Feasibility demonstration of using wire electrical-discharge machining, abrasive flow honing, and laser spot welding to manufacture high-precision triangular-pitch Zircaloy-4 fuel-rod-support grids

International Nuclear Information System (INIS)

Horwood, W.A.

1982-05-01

Results are reported supporting the feasibility of manufacturing high precision machined triangular pitch Zircaloy-4 fuel rod support grids for application in water cooled nuclear power reactors. The manufacturing processes investigated included wire electrical discharge machining of the fuel rod and guide tube cells in Zircaloy plate stock to provide the grid body, multistep pickling of the machined grid to provide smooth and corrosion resistant surfaces, and laser welding of thin Zircaloy cover plates to both sides of the grid body to capture separate AM-350 stainless steel insert springs in the grid body. Results indicated that dimensional accuracy better than +- 0.001 and +- 0.002 inch could be obtained on cell shape and position respectively after wire EDM and surface pickling. Results on strength, corrosion resistance, and internal quality of laser spot welds are provided

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

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

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

Process development for fabrication of zircaloy- 4 of dissolver assembly for spent nuclear fuel reprocessing

International Nuclear Information System (INIS)

Tonpe, Sunil; Saibaba, N.; Jairaj, R.N.; Ravi Shankar, A.; Kamachi Mudali, U.; Raj, Baldev

2010-01-01

Spent fuel reprocessing for fast breeder reactor (FBR) requires a dissolver made of a material which has resistance to corrosion as the process involves Nitric Acid as the process medium. Various materials to achieve minimum corrosion rates have been tried for this operation. Particularly the focus was on the use of advanced materials with high performance (corrosion rate and product life) for high concentrations greater than 8 N and temperatures (boiling and vapour) of Nitric Acid employed in the dissolver unit. The different commercially available materials like SS316L , Pure Titanium, Ti - 5% Ta and Ti - 5% Ta - 1.8% Nb were tried and the corrosion behavior of these materials was studied in detail. As this is continuous process of evolution of new materials, it was decided to try out zircaloy - 4 as the material of construction for construction due to its excellent corrosion resistance properties in Nitric Acid environment. The specifications were stringent and the geometrical configurations of the assembly were very intricate in shape. On accepting the challenge of fabrication of dissolver, NFC has made different fixtures for Electron Beam Welding and TIG Welding. Various trials were carried out for optimization of various operating parameter like beam current, Acceleration voltage, welding speed to get adequate weld penetration. Both EB welding and TIG welding process were standardized and qualified by carrying out a number of trials and testing these welds by various weld qualification procedures like radiography, Liquid dye penetrant testing etc. for different intricate weld geometries. All the welds were simulated with samples to optimize the weld parameters. Tests such as include metallographic (for microstructure and HAZ), mechanical (for weld strength) and chemical (material analysis for gases) were conducted and all the weld samples met the acceptable criteria. Finally the dissolver was made meeting stringent specifications. All the welds were checked

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.)

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

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)

Identification of the zirconium hydrides metallography in zircaloy-2; Contribucion al estudio por metalografia de los hidruros de circonio en Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Garcia Gonzalez, F

1968-07-01

Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs.

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

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

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.

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)

Technical Support of Performance Improvement for Resistance Welding Using Zr-4 Endcap and Endplate

Energy Technology Data Exchange (ETDEWEB)

Kim, Soo Sung

2008-10-15

The proper welding process for Zircaloy-4 endplate of PHWR and DUPIC fuel bundle assembly is considered important in respect to the soundness of weldment and the improvement of the performance of nuclear fuel bundle during the operation in reactor. The Zircaloy-4 endplate of PHWR and DUPIC fuel bundles are welded by the projection joint type, connecting the endcaps of fuel elements. Therefore, the purpose of this projection joint is to improve the welding quality of torque strength and welding deformation and to apply the commercial productions for the endplate welding of PHWR and DUPIC nuclear fuel bundle assembly.

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

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)

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

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)

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

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)

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.)

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

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)

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.

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

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)

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.

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

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)

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)

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

Tensile creep of beta phase zircaloy-2

International Nuclear Information System (INIS)

Burton, B.; Reynolds, G.L.; Barnes, J.P.

1977-08-01

The tensile creep and creep rupture properties of beta-phase zircaloy-2 are studied under vacuum in the temperature and stress range 1300-1550 K and 0.5-2 MN/m 2 . The new results are compared with previously reported uniaxial and biaxial data. A small but systematic difference is noted between the uniaxial and biaxial creep data and reasons for this discrepancy are discussed. (author)

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)

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.

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)

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)

Dislocation Arrangements in Deformed and Neutron Irradiated Zirconium and Zircaloy-2

International Nuclear Information System (INIS)

Roy, R.B.

1963-12-01

Dislocation arrangements in deformed and neutron irradiated Zr and Zircaloy-2 have been studied by thin film transmission electron microscopy. Results indicate that the prominent slip system, in both Zr and Zircaloy-2, is the {1010} 1/3 type; no evidence for basal slip was observed. Attractive and repulsive dislocation interactions seem to be more important than the intersection jog reactions. Elongated loops and dipoles were seen at higher deformations and it is suspected that such loops or dipoles are formed due to interactions between dislocations lying in parallel planes. Stacking fault ribbons lying in {1010} plane have been found in 15% cold rolled Zircaloy-2: a rough estimate of stacking fault energy indicates that it is ∼ 65 ergs/cm 2 . Calculations show that the equilibrium separation of partials is ∼ 60 A and a stress as high as 19x10 -3 μ acting along {0010} direction is needed to separate them. It has been suggested that O 2 and N 2 in addition to their solid solution hardening effect may also cause a lowering of the stacking fault energy and Suzuki hardening

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.)

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)

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)

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

Dislocation Arrangements in Deformed and Neutron Irradiated Zirconium and Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Roy, R B

1963-12-15

Dislocation arrangements in deformed and neutron irradiated Zr and Zircaloy-2 have been studied by thin film transmission electron microscopy. Results indicate that the prominent slip system, in both Zr and Zircaloy-2, is the {l_brace}1010{r_brace} 1/3 <1210> type; no evidence for basal slip was observed. Attractive and repulsive dislocation interactions seem to be more important than the intersection jog reactions. Elongated loops and dipoles were seen at higher deformations and it is suspected that such loops or dipoles are formed due to interactions between dislocations lying in parallel planes. Stacking fault ribbons lying in {l_brace}1010{r_brace} plane have been found in 15% cold rolled Zircaloy-2: a rough estimate of stacking fault energy indicates that it is {approx} 65 ergs/cm{sup 2}. Calculations show that the equilibrium separation of partials is {approx} 60 A and a stress as high as 19x10{sup -3} {mu} acting along {l_brace}0010{r_brace} direction is needed to separate them. It has been suggested that O{sub 2} and N{sub 2} in addition to their solid solution hardening effect may also cause a lowering of the stacking fault energy and Suzuki hardening.

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)

A regression approach for Zircaloy-2 in-reactor creep constitutive equations

International Nuclear Information System (INIS)

Yung Liu, Y.; Bement, A.L.

1977-01-01

In this paper the methodology of multiple regressions as applied to Zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor Zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) When there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets. (2) Regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections. Multiple regression analysis performed on a set of carefully selected Zircaloy-2 in-reactor creep data leads to a model which provides excellent correlations for the data. (Auth.)

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

Thermal expansion studies on zircaloy-2

International Nuclear Information System (INIS)

Sivabharathy, M.; Senthilkumar, A.; Palanichamy, P.; Ramachandran, K.

2016-01-01

Zircaloy-2 and Zr-2.5% Nb alloys are widely used in the pressurized heavy water reactors (PHWR) as the material for the pressure tubes. The pressure tube operates at 573 K, 11 MPa internal pressures and is subjected to neutron flux of the order of 1013 n/cm 2 /s. These conditions lead to degradations in the pressure tube with respect to dimensional changes, deterioration in mechanical properties due to irradiation embrittlement, thereby reducing its flaw tolerance, the growth of existing flaws, which were too small or 'insignificant' at the time of installation. Physical and chemical properties of materials are also very essential in nuclear industry and the relations among them is of interest in the selection of materials when they are used in the design and manufacturing of devices particularly for atomic reactors.Studies on the relations between mechanical and thermal properties are of interest to the steel and metal industries as these would give useful information on the relation between hardness and thermal diffusivity (α) of steel. Jayakumar et al have already carried out the ultrasonic and metallographic investigations to see that all the heat-treated specimens retained essentially the martensite structure. In this present work, thermal expansion measurements on useful reactor material, Zircaloy-2 with different sample. Given a β-quenching treatment by heating to 1223 K and holding for 2 h, followed by water quenching. These specimens were then thermally aged for 1 h in the temperature range 473 to 973 K and air-cooled. For all samples, the thermal expansion was carried out and the results are correlated with ultrasonic measurements, metallographic and photoacoustic studies. (author)

Effect of impurity elements Al, Mn, and N2 on the corrosion resistance of zircaloy-2 in high temperature water and steam

International Nuclear Information System (INIS)

Gadiyar, H.S.

1978-01-01

Although the impurity limits are specified in standard zircaloy-2, it is possible that during its manufacture some of the impurities may exceed by a few ppm than the normally set values. It is necessary to understand the corrosion behaviour of such zircaloy-2 which contain a small amount of excessive impurities. This report summarizes some such data of the impurities aluminium, manganese and nitrogen. It is seen that the common impurities which can affect the corrosion of zircaloy-2 significantly are Al and N 2 and to a lesser extent Mn. (author)

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.

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.

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

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

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 anisotropic mechanical behaviour of zircaloy-2

International Nuclear Information System (INIS)

Ballinger, R.; Pelloux, R.M.

1980-01-01

Zirconium alloys used in the LWR industry crystallize in the hexagonal crystal structure below approximately 1136 K and many of the fabrication steps are performed below this temperature. The hexagonal structure possesses a limited number of slip systems and normal deformation processes result in extensive twinning. The twinning process results in the development of a fabrication texture, the type and extent of which is a function of the strain path used in the fabrication process. The texture which develops is important for two reasons. First, the texture at a given point in the fabrication process will determine the ease with which the next strain increment may be taken. Second, the texture of the completed part will have a significant effect on its in service performance because properties such as yield strength, creep strength, and fatigue and stress corrosion cracking resistance are a strong function of texture. Currently there is little data available concerning the evolution of textures as a function of strain path during the fabrication process of Zircaloy. Consequently this experimental investigation was conducted to determine the effect of textures on the mechanical behaviour of Zircaloy-2 with a primary emphasis on the evolution of texture during plastic deformation. (author)

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

Electromigration of hydrogen in zircaloy-2

International Nuclear Information System (INIS)

Parmeswaran, P.; Kamachi Mudali, U.; Raghunathan, V.S.; Govinda Rajan, K.

1989-01-01

Electromigration is a purification technique for removing interstitial impurities from metals like Zr, Ti and Nb. It uses an electric field to induce migration of atoms from one end to other. This paper describes an attempt to purify zircaloy-2 of its hydrogen content by this technique. Resistivity measurement has been used to evaluate the change in impurity concentration that occurs during the process. Results indicate the movement of hydrogen atoms towards the cathode end. The value of the effective charge number, Z * , calculated from the results confirms hydrogen migration to the cathode aided by a positive wind force. (author). 6 refs., 5 figs

Evolution of deformation velocity in narrowing for Zircaloy 2

Energy Technology Data Exchange (ETDEWEB)

Cetlin, P R [Minas Gerais Univ., Belo Horizonte (Brazil). Dept. de Engenharia Metalurgica; Okuda, M Y [Goias Univ., Goiania (Brazil). Inst. de Matematica e Fisica

1980-09-01

Some studies on the deformation instability in strain shows that the differences in this instability may lead to localized narrowing or elongated narrowing, for Zircaloy-2. The variation of velocity deformation with the narrowing evolution is expected to be different for these two cases. The mentioned variation is discussed, a great difference in behavior having been observed for the case of localized narrowing.

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.)

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.

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)

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

Corrosion of electron-irradiated Zr-2.5Nb and Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Woo, O.-T.; McDougall, G.M.; Hutcheon, R.M.; Urbanic, V.F.; Griffiths, M.; Coleman, C.E

2000-07-01

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 {approx}450 deg C. After irradiation, the samples contained needle-like {beta}-Nb precipitates in the {alpha}-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{sub 2}O), and in separate tests, high- and low-temperature irradiated disks were corroded in 300 deg C D{sub 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{sub 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

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

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

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

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)

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

Influence of manufacturing process on the in-reactor creep anisotropy of stress-relieved Zircaloy-2 cladding

International Nuclear Information System (INIS)

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

1995-01-01

A procedure to determine the axial/radial and circumferential/radial contractile strain ratios (the R and P factors respectively in the Backofen-modified von Mises-Hill yield criterion) from post-irradiation dimensional measurements of Zircaloy-2 cladding of BWR fuel rods, tie rods and water rods was developed and has been described previously (S.H. Shann and L.F. van Swam, Creep anisotropy of Zircaloy-2 cladding during irradiation, Trans. SMiRT-11, Vol. C, 1991). The present study employs the procedure to determine the anisotropy factors R and P for textured cold-worked stress-relieved (CWSR) Zircaloy-2 cladding fabricated by various manufacturing processes. The analysis indicates that the cladding manufacturing process can have a pronounced effect on the anisotropy of irradiation-induced creep. Cladding types with identical yield and ultimate tensile strengths but fabricated by different manufacturing processes have different values of R and P during in-reactor creep. ((orig.))

The corrosion of zircaloy 2 in anaerobic synthetic cement pore solution

International Nuclear Information System (INIS)

Hansson, C.M.

1984-12-01

Measurements have been made of the corrosion rates of Zircaloy 2 tubes in anaerobic synthetic cement pore solution of pH 12.0-13.8. The samples were tested in the as-received condition by the polarization resistance technique using a Tafal constant of 52 mV/decade and, for all pH values, corrosion rates of 3.10 -5 A/m 2 (0.03 μm/yr) were determined. These corrosion currents are at the lower limit of the experimental detection range of the technique used. Some samples were then held at a low electrochemical potential, namely -1850 mV SCE, for several days but this treatment had only a minor effect on the behaviour of the Zircaloy: the value of corrosion rate was increased by a factor of 3 and the free potential was temporarily lowered but drifted towards more positive values after the applied potential was removed. Attempts were made to remove the passive film from the surface of the samples by electrochemical reduction. For practical, experimental reasons, this was not successful and, instead, the effect of removing the film by scratching the surface was investigated. At both the free potential and at applied cathodic potentials, an anodic current was detected immediately and the surface was scratched but, in all cases, the scratched area repassivated within a few seconds and the anodic corrosion current fell accordingly. Thus, it may be concluded that active corrosion of Zircaloy 2 in anaerobic concrete will not occur and, by comparison with measurements on steel, it is likely that the passive corrosion rates will be even lower in concrete than those measured in the synthetic pore solution. (Author)

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.)

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.

Investigation of Zircaloy-2 oxidation model for SFP accident analysis

Energy Technology Data Exchange (ETDEWEB)

Nemoto, Yoshiyuki, E-mail: nemoto.yoshiyuki@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki [Global Nuclear Fuel – Japan Co., Ltd., 2-3-1, Uchikawa, Yokosuka-shi, Kanagawa, 239-0836 (Japan)

2017-05-15

The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. - Highlights: •An oxidation model of Zircaloy-2 in air environment was developed. •The oxidation model was validated by the comparison with oxidation tests using long cladding tubes in hypothetical spent fuel pool accident condition. •The oxidation model successfully reproduced the typical oxidation behavior in air.

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

Cumulative damage fatigue tests on nuclear reactor Zircaloy-2 fuel tubes at room temperature and 3000C

International Nuclear Information System (INIS)

Pandarinathan, P.R.; Vasudevan, P.

1980-01-01

Cumulative damage fatigue tests were conducted on the Zircaloy-2 fuel tubes at room temperature and 300 0 C on the modified Moore type, four-point-loaded, deflection-controlled, rotating bending fatigue testing machine. The cumulative cycle ratio at fracture for the Zircaloy-2 fuel tubes was found to depend on the sequence of loading, stress history, number of cycles of application of the pre-stress and the test temperature. A Hi-Lo type fatigue loading was found to be very much damaging at room temperature and this feature was not observed in the tests at 300 0 C. Results indicate significant differences in damage interaction and damage propagation under cumulative damage tests at room temperature and at 300 0 C. Block-loading fatigue tests are suggested as the best method to determine the life-time of Zircaloy-2 fuel tubes under random fatigue loading during their service in the reactor. (orig.)

Plastic strain accumulation during asymmetric cyclic loading of Zircaloy-2 at room temperature

International Nuclear Information System (INIS)

Rajpurohit, R.S.; Santhi Srinivas, N.C.; Singh, Vakil

2016-01-01

Asymmetric cyclic loading leads to accumulation of cyclic plastic strain and reduces the fatigue life of components. This phenomenon is known as ratcheting fatigue. Zircaloy-2 is a important structural material in nuclear reactors and used as pressure tubes and fuel cladding in pressurized light and heavy water nuclear reactors. Due to power fluctuations, these components experience plastic strain cycles in the reactor and their life is reduced due to strain cycles. Power fluctuations also cause asymmetric straining of the material and leads to accumulation of plastic strain. The present investigation deals with the effect of the magnitude of mean stress, stress amplitude and stress rate on hardening/softening behavior of Zircaloy-2 under asymmetric cyclic loading, at room temperature. It was observed that plastic strain accumulation increased with mean stress and stress amplitude; however, it decreased with stress rate. (author)

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.

