induced by serum deprivation in rat endplate chondrocytes in vitro

Energy Technology Data Exchange (ETDEWEB)

Ding, L.; Wu, J.P. [Fudan University, Jinshan Hospital, Department of Orthopaedics, Shanghai, China, Department of Orthopaedics, Jinshan Hospital, Fudan University, Shanghai (China); Xu, G. [Fudan University, Jinshan Hospital, Center Laboratory, Shanghai, China, Center Laboratory, Jinshan Hospital, Fudan University, Shanghai (China); Zhu, B.; Zeng, Q.M.; Li, D.F.; Lu, W. [Fudan University, Jinshan Hospital, Department of Orthopaedics, Shanghai, China, Department of Orthopaedics, Jinshan Hospital, Fudan University, Shanghai (China)

2014-05-09

Current studies find that degenerated cartilage endplates (CEP) of vertebrae, with fewer diffusion areas, decrease nutrient supply and accelerate intervertebral disc degeneration. Many more apoptotic cells have been identified in degenerated than in normal endplates, and may be responsible for the degenerated grade. Previous findings suggest that inhibition of apoptosis is one possible approach to improve disc regeneration. It is postulated that inhibition of CEP cell apoptosis may be responsible for the regeneration of endplates. Caspase-3, involved in the execution phase of apoptosis, is a candidate for regulating the apoptotic process. In the present study, CEP cells were incubated in 1% fetal bovine serum. Activated caspases were detected to identify the apoptotic pathway, and apoptosis was quantified by flow cytometry. Lentiviral caspase-3 short hairpin RNA (shRNA) was employed to study its protective effects against serum deprivation. Silencing of caspase-3 expression was quantified by reverse transcription-polymerase chain reaction and Western blots, and inhibition of apoptosis was quantified by flow cytometry. Serum deprivation increased apoptosis of rat CEP cells through activation of a caspase cascade. Lentiviral caspase-3 shRNA was successfully transduced into CEP cells, and specifically silenced endogenous caspase-3 expression. Surviving cells were protected by the downregulation of caspase-3 expression and activation. Thus, lentiviral caspase-3 shRNA-mediated RNAi successfully silenced endogenous caspase-3 expression, preventing inappropriate or premature apoptosis.

Lentiviral-mediated RNAi targeting caspase-3 inhibits apoptosis induced by serum deprivation in rat endplate chondrocytes in vitro

Directory of Open Access Journals (Sweden)

L. Ding

2014-06-01

Full Text Available Current studies find that degenerated cartilage endplates (CEP of vertebrae, with fewer diffusion areas, decrease nutrient supply and accelerate intervertebral disc degeneration. Many more apoptotic cells have been identified in degenerated than in normal endplates, and may be responsible for the degenerated grade. Previous findings suggest that inhibition of apoptosis is one possible approach to improve disc regeneration. It is postulated that inhibition of CEP cell apoptosis may be responsible for the regeneration of endplates. Caspase-3, involved in the execution phase of apoptosis, is a candidate for regulating the apoptotic process. In the present study, CEP cells were incubated in 1% fetal bovine serum. Activated caspases were detected to identify the apoptotic pathway, and apoptosis was quantified by flow cytometry. Lentiviral caspase-3 short hairpin RNA (shRNA was employed to study its protective effects against serum deprivation. Silencing of caspase-3 expression was quantified by reverse transcription-polymerase chain reaction and Western blots, and inhibition of apoptosis was quantified by flow cytometry. Serum deprivation increased apoptosis of rat CEP cells through activation of a caspase cascade. Lentiviral caspase-3 shRNA was successfully transduced into CEP cells, and specifically silenced endogenous caspase-3 expression. Surviving cells were protected by the downregulation of caspase-3 expression and activation. Thus, lentiviral caspase-3 shRNA-mediated RNAi successfully silenced endogenous caspase-3 expression, preventing inappropriate or premature apoptosis.

Effects of B4C Addition on the Laser Beam Welding Characteristics of Al/SiC MMCs Produced By P/M

Directory of Open Access Journals (Sweden)

Serdar KARAOĞLU

2011-01-01

Full Text Available Fusion weldability characteristics of metal matrix composites (MMC produced by powder metallurgy (P/M are usually insufficient due to unwanted micro-structural changes that occur during welding. This study aims to investigate the effects of B4C addition as reinforcement on the weld quality of Al/SiC MMCs. After the production of Al/SiC MMCs by P/M with or without the addition of B4C, laser beam welding (LBW characteristics of the materials were investigated by focusing on the integrity of the welds. Optical microscopy (OM, scanning electron microscopy (SEM, and energy dispersive X-ray analysis (EDX were utilized for the characterization of the welds. Results show that Al/SiC MMCs produced by P/M can not be easily welded by LBW, but weldability characteristics of the material can be improved by the addition of B4C.