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

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)

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.

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)

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

Reaction diffusion in chromium-zircaloy-2 system

International Nuclear Information System (INIS)

Xiang Wenxin; Ying Shihao

2001-01-01

Reaction diffusion in the chromium-zircaloy-2 diffusion couples is investigated in the temperature range of 1023 - 1123 K. Scanning electron microscope (SEM) and energy dispersive spectrum (EDS) were used to measure the thickness of the reaction layer and to determine the Zr, Fe and Cr concentration penetrate profile in reaction layer, respectively. The growth kinetics of reaction layer has been studied and the results show that the growth of intermetallic compound is controlled by the process of volume diffusion as the layer growth approximately obeys the parabolic law. Interdiffusion coefficients were calculated using Boltzmann-Matano-Heumann model. Calculated interdiffusion coefficients were compared with those obtained on the condition that Cr dissolves in Zr and merely forms dilute solid solution. The comparison indicates that Cr diffuses in dilute solid solution is five orders of magnitude faster than in Zr(Fe, Cr) 2 intermetallic compound

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

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

Parametric studies of cutting zircaloy-2 sheets with a laser beam

International Nuclear Information System (INIS)

Ghosh, S.; Badgujar, B.P.; Goswami, G.L.

1996-01-01

The highly reactive and pyrophoric nature of zirconium alloys limits the use of conventional thermal sources (e.g., plasma arc cutting, oxygen flame cutting, etc.) for the cutting and drilling of these alloys. In this context, a highly coherent laser beam provides a good alternative for the cutting and drilling. In the present paper, laser beam cutting of zircaloy-2 sheets of 1.1 mm and 0.74 mm thickness is performed using a 300 W average power pulsed Nd:YAG laser. Pulse energy, pulse repetition rate, nozzle gap, gas pressure and cutting speed were varied to give different laser cutting conditions. Metallographic study of the cut surfaces showed the presence of transformed beta phase in the heat affected zone (HAZ) near the cut surface. The microhardness value across the cut surface was also measured. It showed a gradual increase in microhardness from the base metal (160 VHN) towards the HAZ having a maximum value of 365 VHN. The results of parametric studies of the cutting indicated that, with proper selection of process parameters, very narrow cuts can be easily made in zircaloy-2 using a pulsed Nd:YAG laser with a saving in material and at a much faster rate than alternative processes such as plasma arc cutting and oxygen flame cutting

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

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.)

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

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

Identification of the zirconium hydrides metallography in zircaloy-2

International Nuclear Information System (INIS)

Garcia Gonzalez, F.

1968-01-01

Technique for the Identification of the zirconium hydrides in metallographic specimens have been developed. Microhardness, quantitative estimation and relative orientation of the present hydrides as well as grain size determination of the different Zircaloy-2 tube specimens have also been made. The specimens used were corrosion- tested in water during various periods of time at 300 degree castrating, prior to the metallographic examination. Reference specimens, as received, and heavily hydride specimens in a hydrogen atmosphere at 800 degree centigrees, have been used in the previous stages of the work. No difficulties have been met in this early stage of acquaintanceship with the zirconium hydrides. (Author) 5 refs

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

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

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)

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

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)

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

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

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

Irradiation effects on Fe distributions in zircaloy-2 and Zr-2.5Nb

International Nuclear Information System (INIS)

Zou, H.; Hood, G.M.; Roy, J.A.

1995-03-01

Irradiation of large-grained Zr-2.5Nb (ZN) and Zircaloy-2 (Zy) with 1.5 MeV Ar ions to a fluence of ∼ 10 20 /m 2 (≡ 10 dpa) at 50, 300 and 420 deg C leads to enhanced α-phase Fe levels of 250-1500 ppma, compared to equivalent non-irradiated state values of ∼ 70 ppma. In ZN the β-phase Fe levels fell from about 6000 to 3500 ppma: this result accords, qualitatively, with the loss of Fe from the β-phase following in-service neutron irradiation. Measurements on Zy showed that the Fe concentrations were higher near the specimen surfaces. Limited data for Ni distributions in Zy show similar (to Fe) behaviour. (author). 18 refs., 2 tabs

Measurements of the effective total and resonance absorption cross sections for zircaloy-2 and zirconium

Energy Technology Data Exchange (ETDEWEB)

Kocic, A; Markovic, V [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

1967-04-15

Zirconium and zircaloy-2 alloy, as constructive materials, have found wide application in reactor technology, especially in heavy water systems for two reasons: a) low neutron absorption cross section, b) good mechanical properties. The thickness of the zirconium and zircaloy-2 for different applications varies from several tenths of a millimeter to about ten millimeters. Therefore, to calculate reactor systems it is desirable to know the effective neutron absorption cross section for the range of thicknesses mention above. The thermal neutron cross sections for these materials are low and no appreciable variation of the effective neutron cross section occurs even for the largest thicknesses. However, this is not true for effective resonance absorption. On the other hand, due to the lack of detailed knowledge of the zirconium resonances, calculations of the effective resonance integrals cannot be performed. Therefore it is necessary to measure the effective total and resonance absorption cross section for zirconium (author)

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)

Report on the visit to Saclay related to the VISA-2 Project, February - March 1962; Izvestaj o poseti Saclayu u vezi projekta VISA-2, februar - mart 1962

Energy Technology Data Exchange (ETDEWEB)

Milasin, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-03-15

This report includes activities related to planned irradiation of samples (Zircaloy, stainless steels, beryllium) within VISA-2 project, namely welding, leak testing and preparing the samples for the irradiation capsules. Visits to nuclear center in Grenoble as well as factories which fabricate reactor materials are included.

Experimental determination of resonance absorption cross sections for Zircaloy-2 and zirconium

Energy Technology Data Exchange (ETDEWEB)

Kocic, A; Markovic, V [Boris Kidric Institute of Nuclear Sciences, Vinca, Beograd (Yugoslavia)

1968-05-15

The integral absorption cross section for the neutron spectrum and the thermal absorption cross section for zircaloy-2 have been determined using the pile oscillator technique. Using both values and a measured ratio of the epithermal to the thermal flux, the effective resonance integrals were obtained. After subtraction of the contributions for alloy and impurity elements, the effective resonance integrals for zirconium were evaluated. An extrapolated value of 0.91{+-}0.10 was obtained for the dilute integral. (author)

Fatigue limit of Zircaloy-2 under variable one-directional tension and temperature 300 deg C; Granica zamora zircaloy-2, pri cisto jednosmerno promenljivom opterecenju (A=1) na zatezanje i temperaturi 300 deg C

Energy Technology Data Exchange (ETDEWEB)

Spasic, Z; Simic, G [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1968-11-15

A vacuum chamber wad designed and constructed. It was suitable for study of materials at higher temperatures in vacuum or controlled atmospheres. Zircaloy-2 fatigue at 300 deg C in argon atmosphere was measured. Character of strain is variable one directional (A=1) tension. Obtained results are presented in tables and in the form of Veler's curve. The obtained fatigue limit was {sigma} - 15 kp/mm{sup 2}. The Locati method was allied as well and fatigue limit value obtained was 15,75 kp/mm{sup 2}. Error calculated in reference to the previous value obtained by classical methods was 5%. Konstruisana je i izvedena vakuum-komora koja se pokazala prikladna za izucavanje osobina materijala na povisenim temperaturama u vakuumu ili kontrolisanim atmosferama. Izvrseno je ispitivanje zamaranja Zircaloy-2 na temperaturi 300 deg C u atmosferi preciscenog argona. Karakter opterecenja je bio cisto jednosmerno promenljivo opterecenje (A=1) na zatezanje. Dobiveni rezultati su dati tabelarno i u obliku Velerove krive. Dobijena je granica zamora {sigma} = 15 kp/mm{sup 2}. Primenjen je i metod Locati-a za priblizno odredjivanje granice zamora i dobijena je vrednost 15,75 kp/mm{sup 2}. Greska u odnosu na prethodnu granicu zamora dobijenu klasicnim metodom iznosi 5% (author)

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

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)

Electrochemical corrosion of Zircaloy-2 under PWR water chemistry but at room temperature

International Nuclear Information System (INIS)

Waheed, Abdel-Aziz Fahmy; Kandil, Abdel-Hakim Taha; Hamed, Hani M.

2016-01-01

Highlights: • There is no simple relation between the corrosion rate and LiOH concentration. • At low concentration, 100 ppm Li, an increase of the rate is due to the pH impact. • LiOH in concentrated solution led to accelerated corrosion by pH effect and porosity. • Boron abates the lithium effect by pH neutralizing and participation in the corrosion. - Abstract: Electrochemical corrosion of Zircaloy-2 was tested at room temperature in lithium hydroxide (LiOH) concentrations that ranged from 2.2 to 7000 ppm and boric acid (H 3 BO 3 ) concentrations that ranged from 50 to 4000 ppm. Following the corrosion experiments, the oxide films of specimens were examined by SEM to examine the oxide existence. LiOH concentrations as high as 1 M (7000-ppm lithium) can lead to significantly increased electrochemical corrosion rate. It is suggested that the accelerated corrosion in concentrated solution is caused by the synergetic effect of LiOH, pH and porosity generation. In solutions containing 100 ppm of lithium, the presence of boron had an ameliorating effect on the corrosion rates of Zircaloy-2. Similar to acceleration of corrosion by lithium, the inhibition by boron is due to a combined effect of pH neutralizing and its participation in the corrosion process.

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

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.

Temperature escalation in PWR fuel rod simulators due to the zircaloy/steam reaction: Tests ESSI-1,2,3

International Nuclear Information System (INIS)

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

1983-08-01

This report discusses the test conduct, results, and posttest appearance of three scoping tests (ESSI-1,2,3) investigating temperature escalation in zircaloy clad fuel rods. The experiments are part of an out-of-pile program using electrically heated fuel rod simulators to investigate PWR fuel element behavior up to temperatures of 2000 0 C. These experiments are part of the PNS Severe Fuel Damage Program. The temperature escalation is caused by the exothermal zircaloy/steam reaction, whose reaction rate increases exponentially with the temperature. The tests were performed using different initial oxide layers as a major parameter, obtained by varying the heatup rates and steam exposure times. (orig./RW) [de

Influence of shielding gas pressure on welding characteristics in CO2 laser-MIG hybrid welding process

Science.gov (United States)

Chen, Yanbin; Lei, Zhenglong; Li, Liqun; Wu, Lin

2006-01-01

The droplet transfer behavior and weld characteristics have been investigated under different pressures of shielding gas in CO2 laser and metal inert/active gas (laser-MIG) hybrid welding process. The experimental results indicate that the inherent droplet transfer frequency and stable welding range of conventional MIG arc are changed due to the interaction between CO2 laser beam and MIG arc in laser-MIG hybrid welding process, and the shielding gas pressure has a crucial effect on welding characteristics. When the pressure of shielding gas is low in comparison with MIG welding, the frequency of droplet transfer decreases, and the droplet transfer becomes unstable in laser-MIG hybrid welding. So the penetration depth decreases, which shows the characteristic of unstable hybrid welding. However, when the pressure of shielding gas increases to a critical value, the hybrid welding characteristic is changed from unstable hybrid welding to stable hybrid welding, and the frequency of droplet transfer and the penetration depth increase significantly.

Observations on the ductility of zircaloy-2 under simultaneous tension and bending

International Nuclear Information System (INIS)

Pettersson, K.

1975-01-01

The ductility of Zircaloy-2 in creep-fatigue interaction tests has been found to exceed the ductility in separate tensile tests. It was shown that the increase of ductility was due to either the suppression of the localized shear band instability causing final failure in a tensile test, or because the hydrostatic tension-shear stress ratio in the creep-fatigue test is lower than in the tensile test. Possible applications of the ductility increase in forming operations are suggested. (author)

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

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

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

A regression approach for zircaloy-2 in-reactor creep constitutive equations

International Nuclear Information System (INIS)

Yung Liu, Y.; Bement, A.L.

1977-01-01

In this paper the methodology of multiple regressions as applied to zircaloy-2 in-reactor creep data analysis and construction of constitutive equation are illustrated. While the resulting constitutive equation can be used in creep analysis of in-reactor zircaloy structural components, the methodology itself is entirely general and can be applied to any creep data analysis. From data analysis and model development point of views, both the assumption of independence and prior committment to specific model forms are unacceptable. One would desire means which can not only estimate the required parameters directly from data but also provide basis for model selections, viz., one model against others. Basic understanding of the physics of deformation is important in choosing the forms of starting physical model equations, but the justifications must rely on their abilities in correlating the overall data. The promising aspects of multiple regression creep data analysis are briefly outlined as follows: (1) when there are more than one variable involved, there is no need to make the assumption that each variable affects the response independently. No separate normalizations are required either and the estimation of parameters is obtained by solving many simultaneous equations. The number of simultaneous equations is equal to the number of data sets, (2) regression statistics such as R 2 - and F-statistics provide measures of the significance of regression creep equation in correlating the overall data. The relative weights of each variable on the response can also be obtained. (3) Special regression techniques such as step-wise, ridge, and robust regressions and residual plots, etc., provide diagnostic tools for model selections

Plasma Arc Augmented CO2 laser welding

DEFF Research Database (Denmark)

Bagger, Claus; Andersen, Mikkel; Frederiksen, Niels

2001-01-01

In order to reduce the hardness of laser beam welded 2.13 mm medium strength steel CMn 250, a plasma arc has been used simultaneously with a 2.6 kW CO2 laser source. In a number of systematic laboratory tests, the plasma arc current, plasma gas flow and distance to the laser source were varied...... with all laser parameters fixed. The welds were quality assessed and hardness measured transversely to the welding direction in the top, middle and root of the seam. In the seams welded by laser alone, hardness values between 275 and 304 HV1 were measured, about the double of the base material, 150 HV1...

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

TEM/STEM study of Zircaloy-2 with protective FeAl(Cr) layers under simulated BWR environment and high-temperature steam exposure

Science.gov (United States)

Park, Donghee; Mouche, Peter A.; Zhong, Weicheng; Mandapaka, Kiran K.; Was, Gary S.; Heuser, Brent J.

2018-04-01

FeAl(Cr) thin-film depositions on Zircaloy-2 were studied using transmission electron microscopy (TEM) and scanning transmission electron microscopy (STEM) with respect to oxidation behavior under simulated boiling water reactor (BWR) conditions and high-temperature steam. Columnar grains of FeAl with Cr in solid solution were formed on Zircaloy-2 coupons using magnetron sputtering. NiFe2O4 precipitates on the surface of the FeAl(Cr) coatings were observed after the sample was exposed to the simulated BWR environment. High-temperature steam exposure resulted in grain growth and consumption of the FeAl(Cr) layer, but no delamination at the interface. Outward Al diffusion from the FeAl(Cr) layer occurred during high-temperature steam exposure (700 °C for 3.6 h) to form a 100-nm-thick alumina oxide layer, which was effective in mitigating oxidation of the Zircaloy-2 coupons. Zr intermetallic precipitates formed near the FeAl(Cr) layer due to the inward diffusion of Fe and Al. The counterflow of vacancies in response to the Al and Fe diffusion led to porosity within the FeAl(Cr) layer.

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

Ductility loss of ion-irradiated zircaloy-2 in iodine

International Nuclear Information System (INIS)

Shimada, M.; Terasawa, M.; Yamamoto, S.; Kamei, H.; Koizumi, K.

1981-01-01

An ion bombardment simulation technique for neutron irradiation was applied to 'thick' materials to study the effect of radiation damage on the ductility change in Zircaloy-2 in an iodine environment. Specimens were prepared from actual cladding tubes and, prior to the irradiation, they were heat-treated in vacuo at 450, 580, and 700/degree/C for 2 h. Irradiation was performed by 52-MeV alpha particles up to the 0.32 displacements per atom (dpa) at 340/degree/C. Ductility loss begins to appear after 0.03 dpa irradiation, both in iodine and argon gas environments. The iodine presence resulted in ductility reduction, compared with the argon result in all irradiation dose ranges examined. The stress applied during irradiation caused ductility loss to commence at lower dosage than in the case of stress-free irradiation. These results are discussed in relation to the existing stress corrosion cracking models

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

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)

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)

Report on the visit to Saclay related to the VISA-2 Project, February - March 1962

International Nuclear Information System (INIS)

Milasin, N.

1963-01-01

This report includes activities related to planned irradiation of samples (Zircaloy, stainless steels, beryllium) within VISA-2 project, namely welding, leak testing and preparing the samples for the irradiation capsules. Visits to nuclear center in Grenoble as well as factories which fabricate reactor materials are included

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)

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)

SSMS near surface analysis of B in irradiated Zircaloy-2: ion implantation standards as a calibration technique

International Nuclear Information System (INIS)

Christie, W.H.; Carter, J.A.; Eby, R.E.; Landau, L.; Musick, W.R.

1980-01-01

Purpose of this study was to determine the amount of 10 B contamination on the surface of Zircaloy-2 clad irradiated fuel elements that had been stored in an aqueous solution containing 5000 wt. ppM enriched B. SMSS indicated that the contamination was less than 0.06 μg/cm 2

CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

International Nuclear Information System (INIS)

Haste, T.J.

1982-07-01

The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

Metallographic Study of the Isothermal Transformation of Beta Phase in Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Oestberg, G

1960-06-15

Observations of the structure of commercial zircaloy-2 have been made in the microscope showing that the high temperature beta phase is transformed isothermally at lower temperatures into alpha plus secondary precipitate. The alpha occurs mainly as Widmanstaetten plates developed by a shear mechanism. The secondary precipitate is formed from the beta - alpha structure at the phase boundary between these phases. This precipitation of particles of secondary phase occurs on account of a eutectoid reaction, alpha also being formed. A time-temperature transformation diagram has been constructed from the observations.

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)

Deformation texture and microtexture development in zircaloy-2

International Nuclear Information System (INIS)

Vanitha, C.; Kiran Kumar, M.; Samajdar, I.; Vishvanathan, N.N.; Dey, G.K.; Tewari, R.; Srivastava, D.; Banerjee, S.

2002-01-01

In the present study, two starting materials used were as-cast Zircaloy-2 with random texture and the finished tube with relatively stronger starting texture. Specimens of the alloys were hot rolled to various strains at different temperature. The texture measurement was carried out and was represented in the form of Orientation Distribution Function which showed a sluggish texture development on high temperature deformation. In the case of as cast alloy with increase in strain at a constant deformation temperature, development in the texture was significant. Upon increasing the working temperature, rate of the overall texture development has been found to reduce. This could be due to reduced slip-twin activities, recovery or due to recrystallization. Microstructural and relative hardening studies were carried out for understanding the mechanisms of deformation texture developments at warm and hot working stages. In the case of finished tube having initially strong texture exhibited slower development in texture on warm and hot rolling. (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)

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

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.

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

Studies on CO2-laser Hybrid-Welding of Copper

DEFF Research Database (Denmark)

Nielsen, Jakob Skov; Olsen, Flemming Ove; Bagger, Claus

2005-01-01

CO2-laser welding of copper is known to be difficult due to the high heat conductivity of the material and the high reflectivity of copper at the wavelength of the CO2-laser light. THis paper presents a study of laser welding of copper, applying laser hybrid welding. Welding was performed as a hy...

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

Effect of current density on the anodization of zircaloy-2

International Nuclear Information System (INIS)

Bhaskar Reddy, P.; Panasa Reddy, A.

2005-01-01

The effect of current density on the kinetics of anodization of Zircaloy-2 in 0.1 M potassium tartarate have been studied at various constant current densities ranging from 2 to 10 mA.cm -2 and at room temperature to investigate the exponential dependence of ionic current density on the field across the oxide. The rate of anodic film formation (dV/dt), the current efficiency the differential field of formation (F) and the ionic current density (i i ) were calculated. It was found that all these parameters were increased with increase of current density. The induction period was decreased with the increase of current density. It was also found that the plot of log (ionic current density) vs differential field gave fairly a linear relationship. The kinetic parameters, half jump distance (a) and height of the energy barrier (W) were calculated. (author)

Selected Welding Techniques, Part 2

National Research Council Canada - National Science Library

1964-01-01

Partial contents: CONVENTIONAL WELD JOINTS VERSUS BUTT JOINTS IN 1-INCH ALUMINUM PLATE, SPECIAL WELD JOINT PREPARATION, UPSET METAL EDGES FOR INCREASED WELD JOINT STRENGTH, OUT-OF-POSITION WELDING OF HEAVY GAGE...

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.)

An assessment of the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes of PHWR

International Nuclear Information System (INIS)

Sah, D.N.

1992-01-01

In view of the deleterious effect of hydriding on the operating life of zircaloy-2 pressure tubes in PHWRs there is an urgent need for the assessment of the status of the pressure tubes with respect to corrosion and hydrogen pick-up in the operating PHWRs. A model has been developed for analysing the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes under reactor operating conditions. This model predicts the axial profiles of oxide layer thickness and hydrogen pick-up in the pressure tubes as a function of the operating time of the reactor. The prediction of hydrogen pick-up by the model in the F-10 pressure tube of RAPS-I have been found to be in good agreement with the measured value of hydrogen content. This report gives a brief description of the model and its predictions on the present status of hydrogen pick-up in the pressure tubes of lead reactor RAPS-II. (author). 6 refs., 5 figs., 2 tabs

The formation and characteristics of hydride blisters in c.w. Zircaloy-2 pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Price, E G [ed.

1994-09-01

Under the auspices of the IAEA, a consultants` meeting was arranged in Vienna, 1994 July 25-29, at which a Canadian delegation, consisting of AECL and Ontario Hydro Technologies personnel, presented information on their knowledge of the behaviour of hydride blisters in Zircaloy-2 pressure tubes. This document contains the 10 papers presented by the Canadian delegation to the meeting. It is believed that they represent a good reference document on hydride blister phenomena.

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)

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

A pneumatic bellows-driven setup for controlled-distance electrochemical impedance measurements of Zircaloy-2 in simulated BWR conditions

International Nuclear Information System (INIS)

Arilahti, E.; Bojinov, M.; Hansson-Lyyra, L.

2004-01-01

This paper describes a novel pneumatic bellows-driven arrangement designed for controlled distance electrochemistry (CDE) measurements. The feasibility of the new arrangement has been verified by performing contact electric impedance measurements to study corrosion of Zircaloy-2 in a re-circulation loop simulating the BWR conditions. Until now, the measurements have been carried out using a step-motor driven controlled-distance electrochemistry (CDE) arrangement. The electrical and electrochemical properties of the pre transition oxide on Zircaloy-2 determined from these measurements were in good agreement with those estimated from measurements with a step-motor driven CDE. Furthermore, the results indicate that the bellows-driven CDE device is less sensitive to the contact pressure variation than the step-motor driven arrangement. This property combined with the bellows driven displacement mechanism provides a clear advantage for future in-core corrosion studies of fuel cladding materials. (Author)

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

Microstructural examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

2005-01-01

Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples 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), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

Comparisons between TiO2- and SiO2-flux assisted TIG welding processes.

Science.gov (United States)

Tseng, Kuang-Hung; Chen, Kuan-Lung

2012-08-01

This study investigates the effects of flux compounds on the weld shape, ferrite content, and hardness profile in the tungsten inert gas (TIG) welding of 6 mm-thick austenitic 316 L stainless steel plates, using TiO2 and SiO2 powders as the activated fluxes. The metallurgical characterizations of weld metal produced with the oxide powders were evaluated using ferritoscope, optical microscopy, and Vickers microhardness test. Under the same welding parameters, the penetration capability of TIG welding with TiO2 and SiO2 fluxes was approximately 240% and 292%, respectively. A plasma column made with SiO2 flux exhibited greater constriction than that made with TiO2 flux. In addition, an anode root made with SiO2 flux exhibited more condensation than that made with TiO2 flux. Results indicate that energy density of SiO2-flux assisted TIG welding is higher than that of TiO2-flux assisted TIG welding.

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.

CO2 and diode laser welding of AZ31 magnesium alloy

International Nuclear Information System (INIS)

Zhu Jinhong; Li Lin; Liu Zhu

2005-01-01

Magnesium alloys are being increasingly used in automotive and aerospace structures. Laser welding is an important joining method in such applications. There are several kinds of industrial lasers available at present, including the conventional CO 2 and Nd:YAG lasers as well as recently available high power diode lasers. A 1.5 kW diode laser and a 2 kW CO 2 laser are used in the present study for the welding of AZ31 alloys. It is found that different welding modes exist, i.e., keyhole welding with the CO 2 laser and conduction welding with both the CO 2 and the diode lasers. This paper characterizes welds in both welding modes. The effect of beam spot size on the weld quality is analyzed. The laser processing parameters are optimized to obtain welds with minimum defects

46 CFR 2.75-70 - Welding procedure and performance qualifications.

Science.gov (United States)

2010-10-01

... 46 Shipping 1 2010-10-01 2010-10-01 false Welding procedure and performance qualifications. 2.75... for Construction Personnel § 2.75-70 Welding procedure and performance qualifications. (a) Welding... requirements for the welding of pressure piping, boilers, pressure vessels, and nonpressure vessel type tanks...

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.

Cyclic softening in annealed Zircaloy-2: Role of edge dislocation dipoles and vacancies

Science.gov (United States)

Sudhakar Rao, G.; Singh, S. R.; Krsjak, Vladimir; Singh, Vakil

2018-04-01

The mechanism of cyclic softening in annealed Zircaloy-2 at low strain amplitudes under strain controlled fatigue at room temperature is rationalized. The unusual softening due to continuous decrease in the phenomenological friction stress is found to be associated with decrease in the resistance against movement of dislocations because of the formation and easy glide of pure edge dislocation dipoles and consequent decrease in friction stress from reduction in the shear modulus. Positron annihilation spectroscopy data strongly support the increase in edge dislocation density containing jogs, from increased positron trapping and increase in annihilation lifetime.

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.

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)

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 on the Microstructure Evolution and Toughness of TIG Weld Metal of 25Cr2Ni2MoV Steel after Post Weld Heat Treatment

Directory of Open Access Journals (Sweden)

Xia Liu

2018-03-01

Full Text Available The microstructure and toughness of tungsten inert gas (TIG backing weld parts in low-pressure steam turbine welded rotors contribute significantly to the total toughness of the weld metal. In this study, the microstructure evolution and toughness of TIG weld metal of 25Cr2Ni2MoV steel low-pressure steam turbine welded rotor under different post-weld heat treatment (PWHT conditions are investigated. The fractography and microstructure of weld metal after PWHT are characterized by optical microscope, SEM, and TEM, respectively. The Charpy impact test is carried out to evaluate the toughness of the weld. The optical microscope and SEM results indicate that the as-welded sample is composed of granular bainite, acicular ferrite and blocky martensite/austenite (M-A constituent. After PWHT at 580 °C, the blocky M-A decomposes into ferrite and carbides. Both the number and size of precipitated carbides increase with holding time. The impact test results show that the toughness decreases dramatically after PWHT and further decreases with holding time at 580 °C. The precipitated carbides are identified as M23C6 carbides by TEM, which leads to the dramatic decrease in the toughness of TIG weld metal of 25Cr2Ni2MoV steel.

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)

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

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.

Crystal plasticity modeling of irradiation growth in Zircaloy-2

Science.gov (United States)

Patra, Anirban; Tomé, Carlos N.; Golubov, Stanislav I.

2017-08-01

A physically based reaction-diffusion model is implemented in the visco-plastic self-consistent (VPSC) crystal plasticity framework to simulate irradiation growth in hcp Zr and its alloys. The reaction-diffusion model accounts for the defects produced by the cascade of displaced atoms, their diffusion to lattice sinks and the contribution to crystallographic strain at the level of single crystals. The VPSC framework accounts for intergranular interactions and irradiation creep, and calculates the strain in the polycrystalline ensemble. A novel scheme is proposed to model the simultaneous evolution of both, number density and radius, of irradiation-induced dislocation loops directly from experimental data of dislocation density evolution during irradiation. This framework is used to predict the irradiation growth behaviour of cold-worked Zircaloy-2 and trends compared to available experimental data. The role of internal stresses in inducing irradiation creep is discussed. Effects of grain size, texture and external stress on the coupled irradiation growth and creep behaviour are also studied and compared with available experimental data.

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.

Microstructural examination of Zr-2.5%Nb alloy welds made by pulsed Nd:YAG laser and TIG welding technique

International Nuclear Information System (INIS)

Bhatt, R.B.; Varma, P.V.S.; Panakkal, J.P.; Srivastava, D.; Dey, G.K.

2009-01-01

The paper describes the weld microstructure of Zr-2.5%Nb alloy material. Bead on plate welds were made using pulsed Nd:YAG laser and TIG welding technique at different parameters. These welds were characterized at macro and microstructural level. Weld pools of Pulsed Laser and TIG welds were not resolved by optical microscopy. SEM too did not reveal much. Orientation imaging microscopy could reveal the presence of fine martensite. It was observed that microstructure is very sensitive to welding parameters. Microhardness studies suggested formation of martensite in the weld pool. It was also observed that laser welds had very sharp weld pool boundary as compared to TIG welds. Variation in microhardness of the weldment is seen and is influenced by overlapping of weld spots causing thermal treatment of previously deposited spots. (author)

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

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

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)

Stochastic model of texture dependence of iodine SCC susceptibility of a zircaloy-2 alloy

International Nuclear Information System (INIS)

Hirao, Keiichi; Yamane, Toshimi; Nakajima, Shinichi; Node, Shunsaku; Fujisawa, Takashi; Minamino, Yoritoshi

1991-01-01

Effects of textures on statistical parameters of tensile elongations in stress corrosion cracking (SCC) of zircaloy-2 using a slow strain rate test (SSRT) method have been investigated by Weibull distribution method based on stochastic process theory. The SCC is analyzed by assuming a probabilistic state transition model. Tensile directions of test pieces were prepared parallel, 45deg and perpendicular to rolling direction of the sheet. The test pieces in evacuated silica tubes were annealed at 1073K for 7.2x10 3 s, and then quenched into ice water. The annealed pieces with tilt angle α between tensile direction and a basal plane {0001} were 0, 18 and 25deg respectively. The tensile elongations of zircaloy-2 in SCC using the SSRT method are found to obey the single Weibull distribution with location parameters, and the SCC phenomena can be described by the Weibull distribution based on the stochastic process. The values of scale parameter η decrease with the tilt angle α, and the SCC susceptibility can be indicated by the values of scale parameter η. The texture dependence of the values of shape parameters m shows the changes of corrosion process in iodine solution and deformation system in air which are observed in the SSRT. The mechanism of decrement in the SCC susceptibility changes with the tilt angle α. The SCC under SSRT method is found to obey the model of probabilistic state transition. The constant load SCC process which obey the model of probabilistic state transition, is found to be effective for estimation of accelerated SCC condition. (author)

Diffusion welding of ZrO2 solid electrolyte cells

International Nuclear Information System (INIS)

Schaefer, W.; Schmidberger, R.

1980-01-01

Zirconia based solid-electrolyte-cells can be applied as electrolysis-cells or fuel cells at high temperatures. Scaling up to technical aggregates must be realized by a gastight electrical series-connection of many tubular single cells. A suitable process for connecting single cells is diffusion welding. Starting materials were sintered zirconia-tubes (16 mm diameter, 10 mm length) and gastight interconnecting rings (16 mm diameter, 0.5-2mm length) from gold, platinum or electrically conducting mixed oxides. ZrO 2 -tubes and interconnecting rings were mounted in alternating sequence and diffusion welded under axial pressure at high temperatures. From economic reasons noble metals cannot be used for technical aggregates. The developments were therefore concentrated on the connection with mixed oxides. Optimized welding parameters are: 1400-1500 0 C welding temperature, 2 hours welding time and an axial pressure of approximately 1 Nmm 2 . Up to now gastight tubes consisting of 20 single cells were preparated by diffusion-welding in one step. The process will be further developed for the production of 50-cell-tubes with a total length of about 60 cm. (orig.) [de

Fatigue limit of Zircaloy-2 under variable one-directional tension and temperature 300 deg C

International Nuclear Information System (INIS)

Spasic, Z.; Simic, G.

1968-11-01

A vacuum chamber wad designed and constructed. It was suitable for study of materials at higher temperatures in vacuum or controlled atmospheres. Zircaloy-2 fatigue at 300 deg C in argon atmosphere was measured. Character of strain is variable one directional (A=1) tension. Obtained results are presented in tables and in the form of Veler's curve. The obtained fatigue limit was σ - 15 kp/mm 2 . The Locati method was allied as well and fatigue limit value obtained was 15,75 kp/mm 2 . Error calculated in reference to the previous value obtained by classical methods was 5% [sr

Influence of texture on fracture toughness of zircaloy cladding

Energy Technology Data Exchange (ETDEWEB)

Grigoriev, V. [Studsvik Material AB, Nykoeping (Sweden); Andersson, Stefan [Royal Inst. of Tech., Stockholm (Sweden)

1997-06-01

The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill`s theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture. With a 2 page summary in Swedish. 32 refs, 18 figs.

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

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.)

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)

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

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)

Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

International Nuclear Information System (INIS)

Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F.

2000-01-01

The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

Recent observations on the evolution of secondary-phase particles in zircaloy-2 under irradiation in a BWR to high burn-up

Energy Technology Data Exchange (ETDEWEB)

Abolhassani, S.; Graber, T.; Gavillet, D.; Groeschel, F

2000-07-01

The influence of radiation on the corrosion of the fuel claddings in a Light Water Reactor (LWR) has been the subject of many investigations, and different aspects of the overall phenomena have been studied by different techniques. Analysis of the evolution of Secondary-Phase Particles (SPPs) for different periods of immersion of the cladding in the reactor enables the rate of corrosion to the structure of the material to be correlated. In the case of Zircaloy-2 in a Boiling Water Reactor (BWR), SPPs are dissolved under irradiation, and their dissolution affects the rate of oxidation and other correlated phenomena. In recent studies, the Zircaloy-2 in claddings loaded in the Leibstadt BWR are analysed after one, three and five cycles. Results are presented, and give an account of the changes which occurred in the materials under irradiation. (authors)

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

Combined effects of radiation damage and hydrides on the ductility of Zircaloy-2

International Nuclear Information System (INIS)

Wisner, S.B.; Adamson, R.B.

1998-01-01

Interest remains high regarding the effects of zirconium hydride precipitates on the ductility of reactor Zircaloy components, particularly in irradiated material. Previous studies have reported that ductility reductions are much greater at room temperature compared to reactor component temperatures. It is often concluded that the effects of irradiation dominate the ductility reduction observed in test specimens, although there is no consensus as to whether hydriding effects are additive. Many of the tests reported in the literature are difficult to interpret due to variations in test specimen geometry and material history. In this paper, we present the results of an experimental program aimed at clearly describing the combined effects of irradiation and hydriding on ductility parameters under conditions of a realistic test specimen design and well characterized hydride content, distribution and orientation. Experiments were conducted at 295 and 605 K, respectively on Zircaloy-2 tubing segments containing 10-800 ppm hydrogen and neutron fluences between 0.9 x 10 25 nm -2 (E>1 MeV). Tests utilized the well proven localized ductility specimen which applies plane strain tension in the hoop direction of the tubing segment. In all cases, hydrides were also oriented in the hoop or circumferential direction and were uniformly distributed across the tubing wall. Results indicate that at 605 K, the ductility of irradiated material was almost independent of hydride content, retaining above 4% uniform elongation and 25% reduction in an area for the highest fluences and hydrogen contents. Even at 295 K, measurable ductility was retained for irradiated material with up to 600 ppm hydrogen. In the paper, results of fractographic analyses and strain rate are also discussed

Zircaloy cladding degradation under repository conditions

International Nuclear Information System (INIS)

Santanam, L.; Raghavan, S.; Chin, B.A.

1990-12-01

Creep, a potential degradation mechanism of Zircaloy cladding after repository disposal of spent nuclear fuel, has been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. Maximum allowable temperatures are 340 degree C (613 K) for typically stressed rods (70--100 MPa) and 300 degree C (573 K) for highly stressed rods (140--160 MPa). 10 refs., 2 figs

Simulation of the interaction between uranium dioxide and zircaloy

International Nuclear Information System (INIS)

Denis, A.; Garcia, E.A.

1984-01-01

The code solves the oxygen diffusion equations of the five phases formed during the UO 2 /Zircaloy interaction, using an implicit finite difference method with parabolic interpolation at the interfaces. Uranium and Zirconium mass conservation are considered. The code gives a good simulation of the experimental results for isothermal conditions. (orig.)

Appropriate welding conditions of temper bead weld repair for SQV2A pressure vessel steel

International Nuclear Information System (INIS)

Mizuno, R.; Matsuda, F.; Brziak, P.; Lomozik, M.

2004-01-01

Temper bead welding technique is one of the most important repair welding methods for large structures for which it is difficult to perform the specified post weld heat treatment. In this study, appropriate temper bead welding conditions to improve the characteristics of heat affected zone (HAZ) are studied using pressure vessel steel SQV2A corresponding to ASTM A533 Type B Class 1. Thermal/mechanical simulator is employed to give specimens welding thermal cycles from single to quadruple cycle. Charpy absorbed energy and hardness of simulated CGHAZ by first cycle were degraded as compared with base metal. Improvability of these degradations by subsequent cycles is discussed and appropriate temper bead thermal cycles are clarified. When the peak temperature lower than Ac1 and near Ac1 in the second thermal cycle is applied to CGAHZ by first thermal cycle, the characteristics of CGHAZ improve enough. When the other peak temperatures (that is, higher than Ac1) in the second thermal cycle are applied to the CGHAZ, third or more thermal cycle temper bead process should be applied to improve the properties. Appropriate weld condition ranges are selected based on the above results. The validity of the selected ranges is verified by the temper bead welding test. (orig.)

High-pressure hydriding of Zircaloy

International Nuclear Information System (INIS)

Kim, Y.S.

1996-01-01

The hydriding characteristics of Zircaloy-2(Zry), sponge zirconium (as a liner on Zry plate), and crystal-bar zirconium exposed to pure H 2 at 0.1 MPa or 7 MPa and 400 C were determined in a thermogravimetric apparatus. The morphology of the hydrided specimens was also examined by optical microscopy. For all specimen types, the rate of hydriding in 7 MPa H 2 was two orders of magnitude greater than in 0.1 MPa H 2 . For Zry, uniform bulk hydriding was revealed by hydride precipitates at room temperature and on one occasion, a sunburst hydride. In addition, all specimen types exhibited a hydride surface layer. In a duplex Zry/sponge-Zr specimen, Zry is more heavily hydrided than the sponge Zr layer. (orig.)

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.

Plastic deformation and fracture behavior of zircaloy-2 fuel cladding tubes under biaxial stress

International Nuclear Information System (INIS)

Maki, Hideo; Ooyama, Masatosi

1975-01-01

Various combinations of biaxial stress were applied on five batches of recrystallized zircaloy-2 fuel cladding tubes with different textures; elongation in both axial and circumferential directions of the specimen was measured continuously up to 5% plastic deformation. The anisotropic theory of plasticity proposed by Hill was applied to the resulting data, and anisotropy constants were obtained through the two media of plastic strain loci and plastic strain ratios. Comparison of the results obtained with the two methods proved that the plastic strain loci provide data that are more effective in predicting quantitatively the plastic deformation behavior of the zircaloy-2 tubes. The anisotropy constants change their value with progress of plastic deformation, and judicious application of the effective stress and effective strain obtained on anisotropic materials will permit the relationship between stress and strain under various biaxialities of stresses to be approximated by the work hardening law. The test specimens used in the plastic deformation experiments were then stressed to fracture under the same combination of biaxial stress as in the proceeding experiments, and the deformation in the fractured part was measured. The result proved that the tilt angle of the c-axis which serves as the index of texture is related to fracture ductility under biaxial stress. Based on this relationship, it was concluded that material with a tilt angle ranging from 10 0 to 15 0 is the most suitable for fuel cladding tubes, from the viewpoint of fracture ductility, at least in the case of unirradiated material. (auth.)

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

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.

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)

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.)

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

Delayed hydride cracking behavior for zircaloy-2 plate

International Nuclear Information System (INIS)

Mills, J.W.; Huang, F.H.

1991-01-01

The delayed hydride cracking (DHC) behaviour for Zircaloy-2 plate was characterized at temperatures ranging from 300 to 550 o F. Specimens with a longitudinal (T-L) orientation exhibited a classic two-stage DHC response. At K values slightly above the threshold level (K th ), crack-growth rates increased dramatically with increasing K values (stage I). The K th value was found to be 11 and 14 ksi√ in at 400 and 500 o F. At high K values (stage II), cracking rates were relatively insensitive to applied K levels. Stage II crack growth was a thermally activated process described by an Arrhenius-type relationship with an activation energy of 65 kJ/mol. This energy level agreed with the theoretical activation energy for hydrogen diffusion into the triaxial stress field ahead of a crack. Above a critical temperature (300 o F), an overtemperature cycle was required to initiate DHC. The magnitude of the thermal excursion required to initiate cracking was found to increase at higher test temperatures. Specimens with a transverse(L-T) orientation showed a very low sensitivity to DHC because of an unfavorable crystallographic orientation for hydride reorientation. Metallographic and fractographic examinations were performed to understand the DHC mechanism. (author)

Hydraulic burst tests at elevated temperatures on Zircaloy cladding from fuel rods irradiated in the Winfrith SGHWR

International Nuclear Information System (INIS)

Garlick, A.; Hindmarch, P.

1980-09-01

Closed-end hydraulic burst tests have been carried out at 613K on lengths of cladding cut from fuel rods that had been irradiated in the SGHWR to 25 n/m 2 . The effects of reactor exposure on the mechanical properties of the Zircaloy cladding, initially in the stress-relieved and fully recrystallised conditions, have been evaluated from measurements of the 0.2% proof stress, the ultimate burst stress, the total circumferential elongation and the reduction in wall thickness at fracture. It is shown that after irradiation, the measured strength properties of stress-relieved cladding remained higher than for that in the fully recrystallised condition, although the large differences observed before irradiation were considerably reduced. The irradiation-induced increase in proof stress measured during these tests was compared with US results from uniaxial tensile tests and, after correcting for the effect of stress-ratio, it is concluded that close agreement exists between the two sets of data for Zircaloy in the fully recrystallised condition. In contrast, the agreement for stress-relieved Zircaloy is less good, although the maximum increase in proof stress after high neutron doses for this material is similar for data from the two sources. After irradiation, the ductility of fully recrystallised Zircaloy remained higher than that of stress-relieved material and there was no evidence to suggest that a serious loss of ductility had occurred for Zircaloy in either condition of heat-treatment as a result of reactor exposure. (author)

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

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)

Dictionary: Welding, cutting and allied processes. Pt. 2

International Nuclear Information System (INIS)

Kleiber, A.W.

1987-01-01

The dictionary contains approximately 40 000 entries covering all aspects of welding technology. It is based on the evaluation of numerous English, American and German sources. This comprehensive and up to date dictionary will be a reliable and helpful aid in evaluation and translating. The dictionary covers the following areas: Welding: gas welding, arc welding, gas shielded welding, resistance welding, welding of plastics, special welding processes; Cutting: flame cutting, arc cutting and special thermal cutting processes; Soldering: brazing and soldering; Other topics: thermal spraying, metal to metal adhesion, welding filler materials and other consumables, test methods, plant and equipment, accessories, automation, welding trade, general welding terminology. (orig./HP) [de

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

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)

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 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

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

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

Studies of irradiated zircaloy fuel sheathing using XPS

Energy Technology Data Exchange (ETDEWEB)

Chan, P K; Irving, K G [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Hocking, W H; Duclos, A M; Gerwing, A F [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-31

The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO{sub 2}) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs.

Studies of irradiated zircaloy fuel sheathing using XPS

International Nuclear Information System (INIS)

Chan, P.K.; Irving, K.G.; Hocking, W.H.; Duclos, A.M.; Gerwing, A.F.

1995-01-01

The preliminary results reported here support the hypothesis that CANLUB graphite coating reduces the rate at which oxygen can react with fuel sheathing. X-ray photoelectron spectroscopic (XPS) characterization of Zircaloy sheathing obtained from extended-burnup Bruce-type elements (BDL-406-XY (555 MW.h/kgU) and BDL-406-AAH (731 MW.h/kgU)) irradiated in NRU indicates that CANLUB may reduce fuel sheath oxidation, and hence that fission-liberated oxygen may remain in the fuel. Chemical shifts in the Zr 3d spectra suggest that a stoichiometric (ZrO 2 ) oxide film was formed only on Zircaloy in direct contact with fuel. Particulate fuel adhering to the sheath was also determined to be systematically more oxidized on surfaces with CANLUB than on those without it. The unique association of tin on sheathing specimens with the non-CANLUB-coated specimens might also suggest that the tin had segregated from the sheathing. It must be emphasized that further experiments are required to better define the effect of CANLUB on fuel oxidation. (author). 14 refs., 1 tab., 3 figs

Influence of texture on fracture toughness of zircaloy cladding

International Nuclear Information System (INIS)

Grigoriev, V.; Andersson, Stefan

1997-06-01

The correlation between texture and fracture toughness of Zircaloy 2 cladding has been investigated in connection with axial cracks in fuel rods. The texture of the cladding determines the anisotropy of plasticity of the cladding which, in turn, should influence the strain conditions at the crack-tip. Plastic strains in the cladding under uniaxial tension were characterised by means of the anisotropy constants F, G and H calculated according to Hill's theory. Test temperatures between 20 and 300 deg C do not influence the F, G and H values. Any significant effect of hydrogen (about 500 wtppm) on the anisotropy constants F, G and H has not been revealed at a test temperature of 300 deg C. The results, obtained for stress-relieved and recrystallized cladding with different texture, show an obvious influence of texture on the fracture toughness of Zircaloy cladding. A higher fracture toughness has been found for cladding with more radial texture

Hot hardness studies on zircaloy 2 pressure tube along three orientations

International Nuclear Information System (INIS)

Kutty, T.R.G.; Ravi, K.; Jarvis, T.; Sengupta, A.K.; Majumdar, S.; Tewari, R.; Shrivastava, D.; Dey, G.K.

2002-01-01

Zirconium based alloys are the natural choice for both the fuel element cans and in-core structural components in water cooled nuclear reactors. In this paper, the hot hardness behaviour of zircaloy 2 pressure tubes has been examined from room temperature to 400 degC using a hot hardness tester. For the purpose of comparison, the hardness of the as cast and room temperature rolled specimens has also been carried out. For this, the samples were cut along three orientations and hardness was measured in each of these directions using Vickers diamond pyramid indenter. The variation in hardness of the pressure tube samples show that the hardness was highest along circumferential direction and least along the axial direction. The room temperature rolled samples showed highest hardness along the rolling planes. These variations in hardness could be explained in terms of development of texture during working on the material. (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

Stress corrosion cracking behavior of zircaloy-2 in iodine environment

International Nuclear Information System (INIS)

Ikeda, Seiichi

1983-01-01

The effects of strain rates, iodine partial pressure and testing temperature on SCC behavior of zircaloy-2 in iodine environment were studied by means of slow strain rate technique (SSRT). SCC behavior of recrystallized specimens in iodine environment was remarkably influenced by the testing temperatures, and the susceptibility to SCC of specimens tested at 623 K was higher than that at 573 K. The susceptibility to SCC of recrystallized specimens increased with increasing iodine partial pressure at the lower strain rates of 4.2 x 10 -6 s -1 and 8.3 x 10 -7 s -1 . Cold worked specimens indicate no SCC failure in iodine environment regardless of strain rates, although those were tested only at 573 K. Fractographic observation revealed that SCC features of recrystallized specimens can be classified into two groups. One group, mostly specimens tested at 573 K, are characterized by the fact that cracks are initiated from corrosion pits. The other group are characterized by transgranuler SCC in the absence of pitting. This type of crack is found on specimens tested in environments containing more than 570 Pa iodine and seems to be produced by iodine embrittlement. (author)

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

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

A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Choi, Kwang Sik

1992-02-15

Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values.

A study on the delayed hydride cracking mechanism in cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and zircaloy-2 pressure tubes

International Nuclear Information System (INIS)

Choi, Kwang Sik

1992-02-01

Cold worked Zr-2.5Nb, heat treated Zr-2.5Nb and Zircaloy-2 pressure tubes were hydrided to the hydrogen concentration of 68 ppm, 49 ppm and 242-411 ppm, respectively, and compact tension specimens were machined from the hydrided materials. The crack growth rate by delayed hydride cracking was measured by potential drop method at various temperatures on the above mentioned three types of specimens. The activation energy obtained were 43 KJ/mol for cold worked Zr-2.5Nb and 37 KJ/mol for heat treated Zr-2.5Nb, which were in good agreements with that of Coleman (1977), while they were lower than the activation energy of 65.5 KJ/mol obtained by Simpson-puls (1979) and 71.5 KJ/mol by Ambler (1984). The DHC growth rate in Zircaloy-2 were about one fifth of that of Zr-2.5Nb, which is due to the texture and material strength effects. Striations which indicate stepwise DHC growth were observed at fracture surface by scanning electron microscope and unsymmetric crack tunnellings were also observed, which seems to be due to the difference in hydrogen diffusion rate caused by the difference in stress fields between inner and outer surface. The comparison of test results with the DHC growth rate calculated by Simpson-puls model showed good agreement at high temperatures, whereas at the lower temperatures the crack growth rates were 2.5 times higher than the calculated values

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

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)

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

16-8-2 weld metal design data for 316L(N) steel

Energy Technology Data Exchange (ETDEWEB)

Tavassoli, A.-A.F. [Commissariat a l' Energie Atomique, CEA/Saclay, 91191 Gif sur Yvette (France)], E-mail: tavassoli@cea.fr

2008-12-15

ITER materials properties documentation is extended to weld metals used for welding Type 316L(N) steel, i.e. the structural material retained for manufacturing ITER major components, such as the vacuum vessel. The data presented here are mainly for the Type 16-8-2 and complete those already reported for the low temperature (Type 316L) and the high temperature (Type 19-12-2) filler metals. The weld metal properties data for Type 16-8-2 filler metal and its joints are collected, sorted and analysed according to the French design and construction rules for nuclear components (RCC-MR). Particular attention is paid to the type of weld metal (e.g. wire for TIG, covered electrode for manual arc, flux wire for automatic welding), as well as, to the weld geometry and welding position. Design allowables are derived from validated data for each category of weld and compared with those of the base metal. In most cases, the analyses performed are extended beyond the conventional analyses required for codes to cover specific needs of ITER. These include effects of exposures to high temperature cycles during component fabrication, e.g. HIPing and low dose neutron irradiation at low and medium temperatures. The ITER Materials Properties Handbook (MPH) is, here, enriched with files for physical and mechanical properties of Type 16-8-2 weld metal. These files, combined with the codification and inspection files, are part of the documentation required for ITER licensing needs. They show that all three weld-metals satisfy the code requirements, provided compositions and types of welds used correspond to those specified in RCC-MR.

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)

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

Assessment of hydrogen levels in Zircaloy-2 by non-destructive testing

International Nuclear Information System (INIS)

De, P.K.; John, J.T.; Banerjee, S.; Jayakumar, T.; Thavasimuthu, M.; Raj, B.

1998-01-01

A non-destructive assessment of Zircaloy-2 samples charged with hydrogen in the range of 50 to 1150 mg/kg has been made using ultrasonic and eddy current testing. It has been found that the ratio of the longitudinal to the shear wave velocity is a parameter which can be directly correlated with the hydrogen content up to a level of 100 to 200 mg/kg. This parameter together with the values of longitudinal and shear wave velocities can be utilized in a multi-parametric correlation approach for estimation of higher levels of the hydrogen content (up to 1150 mg/kg). The sensitivity at different ranges has been found to be acceptable. Ultrasonic attenuation measurements at higher frequencies and eddy current test parameter are also effective for estimation of hydrogen levels above 250 mg/kg in zirconium alloys. Microstructural characterization including TEM studies have been carried out for studying the influence of the type and the morphology of hydride precipitates on ultrasonic parameters. (orig.)

Welding zinc coated steel with a CO/sub 2/ laser

International Nuclear Information System (INIS)

Akhter, R.; Steen, W.M.

1993-01-01

Welding of zinc coated steel has been studied using a high power CO/sub 2/ laser. This process is of great interest to the manufactures of car, washing machines and other components made from sheet steel and subject to corrosion. The problem associated with the welding of zinc coated steel is the low boiling point of zinc (906C) relative to the high melting point of steel (1500C). The problem is particularly important in lap welding where the zinc layer is between the lapped sheets. Under these conditions the laser 'keyhole' will generate very high vapour pressure in the zinc layer with a consequent severe risk of vapour eruption destroying the continuity of the weld bead. Several techniques are presented for the removal of zinc vapours from the interface between the two sheets. It is shown that this problem solved by suitable gap between the sheets during lap welding. Hence full penetration welds without deterioration of the weld bead can be obtained. A theory has been presented which predicted an exact gap size needed to exhaust the zinc vapour. The gap depends upon the welding speed, zinc coating thickness and thickness of the sheet. The theory predicts the weld quality satisfactorily. (author)

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...

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

Experimental investigations of the meltdown phase of UO2-Zircaloy fuel rods under conditions of failure of emergency cooling

International Nuclear Information System (INIS)

Hagen, S.; Mack, A.; Malauschek, H.; Wallenfels, K.

1975-01-01

In the monoxidizing helium atmosphere at 1,850 0 C Zircaloy and UO 2 interact violently. The result is a combined meltdown of pellets and can. This phenomenon appears independent of the velocity of temperature rise. In air the oxid skin splits open at 1,890 0 C and the earlier molten material of the interior begins to flow out. When heating up to more than 2,200 0 C the oxid skin remains solid nevertheless. (orig.) [de

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

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

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)

Requirements to gap widths and clamping for CO2 laser butt welding

DEFF Research Database (Denmark)

Gong, Hui; Juhl, Thomas Winther

1999-01-01

In the experimental study of fixturing and gap width requirements a clamping device for laser butt welding of steel sheets has been developed and tested. It has fulfilled the work and made the gap width experiments possible.It has shown that the maximum allowable gap width to some extent...... is inversely related to the welding speed. Also larger laser power leads to bigger allowable gap widths. The focal point position, though, has little influence on the maximum allowable gap width.During analysis X-ray photos show no interior porosity in the weld seam. Other methods have been applied to measure...... responses from variations in welding parameters.The table below lists the results of the study, showing the maximum allowable gap widths and some corresponding welding parameters.Maximum allowable Gap Width; Welding Speed; Laser Power:0.10 mm2 m/min2, 2.6 kW0.15 mm1 m/min2 kW0.20 mm1 m/min2.6 kW0.30 mm0.5 m...

Ultra high frequency induction welding of powder metal compacts

Directory of Open Access Journals (Sweden)

Çavdar, Uǧur

2014-06-01

Full Text Available The application of the iron based Powder Metal (PM compacts in Ultra High Frequency Induction Welding (UHFIW were reviewed. These PM compacts are used to produce cogs. This study investigates the methods of joining PM materials enforceability with UHFIW in the industry application. Maximum stress and maximum strain of welded PM compacts were determined by three point bending and strength tests. Microhardness and microstructure of induction welded compacts were determined.Soldadura por inducción de ultra alta frecuencia de polvos de metal compactados. Se ha realizado un estudio de la aplicación de polvos de metal (PM de base hierro compactados por soldadura por inducción de ultra alta frecuencia (UHFIW. Estos polvos de metal compactados se utilizan para producir engranajes. Este estudio investiga los métodos de uni.n de los materiales de PM con UHFIW en su aplicación en la industria. La máxima tensión y la máxima deformación de los polvos de metal compactados soldados fueron determinadas por flexión en tres puntos y prueba de resistencia. Se determinó la microdureza y la microestructura de los polvos compactados por soldadura por inducción.

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

HAZ microstructure in joints made of X13CrMoCoVNbNB9-2-1 (PB2 steel welded with and without post-weld heat treatment

Directory of Open Access Journals (Sweden)

M. Łomozik

2016-07-01

Full Text Available The article presents the results of research butt welded joints made of X13CrMoCoVNbNB9-2-1 steel. The joints were welded with post-weld heat treatment PWHT and without PWHT, using the temper bead technique TBT. After welding the joint welded with PWHT underwent stress-relief annealing at 770 °C for 3 hours. The scope of structural tests included the microstructural examination of the coarse-grained heat affected zone (HAZ areas of the joints, the comparison of the morphology of these areas and the determination of carbide precipitate types of the coarse grain heat affected zone (CGHAZ of the joints welded with and without PWHT.

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.)

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)

Mechanical properties of CO2/MIG welded structural rolled steel and stainless steel

International Nuclear Information System (INIS)

Lim, Jong Young; Yoon, Myong Jin; Kim, Sang Youn; Kim, Tae Gyu; Shin, Hyeon Seung

2015-01-01

To accomplish long-term use of specific parts of steel, welding technology is widely applied. In this study, to compare the efficiency in improving mechanical properties, rolled steel (SS400) was welded with stainless steel (STS304) by both CO 2 welding method and MIG (metal inert gas) welding method, respectively. Multi-tests were conducted on the welded specimen, such as X-ray irradiation, Vickers' Hardness, tensile test, fatigue test and fatigue crack growth test. Based on the fatigue crack growth test performed by two different methods, the relationship of da/dN was analyzed. Although the hardness by the two methods was similar, tensile test and fatigue properties of MIG welded specimen are superior to CO 2 welded one.

An improved Zircaloy-steam reaction model for use with the March 2 (Meltdown Accident Response Characteristics) code

International Nuclear Information System (INIS)

Manahan, M.P.

1983-01-01

An improved Zircaloy-steam oxidation reaction model has been incorporated into the MARCH 2 code which includes: (1) improved physical modeling for solid-state process oxidation, (2) improved geometric modeling for gaseous diffusion oxidation, (3) chemisorption/dissociation retardation due to high hydrogen partial pressures, and (4) laminar and turbulent flow conditions. Several accident sequences have been analyzed using the model, and for the sequences considered, the results indicate that the integrated and averaged variables are not significantly altered for the current level of fuel modeling, however, the localized variables such as nodal temperature and oxide thickness are affected

Ratcheting fatigue behavior of Zircaloy-2 at room temperature

Energy Technology Data Exchange (ETDEWEB)

Rajpurohit, R.S., E-mail: rsrajpurohit.rs.met13@iitbhu.ac.in [Department of Metallurgical Engineering, Indian Institute of Technology, Banaras Hindu University, Varanasi, 221005 (India); Sudhakar Rao, G. [Nuclear Energy and Safety Department, Paul Scherrer Institute, Villigen, CH-5232 (Switzerland); Chattopadhyay, K.; Santhi Srinivas, N.C.; Singh, Vakil [Department of Metallurgical Engineering, Indian Institute of Technology, Banaras Hindu University, Varanasi, 221005 (India)

2016-08-15

Nuclear core components of zirconium alloys experience asymmetric stress or strain cycling during service which leads to plastic strain accumulation and drastic reduction in fatigue life as well as dimensional instability of the component. Variables like loading rate, mean stress, and stress amplitude affect the influence of asymmetric loading. In the present investigation asymmetric stress controlled fatigue tests were conducted with mean stress from 80 to 150 MPa, stress amplitude from 270 to 340 MPa and stress rate from 30 to 750 MPa/s to study the process of plastic strain accumulation and its effect on fatigue life of Zircaloy-2 at room temperature. It was observed that with increase in mean stress and stress amplitude accumulation of ratcheting strain was increased and fatigue life was reduced. However, increase in stress rate led to improvement in fatigue life due to less accumulation of ratcheting strain. - Highlights: • Ratcheting strain accumulation occurred due to asymmetric cyclic loading. • Accumulation of ratcheting strain increased with mean stress and stress amplitude. • Ratcheting strain accumulation decreased with increase in stress rate. • With increase in mean stress and stress amplitude there was reduction in fatigue life. • Fatigue life is improved with increase in stress rate.

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)

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)

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)

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

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)

Interaction of both plasmas in CO2 laser-MAG hybrid welding of carbon steel

Science.gov (United States)

Kutsuna, Muneharu; Chen, Liang

2003-03-01

Researches and developments of laser and arc hybrid welding has been curried out since in 1978. Especially, CO2 laser and TIG hybrid welding has been studied for increasing the penetration depth and welding speed. Recently laser and MIG/MAG/Plasma hybrid welding processes have been developed and applied to industries. It was recognized as a new welding process that promote the flexibility of the process for increasing the penetration depth, welding speed and allowable joint gap and improving the quality of the welds. In the present work, CO2 Laser-MAG hybrid welding of carbon steel (SM490) was investigated to make clear the phenomenon and characteristics of hybrid welding process comparing with laser welding and MAG process. The effects of many process parameters such as welding current, arc voltage, welding speed, defocusing distance, laser-to-arc distance on penetration depth, bead shape, spatter, arc stability and plasma formation were investigated in the present work. Especially, the interaction of laser plasma and MAG arc plasma was considered by changing the laser to arc distance (=DLA).

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

Assisting Gas Optimization in CO2 Laser Welding

DEFF Research Database (Denmark)

Gong, Hui; Olsen, Flemming Ove

1996-01-01

High quality laser welding is achieved under the condition of optimizing all process parameters. Assisting gas plays an important role for sound welds. In the conventional welding process assisting gas is used as a shielding gas to prevent that the weld seam oxidates. In the laser welding process...... assisting gas is also needed to control the laser induced plasma.Assisting gas is one of the most important parameters in the laser welding process. It is responsible for obtaining a quality weld which is characterized by deep penetration, no interior imperfections, i.e. porosity, no crack, homogeneous seam...... surface, etc. In this work a specially designed flexible off-axis nozzle capable of adjusting the angle of the nozzle, the diameter of the nozzle, and the distance between the nozzle end and the welding zone is tested. In addition to the nozzle parameters three gases, Nitrogen, Argon, and Helium...

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

Fatigue strength of welds and welded materials of high-temperature steels resistant to pressurized hydrogen of the type 2.25% Cr/1% Mo

International Nuclear Information System (INIS)

Burlat, J.; Cheviet, A.; Million, A.

1986-01-01

The aim of the study is to examine systematically the creep strength of welded joints (base material, heat influence zone and welded seam) and of pure welding materials of the type 2 1/4-3% Cr/1% Mo. According to the AD standard rules, the rule which stipulates that the creep strength of welded seams under full stress be calculated with the strength characteristic value reduced by 20% applies to all heat-resistant steels, if no rupture stress values for the welded joints are available. Manufacturers of steel and weld fillers together with the Union of Technical Control Associations (VdTUeV) have prepared a test programme according to which on the one hand welded joints are tested at right angles to their seams, and on the other pure welding material is tested with respect to its creep strength. The development of the testes and their results have been described. The first results are available as VdTUeV material performance sheets, for 2 materials, and as provisional VdTUeV specification sheets, for 3 weld fillers. With the tested materials, it becomes practically feasible to reduce the creep strength of longitudinally welded pressure-bearing components by about 20% of wall thickness. (orig.) [de

The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

CO2 laser welding of galvanized steel sheets using vent holes

International Nuclear Information System (INIS)

Chen Weichiat; Ackerson, Paul; Molian, Pal

2009-01-01

Joining of galvanized steels is a challenging issue in the automotive industry because of the vaporization of zinc at 906 deg. C during fusion welding of steel (>1530 deg. C). In this work, hot-dip galvanized steel sheets of 0.68 mm thick (24-gage) were pre-drilled using a pulsed Nd:YAG laser to form vent holes along the weld line and then seam welded in the lap-joint configuration using a continuous wave CO 2 laser. The welds were evaluated through optical and scanning electron microscopy and tensile/hardness tests. The vent holes allowed zinc vapors to escape through the weld zone without causing expulsion of molten metal, thereby eliminating the defects such as porosity, spatter, and loss of penetration. In addition, riveting of welds occurred so long as the weld width was greater than the hole diameter that in turn provided much higher strength over the traditional 'joint gap' method

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)

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

Toughness of 2,25Cr-1Mo steel and weld metal

Science.gov (United States)

Acarer, Mustafa; Arici, Gökhan; Acar, Filiz Kumdali; Keskinkilic, Selcuk; Kabakci, Fikret

2017-09-01

2,25Cr-1Mo steel is extensively used at elevated temperature structural applications in fossil fire power plants for steam pipes, nozzle chambers and petrochemical industry for hydrocracking unit due to its excellent creep resistance and good redundant to oxidation. Also they should have acceptable weldability and toughness. The steels are supplied in quenched and tempered condition and their welded components are subjected to post-weld heat treatment (PWHT). Tempering process is carried out at 690-710°C to improve toughness properties. However they are sensitive to reheat cracking and temper embrittlement. To measure temper embrittlement of the steels and their weld metal, temper embrittlement factor and formula (J factor - Watanabe and X formula- Bruscato) are used. Step cooling heat treatment is also applied to determine temper embrittlement. In this study, toughness properties of Cr Mo (W) steels were reviewed. Also transition temperature curves of 2,25Cr-1Mo steel and its weld metal were constructed before and after step cool heat treatment as experimental study. While 2,25Cr-1Mo steel as base metal was supplied, all weld metal samples were produced in Gedik Welding Company. Hardness measurements and microstructure evaluation were also carried out.

Annealing studies of zircaloy-2 cladding at 580-8500C

International Nuclear Information System (INIS)

Hindle, E.D.

1978-05-01

For fuel element cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant accidents (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then much experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 0 C for returning Zircaloy cladding to the annealed condition, so that for any transient, a point can be specified where the material should have annealed. An equation has been derived to give this information. (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)

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

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.

The effect of stimulated fission products on the structure and the mechanical properties of zircaloy

International Nuclear Information System (INIS)

Holub, F.

1982-01-01

The objective of investigation was to study the long-term effects of individual simulated fission products on the mechanical properties and the structure of Zircaloy. Tensile Test specimens of Zircaloy were annealed with important simulated fission products at 350 0 C up to 10,000 hours and at higher temperatures (500, 700 0 C) up to 2,000 hours. The principal methods of investigation on annealed Zircaloy specimens were tension tests at room temperature and at 400 0 C, scanning electron microscopy and microprobe technique, X-ray diffraction, X-ray fluorescence, optical metallography. The action of fission products at normal temperatures of reactor operation will give rise to a small enhancement of strength and a small drop of ductility of the fuel cladding material only. At high fuel pin temperatures which may be realized under abnormal operation conditions, some of the fission products potentially will produce detrimental consequences on the integrity of fuel pins. The most effective fission products will be: lanthanum oxide, followed by the earth alkaline oxides and the other rare earth oxides, molybdenum, iodine and cadmium

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

Contribution to the study of corrosion of zirconium and zircaloy-2 in superheated steam at 400 deg C (105 kg /cm{sup 2}); Contribution a l'etude de la corrosion du zirconium et du zircaloy-2 dans la vapeur d'eau surchauffee a 400 deg C (105 kg /cm{sup 2})

Energy Technology Data Exchange (ETDEWEB)

Coriou, H; Gauduchau, J; Grall, L; Hure, J; Pelras, M [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

1959-07-01

The corrosion kinetics of zircaloy-2 in water and steam at temperatures between 300 deg. C and 400 deg. C are represented by a curve sharply divided into two stages separated by a so-called transition point. After a first period of decreasing corrosion rate there follows a second period with much faster kinetics in which the speed is constant. After carrying out a methodical study of the corrosion of 'zircaloy-2 in the form of sheets and tubes. We have demonstrated, at 400 deg. C in steam, a systematic anomaly which appears at the transition point. The curve presents three quite distinct points; after the first period a fast corrosion is observed, followed by a third period at a slower speed. This leads us to believe that there may be not a single point but a transition zone, separating two types of kinetic behaviour and corresponding to modifications in the properties of the oxide layer. After this readjustment period a new corrosion law is established, lasting a considerable time, the corrosion speed being slower than that indicated so far. A study of the morphology of the oxide films which develop under these conditions has demonstrated the special part played by mechanical, physical and metallurgical factors in the case of zirconium. Deep penetration of oxide can thus show up on the inner wall of hammer-hardened tubes. Simultaneously a very considerable hydride formation occurs in the metal. (author) [French] La cinetique de corrosion du zircaloy-2 dans l'eau et la vapeur a des temperatures comprises entre 300 et 400 deg. C est representee par une courbe a deux periodes separees par un point singulier appele point de transition. A une premiere periode a vitesse de corrosion decroissante, succede une deuxieme periode a cinetique beaucoup plus rapide dont la vitesse est constante. Apres une etude systematique de la corrosion du zircaloy-2 sous forme de toles et de tubes, nous avons mis en evidence a 400 deg. C, dans la vapeur, une anomalie systematique qui se

Stress corrosion cracking of zircaloy. The use of laboratory data to predict in-reactor behaviour

International Nuclear Information System (INIS)

Miller, A.K.; Ocken, H.

1981-01-01

Pellet-cladding interaction (PCI) can lead to failure of the Zircaloy tubing used as cladding in water-cooled reactors. Many investigations have shown that the mechanism directly responsible for such fuel rod failures is stress corrosion cracking (SCC) of Zircaloy tubing. Laboratory studies have yielded extensive data on the time-to-failure (tsub(f)) behaviour of Zircaloy tubing specimens as a function of such important variables as the applied hoop stress (σ sub(h)), the iodine concentration (I 2 ), the temperature (T) and the fluence (F). These data have been used to predict the response of Zircaloy tubing exposed in-reactor. A typical approach is to fit laboratory data to obtain an empirical equation for tsub(f) in terms of the variables identified above. The question can then be posed as to whether it is appropriate to use such an empirical expression for predicting in-reactor behaviour. This paper describes the approach which has been taken in modelling the SCC process. It first reviews the experimental observations upon which the model is based. A summary of the key features of the model is then presented. The model's capabilities, emphasizing those predictions that are independent of data used to evaluate empirical constants, are briefly discussed. Finally, it is shown how the model can be used to predict important differences between the response of tubing specimens exposed in the laboratory and the response of large quantities of tubing exposed in-reactor

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)

Welded joint properties of steel 2.25Cr1MoNiNb

International Nuclear Information System (INIS)

Gladis, R.; Ivanek, J.; Gottwald, M.

1981-01-01

Welded joints of steel 08Cr2.25Mo1NiNb for fast reactor steam generators made using manual arc welding with electrodes of identical compositions attain short-term mechanical properties and times to fracture when creep tested that match those of the base material. The reduction of the carbidic phase content in the steel and the welded joint metal did not adversely affect the tensile properties of the welded joint while increasing notch toughness of the heat-affected zone. Reduced carbon and niobium contents in the steel and the welded joint resulted in significant reduction in the proportion of carbidic eutectic particles in both the heat-affected zone and the weld metal. (Ha)

A study of the accelerated zircaloy-4 oxidation reaction with H2O/H2 mixture gas

International Nuclear Information System (INIS)

Kim, Y. S.; Cho, I. J.

2001-01-01

A study of the Zircaloy-4 reaction with H 2 O/H 2 mixture gas is carried out by using TGA (Thermo Gravimetric Apparatus) to estimate the hydrogen embrittlement which can possibly cause catastrophic nuclear fuel rod failure. Reaction rates are measured as a function of H 2 /H 2 O. In the experiments reaction temperature is set at 500 .deg. C and total pressure of the mixture gas is maintained at 1 atm. Experimental results reveal that hydriding and oxidation reaction are competing. In early stage, hydriding kinetics is faster than oxidation, however, oxidant in H 2 O forms oxide on the surface as steam environment is maintained, thus, this growing oxide begins to protect the zirconium base metal against hydrogen permeation. In this second stage, the total kinetic rate follows enhanced oxidation kinetics. In the final stage, it is observed that the oxide is broken down and massive hydriding takes place through the mechanical defects in the oxide, whose kinetics is similar to pure hydriding kinetics. These results are confirmed by SEM and EDX analysis along with hydrogen concentration measurements

Role of internal stresses in the transient of irradiation growth of zircaloy-2

International Nuclear Information System (INIS)

Tome, C.N.; Christodoulou, N.; Turner, P.A.; Miller, M.A.; Woo, C.H.; Root, J.; Holden, T.M.

1995-07-01

A 'self-consistent' polycrystalline model is used to simulate irradiation growth of Zircaloy-2 samples irradiated at about 330 K. The predictions of the model are compared with experimental measurements obtained from specimens irradiated in the Advanced Test Reactor (ATR) at Idaho Falls. Three types of material are studied here: annealed, cold worked in tension and cold worked by rolling. In general, the growth rate attains a steady-state value after it goes through a transient that depends on the initial state of the material. The transient growth behaviour is explained in terms of the evolution of intergranular residual stresses that are present in the sample, and in terms of the dislocation structure. From this study, information regarding irradiation creep and growth mechanisms occurring at the single crystal level is obtained. (author). 28 refs., 1 tab., 4 figs

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.

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)

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)

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)

Gap conductance in Zircaloy-clad LWR fuel rods

International Nuclear Information System (INIS)

Ainscough, J.B.

1982-04-01

This report describes the procedures currently used to calculate fuel-cladding gap conductance in light water reactor fuel rods containing pelleted UO 2 in Zircaloy cladding, under both steady-state and transient conditions. The relevant theory is discussed together with some of the approximations usually made in performance modelling codes. The state of the physical property data which are needed for heat transfer calculations is examined and some of the relevant in- and out-of-reactor experimental work on fuel rod conductance is reviewed

Welding of dissimilar metals by CO2 lasers

International Nuclear Information System (INIS)

Garciandia, F.; Zubiri, F.; Etayo, J.L.; Cervantes, R.; Iriberri, I.

1998-01-01

The work carried out in CETENASA (laser department) in order to weld dissimilar metals is summarized. The involved metallic pair is M-35 and F-143, a high speed steel and a spring steel, respectively. Looking at the chemical composition of the involved alloys that will appear later, it can be easily understood the difficulty to obtain welded parts with structures metallurgically acceptable because of the high cracking degree that these materials show, specially M-35. The principles of a study which is being developed in the authors laboratory and which shows some interesting CO 2 laser possibilities are presented. (Author) 2 refs

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

Microstructure and Porosity of Laser-welded Dissimilar Material Joints of HR-2 and J75

Science.gov (United States)

Shen, Xianfeng; Teng, Wenhua; Zhao, Shuming; He, Wenpei

Dissimilar laser welding of HR-2 and J75 has a wide range of applications in high-and low-temperature hydrogen storage. The porosity distributions of the welded joints were investigated at different line energies, penetration status, and welding positions (1G, 2G, and 3G). The effect of the welding position on the welding appearance was evident only at high line energies because of the essential effect of gravity of the larger and longer dwelling molten pool. The porosity of the welded joints between the solutionised and aged J75 and HR-2 at the 3G position and partial penetration was located at the weld centre line, while the porosity at the 2G position with full penetration was distributed at the weld edges, which is consistent with the distribution of floating slag. Full keyhole penetration resulted in minimum porosity, partial penetration resulted in moderate porosity, and periodic molten pool penetration resulted in maximum porosity.

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

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

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)

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.)

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.

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.)

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)

Estudios Experimentales 2 Parte: Estudios Cuasi-Experimentales

OpenAIRE

Manterola, Carlos; Otzen, Tamara

2015-01-01

Los estudios experimentales, se caracterizan por la valoración del efecto de una o más intervenciones, habitualmente de forma comparativa con otra intervención, o un placebo; y el carácter prospectivo, de la recolección de datos y seguimiento. Agrupados bajo esta denominación, existe una diversidad de diseños, entre los que se encuentran los estudios cuasi-experimentales (ECE), que se caracterizan especialmente por la ausencia de asignación aleatoria. El objetivo de este manuscrito, es report...

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.

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

Studies of the Effective Total and Resonance Absorption Cross Sections for Zircaloy 2 and Zirconium

Energy Technology Data Exchange (ETDEWEB)

Hellstrand, E; Lindahl, G; Lundgren, G

1961-06-15

Using pile oscillator technique, the total absorption cross section for zircaloy 2 plates has been determined in the neutron spectrum of the reactor R1. The plate thickness was varied in six steps from 0. 2 mm to 6. 4 mm. The thermal cross section for the alloy was calculated from cross section data and the known composition of the alloy. By subtracting this value from the measured cross sections and dividing by the factor {alpha}=2/{radical}({pi}) x r x {radical}(T/T{sub 0}) the effective resonance integrals were obtained. After subtraction of a constant amount for resonance contributions from hafnium, tin etc., effective resonance integrals for zirconium could be evaluated. An extrapolated value of 0.85 {+-} 0.15 b was obtained for the infinitely dilute integral (l/v part excluded). The ratio of the resonance integral at plate thicknesses 0.2 and 6.4 mm came out as 1.65 {+-} 0.25.

Welding of heterogeneous 12Kh2MFSR steels with the Mn-Cr-Si-Ni system

International Nuclear Information System (INIS)

Smirnov, A.N.; Belogolov, E.I.

1978-01-01

The process of welding pipes of the 12Kh2MFSR pearlitic steels and austenitic steels of the Mn-Cr-Si-Ni system was studied. The filler materials were selected, and the working capacity of welded joints was examined in ageing and cyclic heatings. The microhardness of steels was measured, and the ultimate strength of welded joints was determined. The following has been established: the composite joints of steels of the Mn-Cr-Si-Ni system and 12Kh2MFSR steel are advisable to be welded on a coating layer welded by the EhA395/9 electrodes on the surface of a pipe of the 12Kh2MFSR pearlitic steel; this guarantees the sufficient working capacity of welded joints

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

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.

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)

Modeling of Zircaloy cladding degradation under repository conditions

International Nuclear Information System (INIS)

Santanam, L.; Raghavan, S.; Chin, B.A.

1989-07-01

Two potential degradation mechanisms, creep and stress corrosion cracking, of Zircaloy cladding during repository storage of spent nuclear fuel have been investigated. The deformation and fracture map methodology has been used to predict maximum allowable initial storage temperatures to achieve a thousand year life without rupture as a function of spent-fuel history. A stress analysis of fuel rods has been performed. Stresses in the outer zirconium oxide layer and the inner Zircaloy tube have been predicted for typical internal pressurization, oxide layer thickness, volume expansion from formation of the oxide layer and thermal expansion coefficients of the cladding and oxide. Stress relaxation occurring in-reactor has also been taken into account. The calculations indicate that for the anticipated storage conditions investigated, the outer zirconium oxide layer is in a state of compression thus making it unlikely that stress corrosion cracking of the exterior surface will occur. 20 refs., 6 figs., 9 tabs

Factors affecting in-core dimensional stability of Zircaloy-2 calandria tubes

International Nuclear Information System (INIS)

Fidleris, V.; Causey, A.R.; Holt, R.A.

1985-01-01

In CANDU PHW reactors, the heavy water moderator is contained in a cylindrical vessel (calandria) which is penetrated by 380 horizontal fuel channel assemblies. The outer Zircaloy-2 tube of each assembly (the calandria tube) is rolled into the end shields to seal the calandria. The calandria tubes operate at ≅340 K with axial stresses that range from -10 to +40 MPa and experience fast neutron fluxes as large as 3 x 10 17 n m -2 s -1 , E > 1.0 MeV. In this environment tubes elongate and sag due to irradiation-induced creep and growth. Our understanding of these irradiation effects is based on creep, stress relaxation and irradiation growth experiments on calandria tube materials irradiated to neutron fluences of 7 x 10 25 n m -2 , E > 1.0 MeV. Both creep and growth strains decrease with the proportion of grains that have basal plane normals in the direction of testing. Cold work increases the creep rate but appears to introduce a negative component of growth in the working direction due to neutron induced stress relief that persists up to at least 7 x 10 25 n m -2 . Thermal stress relief restores the positive growth rate in the working direction. There is little effect of grain size in the range 10 TO 30 μm. This information can be used to select fabrication routes that will minimize dimensional changes of tubes during service

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)

Automatic welding of fuel elements

International Nuclear Information System (INIS)

Briola, J.

1958-01-01

The welding process depends on the type of fuel element, the can material and the number of cartridges to be welded: - inert-gas welding (used for G2 and the 1. set of EL3), - inert atmosphere arc welding (used for welding uranium and zirconium), - electronic welding (used for the 2. set of EL3 and the tank of Proserpine). (author) [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

thick oxide outer layer over a thin barrier layer. From thermodynamic analysis, it is postulated that a hydride barrier layer forms under PWR coolant conditions whereas an oxide barrier layer forms under BWR primary coolant conditions. Thus, the introduction of hydrogen into the solution lowers the corrosion potential of zirconium to the extent that the formation of ZrH2 is predicted to be spontaneous rather than the ZrO2. Mott-Schottky analysis shows that the passive film formed on zirconium is n-type, which is consistent with the PDM, corresponding to a preponderance of oxygen/hydrogen vacancies and/or zirconium interstitials in the barrier layer. The model parameter values were extracted from electrochemical impedance spectroscopic data for zirconium in high temperature, de-aerated and hydrogenated environments by optimization. The results indicate that the corrosion resistance of zirconium is dominated by the porosity and thickness of the outer layer for both cases. The impedance model based on the PDM provides a good account of the growth of the bi-layer passive films described above, and the extracted model parameter values might be used, for example, for predicting the accumulation of general corrosion damage to Zircaloy fuel sheath in BWR and PWR operating environments. Transients in current density and film thickness for passive film formation on zirconium in dearated and hydrogenated coolant conditions have confirmed that the rate law afforded by the Point Defect Model (PDM) adequately describes the growth and thinning of the passive film. The experimental results demonstrate that the kinetics of oxygen or hydrogen vacancy generation at the metal/film interface control the rate of film growth, when the potential is displaced in the positive direction, whereas the kinetics of dissolution of the barrier layer at the barrier layer/solution interface control the rate of passive film thinning when the potential is stepped in the negative direction. In addition, the

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)

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)

Quantification of Microtexture at Weld Nugget of Friction Stir-Welded Carbon Steel

Science.gov (United States)

Husain, Md M.; Sarkar, R.; Pal, T. K.; Ghosh, M.; Prabhu, N.

2017-05-01

Friction stir welding of C-Mn steel was carried out under 800-1400 rpm tool rotation. Tool traversing speed of 50 mm/min remained same for all joints. Effect of thermal state and deformation on texture and microstructure at weld nugget was investigated. Weld nugget consisted of ferrite + bainite/Widmanstatten ferrite with different matrix grain sizes depending on peak temperature. A texture around ( ϕ 2 = 0°, φ = 30°, ϕ 2 = 45°) was developed at weld nugget. Grain boundary misorientation at weld nugget indicated that continuous dynamic recrystallization influenced the development of fine equiaxed grain structure. Pole figures and orientation distribution function were used to determine crystallographic texture at weld nugget and base metal. Shear texture components D1, D2 and F were present at weld nugget. D1 shear texture was more prominent among all. Large number of high-angle grain boundaries ( 60-70%) was observed at weld nugget and was the resultant of accumulation of high amount of dislocation, followed by subgrain formation.

Generation rate of carbon monoxide from CO2 arc welding.

Science.gov (United States)

Ojima, Jun

2013-01-01

CO poisoning has been a serious industrial hazard in Japanese workplaces. Although incomplete combustion is the major cause of CO generation, there is a risk of CO poisoning during some welding operations. The aim of the present study was to evaluate the generation rate of CO from CO2 arc welding under controlled laboratory conditions and estimate the ventilation requirements for the prevention of CO poisoning. Bead on plate welding was carried out with an automatic welding robot on a rolled steel base metal under several conditions. The concentration of emitted CO from the welding was measured by a real-time CO monitor in a well-ventilated laboratory that was free from ambient CO contamination. The generation rate of CO was obtained from the three measurements-the flow rate of the welding exhaust gas, CO concentration in the exhaust gas and the arcing time. Then the ventilation requirement to prevent CO poisoning was calculated. The generation rate of CO was found to be 386-883 ml/min with a solid wire and 331-1,293 ml/min with a flux cored wire respectively. It was found that the CO concentration in a room would be maintained theoretically below the OSHA PEL (50 ppm) providing the ventilation rate in the room was 6.6-25.9 m3/min. The actual ventilation requirement was then estimated to be 6.6-259 m3/min considering incomplete mixing. In order to prevent CO poisoning, some countermeasures against gaseous emission as well as welding fumes should be taken eagerly.

Welding procedure for 06Kh13N7D2 steel

International Nuclear Information System (INIS)

Muromtsev, B.I.; Turkov, I.I.

1990-01-01

Based on the results of investigations into the process strength, mechanical and corrosion properties of 08Kh13N7D2 steel welded joints, the optimal method of its welding and a possibility of applying it for high-strength mounting in nuclear power plants are determined

In situ measurement of the effect of LiOH on the stability of zircaloy-2 surface film in PWR water

International Nuclear Information System (INIS)

Saario, T.; Taehtinen, S.

1997-01-01

Surface films on the metals play a major role in corrosion assisted cracking. A new method called Contact Electric Resistance (CER) method has been recently developed for in situ measurement of the electric resistance of surface films in high temperature and high pressure environments. The technique has been used to determine in situ the electric resistance of films on metals when in contact with water and dissolved anions, during formation and destruction of oxides and hydrides and during electroplating of metals. Electric resistance data can be measured with a frequency of the order of one hertz, which makes it possible to investigate in situ the kinetics of surface film related processes which are dependent on the environment, temperature, pH and electrochemical potential. This paper presents the results of the CER investigation on the effects of LiOH on the stability of Zircaloy-2 surface film in water with 2000 ppm H 3 BO 3 . At 300 deg. C the LiOH concentrations higher than 10 -2 M (roughly 70 ppm of Li + ) were found to markedly reduce the electric resistance of the Zircaloy-2 surface film during a test period of less than two hours. The decrease of the film resistance is very abrupt, possibly indicating a phase transformation. Moreover, the advantages of the CER technique over the other competing techniques which rely on the measurement of current are discussed. (author)

In situ measurement of the effect of LiOH on the stability of zircaloy-2 surface film in PWR water

Energy Technology Data Exchange (ETDEWEB)

Saario, T; Taehtinen, S [Technical Research Centre of Finland, Espoo (Finland)

1997-02-01

Surface films on the metals play a major role in corrosion assisted cracking. A new method called Contact Electric Resistance (CER) method has been recently developed for in situ measurement of the electric resistance of surface films in high temperature and high pressure environments. The technique has been used to determine in situ the electric resistance of films on metals when in contact with water and dissolved anions, during formation and destruction of oxides and hydrides and during electroplating of metals. Electric resistance data can be measured with a frequency of the order of one hertz, which makes it possible to investigate in situ the kinetics of surface film related processes which are dependent on the environment, temperature, pH and electrochemical potential. This paper presents the results of the CER investigation on the effects of LiOH on the stability of Zircaloy-2 surface film in water with 2000 ppm H{sub 3}BO{sub 3}. At 300 deg. C the LiOH concentrations higher than 10{sup -2} M (roughly 70 ppm of Li{sup +}) were found to markedly reduce the electric resistance of the Zircaloy-2 surface film during a test period of less than two hours. The decrease of the film resistance is very abrupt, possibly indicating a phase transformation. Moreover, the advantages of the CER technique over the other competing techniques which rely on the measurement of current are discussed. (author).

Experimental investigation on the weld pool formation process in plasma keyhole arc welding

Science.gov (United States)

Van Anh, Nguyen; Tashiro, Shinichi; Van Hanh, Bui; Tanaka, Manabu

2018-01-01

This paper seeks to clarify the weld pool formation process in plasma keyhole arc welding (PKAW). We adopted, for the first time, the measurement of the 3D convection inside the weld pool in PKAW by stereo synchronous imaging of tungsten tracer particles using two sets of x-ray transmission systems. The 2D convection on the weld pool surface was also measured using zirconia tracer particles. Through these measurements, the convection in a wide range of weld pools from the vicinity of the keyhole to the rear region was successfully visualized. In order to discuss the heat transport process in a weld pool, the 2D temperature distribution on the weld pool surface was also measured by two-color pyrometry. The results of the comprehensive experimental measurement indicate that the shear force due to plasma flow is found to be the dominant driving force in the weld pool formation process in PKAW. Thus, heat transport in a weld pool is considered to be governed by two large convective patterns near the keyhole: (1) eddy pairs on the surface (perpendicular to the torch axis), and (2) eddy pairs on the bulk of the weld pool (on the plane of the torch). They are formed with an equal velocity of approximately 0.35 m s-1 and are mainly driven by shear force. Furthermore, the flow velocity of the weld pool convection becomes considerably higher than that of other welding processes, such as TIG welding and GMA welding, due to larger plasma flow velocity.

Phase transformations in neutron-irradiated Zircaloys

International Nuclear Information System (INIS)

Chung, H.M.

1986-04-01

Microstructural evolution in Zircaloy-2 and -4 spent-fuel cladding specimens after ∼3 years of irradiation in commercial power reactors has been investigated by TEM and HVEM. Two kinds of precipitates induced by the fast-neutron irradiation in the reactors have been identified, i.e., Zr 3 O and cubic-ZrO 2 particles approximately 2 to 10 nm in size. By means of a weak-beam dark-field ''2-1/2D-microscopy'' technique, the bulk nature of the precipitates and the surficial nature of artifact oxide and hydride phases could be discerned. The Zr(Fe/sub x/,Cr/sub 1-x/) 2 and Zr 2 (Fe/sub x/,Ni/sub 1-x/) intermetallic precipitates normally present in the as-fabricated material virtually dissolved in the spent-fuel cladding specimens after a fast-neutron fluence of ∼4 x 10 21 ncm -2 in the power reactors. The observed radiation-induced phase transformations are compared with predictions based on the currently available understanding of the alloy characteristics. 29 refs

Simulation of Zircaloy cladding deformation under accident conditions derived from analysis of data from Three Mile Island-2

International Nuclear Information System (INIS)

Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

1981-01-01

A limited series of tests has been carried out based on a published analysis of Three Mile Island data. Zircaloy PWR cladding specimens were pressurised to 6.9 MPa at 500 deg C and heated at 0.2-1.0 deg C/sec in slowly flowing steam until they failed. The temperature at which rupture occurred ranged from 700 to 760 deg C. Three specimens were directly heated, and one was indirectly heated using an internal heater. The lengths of cladding strained greater than 33% ranged from 5.7 to 9.7 times the original diameter

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.

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

irradiation growth in annealed Zr2.5wt%Nb at 3530K

International Nuclear Information System (INIS)

Rogerson, A.; Murgatroyd, R.A.

1978-10-01

Zr 2.5wt%Nb growth specimens have been irradiated at 353 0 K to a fast neutron dose of approximately 4.0 x 10 25 n/m 2 . Specimens were taken from the longitudinal and transverse directions of a nominally annealed, seam-welded tube and irradiated in both the stress relieved and fully annealed conditions. Growth in these specimens is characterised by large positive and negative strains in the longitudinal and transverse directions respectively, with dimensional changes in weld material exhibiting intermediate growth behaviour. The results are compared with growth data on both annealed and cold worked Zircaloy-2 at 353 0 K and discussed in terms of the effect of texture, grain size, and cold work on irradiation growth. It is concluded that the continuation of growth to high doses in annealed Zr-2.5wt%Nb at 353 0 K results from interstitial induced dislocation climb with vacancies diffusing to grain boundaries. (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)

Soldadura (Welding). Spanish Translations for Welding.

Science.gov (United States)

Hohhertz, Durwin

Thirty transparency masters with Spanish subtitles for key words are provided for a welding/general mechanical repair course. The transparency masters are on such topics as oxyacetylene welding; oxyacetylene welding equipment; welding safety; different types of welds; braze welding; cutting torches; cutting with a torch; protective equipment; arc…

Effects of Mars Atmosphere on Arc Welds: Phase 2

Science.gov (United States)

Courtright, Z. S.

2018-01-01

Gas tungsten arc welding (GTAW) is a vital fusion welding process widely used throughout the aerospace industry. Its use may be critical for the repair or manufacture of systems, rockets, or facilities on the Martian surface. Aluminum alloy AA2219-T87 and titanium alloy Ti-6Al-4V butt welds have been investigated for weldability and weld properties in a simulated Martian gas environment. The resulting simulated Martian welds were compared to welds made in a terrestrial atmosphere, all of which used argon shielding gas. It was found that GTAW is a process that may be used in a Martian gas environment, not accounting for pressure and gravitational effects, as long as adequate argon shielding gas is used to protect the weld metal. Simulated Martian welds exhibited higher hardness in all cases and higher tensile strength in the case of AA2219-T87. This has been attributed to the absorption of carbon into the fusion zone, causing carbide precipitates to form. These precipitates may act to pin dislocations upon tensile testing of AA2219-T87. Dissolved carbon may have also led to carburization, which may have caused the increase in hardness within the fusion zone of the welds. Based on the results of this experiment and other similar experiments, GTAW appears to be a promising process for welding in a Martian gas environment. Additional funding and experimentation is necessary to determine the effects of the low pressure and low gravity environment found on Mars on GTAW.

Reaction- and melting behaviour of LWR-core components UO2, Zircaloy and steel during the meltdown period

International Nuclear Information System (INIS)

Hofmann, P.

1976-07-01

The reaction behaviour of the UO 2 , Zircaloy-4 and austenitic steel core components was investigated as a function of temperature (till melting temperatures) under inert and oxidizing conditions. Component concentrations varied between that of Corium-A (65 wt.% UO 2 , 18% Zry, 17% steel) and that of Corium-E (35 wt.% UO 2 , 10% Zry, 55% steel). In addition, Zircaloy and stainless steel were used with different degrees of oxidation. The paper describes systematically the phases that arise during heating and melting. The integral composition of the melts and the qualitative as well as quantitative analysis of the phases present in solidified corium are given. In some cases melting points have been determined. The reaction and melting behaviour of the corium specimens strongly depends on the concentration and on the degree of oxidation of the core components. First liquid phases are formed at the Zry-steel interface at about 1,350 0 C. The maximum temperatures of about 2,500 0 C for the complete melting of the corium-specimens are well below the UO 2 melting point. Depending on the steel content and/or degree of oxidation of Zry and steel, a homogeneous metallic or oxide melt or two immiscible melts - one oxide and the other metallic - are obtained. During the melting experiments performed under inert gas conditions the chemical composition of the molten specimens generally change by evaporation losses of single elements, especially of uranium, zirconium and oxygen. The total weight losses go up to 30%; under oxidizing conditions they are substantially smaller due to the occurrence of different phases. In air or water vapor, the occurrence of the phases and the melting behaviour of the core components are strongly influenced by the oxidation rate and the oxygen supply to the surface of the melt. In the case of the hypothetical core melting accident, a heterogeneous melt (oxide and metallic) is probable after the meltdown period. (orig./RW) [de

Measurements of delayed hydride cracking propagation rate in the radial direction of Zircaloy-2 cladding tubes

Energy Technology Data Exchange (ETDEWEB)

Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan); Uchikoshi, H. [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan)

2012-08-15

Highlights: Black-Right-Pointing-Pointer The delayed hydride cracking (DHC) velocity of Zircaloy-2 was measured. Black-Right-Pointing-Pointer The velocity followed the Arrhenius law up to 270 Degree-Sign C. Activation energy was 49 kJ/mol. Black-Right-Pointing-Pointer The threshold stress intensity factor for the DHC was from 4 to 6 MPa m{sup 1/2}. Black-Right-Pointing-Pointer An increase in material strength accelerated the DHC. Black-Right-Pointing-Pointer Precipitation and fracture of hydrides at a crack tip is responsible for the DHC. - Abstract: Delayed hydride cracking (DHC) tests of Zircaloy-2 cladding tubes were performed in the chamber of a scanning electron microscope (SEM) to directly observe the crack propagation and measure the crack velocity in the radial direction of the tubes. Pre-cracks were produced at the outer surfaces of the tubes. Hydrogen contents of the tubes were from 90 ppm to 130 ppm and test temperatures were from 225 Degree-Sign C to 300 Degree-Sign C. The crack velocity followed the Arrhenius law at temperatures lower than about 270 Degree-Sign C with apparent activation energy of about 49 kJ/mol. The upper temperature limit for DHC, above which DHC did not occur, was about 280 Degree-Sign C. The threshold stress intensity factor for the initiation of the crack propagation, K{sub IH}, was from about 4 MPa m{sup 1/2} to 6 MPa m{sup 1/2}, almost independent of temperature. An increase in 0.2% offset yield stress of the material accelerated the crack velocity and slightly decreased K{sub IH}. Detailed observations of crack tip movement showed that cracks propagated in an intermittent fashion and the propagation gradually approached the steady state as the crack depth increased. The SEM observations also showed that hydrides were formed at a crack tip and a number of micro-cracks were found in the hydrides. It was presumed from these observations that the repetition of precipitation and fracture of hydrides at the crack tip would be

Hybrid 2D-3D modelling of GTA welding with filler wire addition

KAUST Repository

Traidia, Abderrazak

2012-07-01

A hybrid 2D-3D model for the numerical simulation of Gas Tungsten Arc welding is proposed in this paper. It offers the possibility to predict the temperature field as well as the shape of the solidified weld joint for different operating parameters, with relatively good accuracy and reasonable computational cost. Also, an original approach to simulate the effect of immersing a cold filler wire in the weld pool is presented. The simulation results reveal two important observations. First, the weld pool depth is locally decreased in the presence of filler metal, which is due to the energy absorption by the cold feeding wire from the hot molten pool. In addition, the weld shape, maximum temperature and thermal cycles in the workpiece are relatively well predicted even when a 2D model for the arc plasma region is used. © 2012 Elsevier Ltd. All rights reserved.

Irradiation capsule design capable of continuously monitoring the creepdown of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Thoms, K.R.; Dodd, C.V.; van der Kaa, T.; Hobson, D.O.

1978-01-01

An irradiation capsule which permits continuous monitoring of the creepdown of Zircaloy tubing has been designed and given preliminary tests. This design effort is the major element of a cooperative research program between the United States Nuclear Regulatory Commission and the Netherlands Energy Research Foundation (ECN) and is a part of the NRC-sponsored Zircaloy creepdown program. The purpose of the Zircaloy creepdown program is to provide data on the deformation characteristics of Zircaloy tubes, typical of LWR fuel element cladding, under combined axial and tangential compressive stresses. These data will be used to verify and improve the material behavior codes that are used for the description of fuel pin behavior. The first capsule of this series contains a mockup test specimen which was machined with three different diameters, nominally 10.92-mm, 10.54-mm and 11.30-mm (.430-in., .415-in., and .445-in.). This test specimen can be moved axially thereby varying the lift-off and serving as a calibration device for the eddy-current deformation monitoring probes. Fabrication of this capsule has been completed and during out-or-reactor checkout we were able to obtain a resolution of better than 0.01-mm (0.0004-in.). The capsule is scheduled for installation in the HFR on February 8, 1978, for a 26 day irradiation test. The first pressurized capsule, and therefore the first one to monitor in-reactor cladding deformation, will be installed in the HFR on May 3, 1978

Annealing studies of Zircaloy-2 cladding at 580-850 deg C

International Nuclear Information System (INIS)

Hindle, E.D.

1983-01-01

For fuel rod cladding it is important to determine if prior metallurgical condition combined with irradiation damage can influence high temperature deformation, because studies of such deformation are required to produce data for the cladding ballooning models which are used in analysing loss-of-coolant (LOCA). If the behaviour of all cladding conditions during a LOCA can be represented by, say, the annealed condition, then a great deal of experimental work on a multiplicity of cladding conditions can be avoided. By examining the metallographic structure and hardness, the present study determines the time required in the range 580 to 850 deg C for returning Zircaloy cladding to the annealed condition, so that for any transient a point can be specified where the material should have annealed. An equation has been derived to give this information. (author)

Modelling of Zircaloy-steam-oxidation under severe fuel damage conditions

International Nuclear Information System (INIS)

Malang, S.; Neitzel, H.J.

1983-01-01

Small break loss-of-coolant accidents and special transients in an LWR, in combination with loss of required safety systems, may lead to an uncovered core for an extended period of time. As a consequence, the cladding temperature could rise up to the melting point due to the decay heat, resulting in severely damaged fuel rods. During heat-up the claddings oxidize due to oxygen uptake from the steam atmosphere in the core. The modeling and assessment of the Zircaloy-steam oxidation under such conditions is important, mainly for two reasons: The oxidation of the cladding influences the temperature transients due to the exothermic heat of reaction; the amount of liquified fuel depends on the oxide layer thickness and the oxygen content of the remaining Zircaloy metal when the melting point is reached. (author)

A comparison of Zircaloy oxide thicknesses on Millstone-3 and North Anna-1 PWR fuel cladding

International Nuclear Information System (INIS)

Polley, M.V.; Evans, H.E.

1993-08-01

High concentrations of lithium in the coolant may enhance the corrosion rate of Zircaloy fuel cladding. In the present work, oxide thicknesses on fuel cladding from the Millstone 3 PWR were compared with those from the North Anna 1 PWR. The intention was to identify whether the higher lithium levels (up to 3.5 ppM) in the Millstone 3 primary coolant during cycles 2 and 3 led to significantly greater oxidation rates than in North Anna 1 which operated generally with lithium levels lower than 2.2 ppM. The comparisons were made by comparing the measurements with code predictions of Zircaloy oxidation in order to factor out the effect of operational variables on the oxide thicknesses achieved. Overall, Millstone 3 oxide thicknesses were found to be approximately 14% greater than North Anna 1 values. However, approximately 29% lower oxide thicknesses were found on reload Millstone 3 rods exposed to one cycle of elevated lithium chemistry than on Millstone 3 initial fuel exposed to one cycle of normal lithium chemistry during cycle 1. Furthermore, oxide thicknesses on Millstone 3 rods exposed to two cycles of elevated lithium chemistry were approximately 36% lower than on Millstone 3 rods exposed to one cycle of normal lithium chemistry plus one cycle of elevated lithium chemistry. Therefore, it cannot be concluded that elevated lithium operation in Millstone 3 led to enhanced Zircaloy fuel clad corrosion

A Study to Increase Weld Penetration in P91 Steel During TIG Welding by using Activating Fluxes

Science.gov (United States)

Singh, Akhilesh Kumar; Kumar, Mayank; Dey, Vidyut; Naresh Rai, Ram

2017-08-01

Activated Flux TIG (ATIG) welding is a unique joining process, invented at Paton Institute of electric welding in 1960. ATIG welding process is also known as flux zoned TIG (FZTIG). In this process, a thin layer of activating flux is applied along the line on the surface of the material where the welding is to be carries out. The ATIG process aids to increase the weld penetration in thick materials. Activating fluxes used in the literature show the use of oxides like TiO2, SiO2, Cr2O3, ZnO, CaO, Fe2O3, and MnO2 during welding of steels. In the present study, ATIG was carried out on P-91 steel. Though, Tungsten Inert Gas welding gives excellent quality welds, but the penetration obtained in such welding is still demanding. P91 steel which is ferritic steel is used in high temperature applications. As this steel is, generally, used in thick sections, fabrication of such structures with TIG welding is limited, due to its low depth of penetration. To increase the depth of penetration in P91while welding with ATIG, the role of various oxides were investigated. Apart from the oxides mentioned above, in the present study the role of B2O3, V2O5 and MgO, during ATIG welding of P91 was investigated. It was seen that, compared to TIG welding, there was phenomenal increase in weld penetration during ATIG welding. Amongst all the oxides used in this study, maximum penetration was achieved in case of B2O3. The measurements of weld penetration, bead width and heat affected zone of the weldings were carried out using an image analysis technique.

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.

Application of Factorial Design for Gas Parameter Optimization in CO2 Laser Welding

DEFF Research Database (Denmark)

Gong, Hui; Dragsted, Birgitte; Olsen, Flemming Ove

1997-01-01

The effect of different gas process parameters involved in CO2 laser welding has been studied by applying two-set of three-level complete factorial designs. In this work 5 gas parameters, gas type, gas flow rate, gas blowing angle, gas nozzle diameter, gas blowing point-offset, are optimized...... to be a very useful tool for parameter optimi-zation in laser welding process. Keywords: CO2 laser welding, gas parameters, factorial design, Analysis of Variance........ The bead-on-plate welding specimens are evaluated by a number of quality char-acteristics, such as the penetration depth and the seam width. The significance of the gas pa-rameters and their interactions are based on the data found by the Analysis of Variance-ANOVA. This statistic methodology is proven...

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.

Determination of Oxygen in Zircaloy Surfaces by Means of Charged Particle Activation Analysis

Energy Technology Data Exchange (ETDEWEB)

Lorenzen, J; Brune, D

1973-01-15

Oxygen in zircaloy surfaces has been determined by means of charged particle activation analysis employing the following two reactions I. 16O (d, n) 17F ->(beta+decay) 17O Q = - 1.63 MeV; II. 16O (d, pgamma) 17O Q = + 1.05 MeV. The detection limits for oxygen in such surfaces has been investigated by measuring the promptly emitted 0.87 MeV gamma rays (reaction II) and also the 511 keV annihilation radiation which arises from beta-decay of 17F (reaction I). The correlation between the detection limit for oxygen in zircaloy, the particle energy and the surface thickness analyzed has been evaluated. At a deuteron energy of 3 MeV a detection limit of 0.7 x 10-7 g/cm2 was obtained from the measurement of the prompt gamma radiation arising from the second of these reactions. The analysis carried out by means of this technique is characterized by a high rapidity

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)

The effect of plastic strain on the evolution of crystallographic texture in Zircaloy-2

Science.gov (United States)

Ballinger, R. G.; Lucas, G. E.; Pelloux, R. M.

1984-09-01

The evolution of crystallographic texture during plastic deformation was investigated in Zircaloy-2 using X-ray and metallographic techniques. Inverse pole figures, the resolved fraction of basal poles, and the volume fraction of twinned material, were determined as a function of plastic strain for several strain paths and initial textures at 298 K and 623 K. Incremental transverse platic strain ratios ( R) were mesured as a function of plastic strain. Texture rotation occurs early in the deformation process, after as little as 1.5% plastic strain. For compressive plastic strains, the resolved fraction of basal poles increases in the direction parallel to the strain axis. For tensile plastic strains, the resolved fraction of basal poles decreases in the direction parallel to the strain axis. The rate of change of the resolved fraction of basal poles with plastic strain is a function of the initial resolved fraction of basal poles. The texture rotation can be explained by considering the operation of the principal tensile twinning systems, {101¯2}.

Globalization of Japanese steel industry. Part 2. Welding materials; Tekkogyo no kokusaika. 2. Yozai

Energy Technology Data Exchange (ETDEWEB)

Aida, I. [Kobe Steel, Ltd., Kobe (Japan)

1995-01-01

This paper mainly discusses the current status and problems of arc welding materials. The domestic production of welding materials has decreased. The recent trend of demand is characterized by the change of form make-up of welding materials. Various technologies for welding materials and their operation in Japan have developed with the progress of steel materials. The high quality and high-grade welding technologies, highly efficient production processes, laborsaving, and robotization have been promoted in various fields. In response to the rapid strong yen, quality and cost have to be further pursued, and amenity and cleanliness of welding have to be realized. The welding technologies have to be developed for large structures, such as ultra high-rise buildings, energy and chemical plants, ships, marine structures, etc. For the welding materials which are applied to robots and robot systems, obstruction factors for the operation have to be removed, which include the unsteady arc, re-arc badness, spattering, wear of chip, slag formation, etc. These measures promote the globalization of welding materials. 17 refs., 4 figs.

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)

Examination of Zircaloy-clad spent fuel after extended pool storage

International Nuclear Information System (INIS)

Bradley, E.R.; Bailey, W.J.; Johnson, A.B. Jr.; Lowry, L.M.

1981-09-01

This report presents the results from metallurgical examinations of Zircaloy-clad fuel rods from two bundles (0551 and 0074) of Shippingport PWR Core 1 blanket fuel after extended water storage. Both bundles were exposed to water in the reactor from late 1957 until discharge. The estimated average burnups were 346 GJ/kgU (4000 MWd/MTU) for bundle 0551 and 1550 GJ/kgU (18,000 MWd/MTU) for bundle 0074. Fuel rods from bundle 0551 were stored in deionized water for nearly 21 yr prior to examination in 1980, representing the world's oldest pool-stored Zircaloy-clad fuel. Bundle 0074 has been stored in deionized water since reactor discharge in 1964. Data from the current metallurgical examinations enable a direct assessment of extended pool storage effects because the metallurgical condition of similar fuel rods was investigated and documented soon after reactor discharge. Data from current and past examinations were compared, and no significant degradation of the Zircaloy cladding was indicated after almost 21 yr in water storage. The cladding dimensions and mechanical properties, fission gas release, hydrogen contents of the cladding, and external oxide film thicknesses that were measured during the current examinations were all within the range of measurements made on fuel bundles soon after reactor discharge. The appearance of the external surfaces and the microstructures of the fuel and cladding were also similar to those reported previously. In addition, no evidence of accelerated corrosion or hydride redistribution in the cladding was observed

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

Study of necking stability in tension test of zircaloy-2, on range from 170 0 C to 620 0 C

International Nuclear Information System (INIS)

Okuda, M.Y.

1975-01-01

The objective of this work is to study necking behavior of Zircaloy-2 in a tension test in which the temperature range varies from 170 0 C to 620 0 C by means of a model. This model provides strain rate variations in the beginning of necking and the parameters in the / necking stability. A new parameter Ψ is presented which permits necking / stability description in metals by means of a simple tension test. It is also proceeded a behavioral study of ε versus ε curve after necking formation. (author)

Reduction of Biomechanical and Welding Fume Exposures in Stud Welding.

Science.gov (United States)

Fethke, Nathan B; Peters, Thomas M; Leonard, Stephanie; Metwali, Mahmoud; Mudunkotuwa, Imali A

2016-04-01

The welding of shear stud connectors to structural steel in construction requires a prolonged stooped posture that exposes ironworkers to biomechanical and welding fume hazards. In this study, biomechanical and welding fume exposures during stud welding using conventional methods were compared to exposures associated with use of a prototype system that allowed participants to weld from an upright position. The effect of base material (i.e. bare structural beam versus galvanized decking) on welding fume concentration (particle number and mass), particle size distribution, and particle composition was also explored. Thirty participants completed a series of stud welding simulations in a local apprenticeship training facility. Use of the upright system was associated with substantial reductions in trunk inclination and the activity levels of several muscle groups. Inhalable mass concentrations of welding fume (averaged over ~18 min) when using conventional methods were high (18.2 mg m(-3) for bare beam; 65.7 mg m(-3) for through deck), with estimated mass concentrations of iron (7.8 mg m(-3) for bare beam; 15.8 mg m(-3) for through deck), zinc (0.2 mg m(-3) for bare beam; 15.8 mg m(-3) for through deck), and manganese (0.9 mg m(-3) for bare beam; 1.5 mg m(-3) for through deck) often exceeding the American Conference of Governmental Industrial Hygienists Threshold Limit Values (TLVs). Number and mass concentrations were substantially reduced when using the upright system, although the total inhalable mass concentration remained above the TLV when welding through decking. The average diameters of the welding fume particles for both bare beam (31±17 nm) through deck conditions (34±34 nm) and the chemical composition of the particles indicated the presence of metallic nanoparticles. Stud welding exposes ironworkers to potentially high levels of biomechanical loading (primarily to the low back) and welding fume. The upright system used in this study improved exposure

Critical heat flux on micro-structured zircaloy surfaces for flow boiling of water at low pressures

International Nuclear Information System (INIS)

Haas, C.; Miassoedov, A.; Schulenberg, T.; Wetzel, T.

2012-01-01

The influence of surface structure on critical heat flux for flow boiling of water was investigated for Zircaloy tubes in a vertical annular test section. The objectives were to find suitable surface modification processes for Zircaloy tubes and to test their critical heat flux performance in comparison to the smooth tube. Surface structures with micro-channels, porous layer, oxidized layer, and elevations in micro- and nano-scale were produced on a section of a Zircaloy cladding tube. These modified tubes were tested in an internally heated vertical annulus with a heated length of 326 mm and an inner and outer diameter of 9.5 and 18 mm. The experiments were performed with mass fluxes of 250 and 400 kg/(m 2 s), outlet pressures between 120 and 300 kPa, and constant inlet subcooling enthalpy of 167 kJ/kg. Only a small influence of modified surface structures on critical heat flux was observed for the pressure of 120 kPa in the present test section geometry. However, with increasing pressure the critical heat flux could increase up to 29% using the surface structured tubes with micro-channels, porous and oxidized layers. Capillary effects and increased nucleation site density are assumed to improve the critical heat flux performance. (authors)

Effect of plastic strain on the evolution of crystallographic texture in Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Ballinger, R.G. (Massachusetts Inst. of Tech., Cambridge (USA)); Lucas, G.E. (California Univ., Santa Barbara (USA)); Pelloux, R.M. (Massachusetts Inst. of Tech., Cambridge (USA). Dept. of Materials Science and Engineering)

1984-09-01

The evolution of crystallographic texture during plastic deformation was investigated in Zircaloy-2 using X-ray and metallographic techniques. Inverse pole figures, the resolved fraction of basal poles, and the volume fraction of twinned material, were determined as a function of plastic strain for several strain paths and initial textures at 298 K and 623 K. Incremental transverse platic strain ratios (R) were measured as a function of plastic strain. Texture rotation occurs early in the deformation process, after as little as 1.5% plastic strain. For compressive plastic strains, the resolved fraction of basal poles increases in the direction parallel to the strain axis. For tensile plastic strains, the resolved fraction of basal poles decreases in the direction parallel to the strain axis. The rate of change of the resolved fraction of basal poles with plastic strain is a function of the initial resolved fraction of basal poles. The texture rotation can be explained by considering the operating of the principal tensile twinning systems, (10anti 12), .

Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

International Nuclear Information System (INIS)

Davis, W. Jr.; West, G.A.; Stacy, R.G.

1979-01-01

Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO 2 or 96 to 97% ThO 2 --3 to 4% UO 2 . Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO 2 or ThO 2 --UO 2 sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO 2 from BWRs and of Zircaloy-4-clad UO 2 from PWRs. Median particle sizes of UO 2 from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 μm; particle sizes of ThO 2 --UO 2 , under these same conditions, ranged from 137 to 202 μm. Similarly, median particle sizes of UO 2 from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 μm. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution deduced from experimental data, realistic estimates can be made of fractions of dislodged fuel having dimensions less than specified values

Technical basis for storage of Zircaloy-clad spent fuel in inert gases

International Nuclear Information System (INIS)

Johnson, A.B. Jr.; Gilbert, E.R.

1983-09-01

The technical bases to establish safe conditions for dry storage of Zircaloy-clad fuel are summarized. Dry storage of fuel with zirconium alloy cladding has been licensed in Canada, the Federal Republic of Germany, and Switzerland. Dry storage demonstrations, hot cell tests, and modeling have been conducted using Zircaloy-clad fuel. The demonstrations have included irradiated boiling water reactor, pressurized heavy-water reactor, and pressurized water reactor fuel assemblies. Irradiated fuel has been emplaced in and retrieved from metal casks, dry wells, silos, and a vault. Dry storage tests and demonstrations have involved about 15,000 fuel rods, and about 5600 rods have been monitored during dry storage in inert gases with maximum cladding temperatures ranging from 50 to 570 0 C. Although some tests and demonstrations are still in progress, there is currently no evidence that any rods exposed to inert gases have failed (one PWR rod exposed to an air cover gas failed at about 270 0 C). Based on this favorable experience, it is concluded that there is sufficient information on fuel rod behavior, storage conditions, and potential cladding failure mechanisms to support licensing of dry storage in the US. This licensing position includes a requirement for inert cover gases and a maximum cladding temperature guideline of 380 0 C for Zircaloy-clad fuel. Using an inert cover gas assures that even if fuel with cladding defects were placed in dry storage, or if defects develop during storage, the defects would not propagate. Tests and demonstrations involving Zircaloy-clad rods and assemblies with maximum cladding temperatures above 400 0 C are in progress. When the results from these tests have been evaluated, the viability of higher temperature limits should be examined. Acceptable conditions for storage in air and dry storage of consolidated fuel are issues yet to be resolved

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

Creep properties and simulation of weld repaired low alloy heat resistant CrMo and Mo steels at 540 deg C. Sub-project 2 - Ex-serviced 2.25Cr1M0 weld metal and cross weld repairs

Energy Technology Data Exchange (ETDEWEB)

Rui Wu; Storesund, Jan; Borggreen, Kjeld; Feilitzen, Carl von

2007-12-15

Weld repair has been carried out in an ex-serviced 10 CrMo 9 10 pipe by using 10 CrMo 9 10, 13 CrMo 4 4 and 15 Mo 3 consumables. Application of current welding procedure and consumables results in an over matched weld repair. This is verified by both creep tests and the creep simulations at even lower stresses than tested. Creep specimens have been extracted from ex-serviced 10 CrMo 9 10 parent metal (PM) and weld metal (WM), from virgin 10 CrMo 9 10 WM, from virgin 13 CrMo 4 4 WM, and from virgin 15 Mo 3 WM. In addition, cross weld specimens including weld metal, heat affected zone (HAZ) and parent metal have been taken from the ex-serviced 10 CrMo 9 10 weld joint, and from three weld repairs. In total, there are nine test series. The sequence of creep lifetime at 540 deg C at given stresses is; virgin 10 CrMo 9 10 weld metal > virgin 15 Mo 3 weld metal approx virgin 13 CrMo 4 4 weld metal approx ex-serviced 10 CrMo 9 10 weld metal >> ex-serviced 10 CrMo 9 10 parent metal > ex-serviced 10 CrMo 9 10 cross weld approx 10 CrMo 9 10 cross weld repair approx 13 CrMo 4 4 cross weld repair approx and 15 Mo 3 cross weld repair. All the series show good creep ductility. The ex-serviced 10 CrMo 9 10 parent metal shows a creep lifetime about one order of magnitude shorter than that for both the virgin parent metal and the ex-serviced 10 CrMo 9 10 weld metal, independent of stresses. Differences in creep lifetime among the ex-serviced 10 CrMo 9 10 cross weld and other cross weld repairs are negligible, simply because rupture always occurred in the ex-serviced 10 CrMo 9 10 parent metal, approximately 10 mm from HAZ, for all the cross welds. Necking is frequently observed in the ex-serviced 10 CrMo 9 10 parent metal at the opposite side of the fracture. Creep damage to a large and a small extend is found adjacent to the fracture and at the necking area, respectively. Other parts of the weld joint like weld metal and HAZ are damage-free, independent of stress, weld metal and

Influence of Loading Direction and Weld Reinforcement on Fatigue Performance of TIG Weld Seam

Directory of Open Access Journals (Sweden)

HUI Li

2018-02-01

Full Text Available The influence of loading direction and weld reinforcement on fatigue performance of TC2 titanium alloy TIG weld seam was investigated via fatigue experiments and SEM fracture observation. The results show that the fatigue life of retaining weld reinforcement specimens is lower than that of removing one in the same weld direction. The fatigue life of oblique weld specimens is higher than that of straight one with the same weld reinforcement treatment. The initiation of removing weld reinforcement specimens' fatigue crack sources is in the hole defect, but the weld reinforcement specimen initiate at the weld toes. During the early stage of fatigue crack propagation, the cracks all grow inside the weld seam metal with obvious fatigue striation. And the fatigue cracks of oblique weld specimens pass through the weld seam into the base with a typical toughness fatigue striation during the last stage of fatigue crack propagation. The dimple of straight weld specimens is little and shallow in the final fracture zone. The oblique weld specimens broke in the base metal area, and the dimple is dense.

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)

Macrostructural and microstructural features of 1 000 MPa grade TRIP steel joint by CO2 laser welding

Institute of Scientific and Technical Information of China (English)

Wang Wenquan; Sun Daqian; Kang Chungyun

2008-01-01

Bead-on-plate CO2 laser welding of 1 000 MPa grade transformation induced plasticity (TRIP) steel was conducted under different welding powers, welding speeds and shield gases. The macrostructural and microstructural features of the welded joint were investigated. The increase of welding speed reduced the width of the weld bead and the porosities in the weld bead resulting from the different flow mode of melted metal in weld pool. The decrease of welding power or use of shield gas of helium also contributed to the reduction of porosity in the weld bead due to the alleviation of induced plasma formation, thus stabilizing the keyhole. The porosity formation intimately correlated with the evaporation of alloy element Mn in the base metal. The laser welded metal had same martensite microstructure as that of water-quenched base metal. The welding parameters which increased cooling rate all led to fine microstructures of the weld bead.

Deformation behavior of a 16-8-2 GTA weld as influenced by its solidification substructure

International Nuclear Information System (INIS)

Foulds, J.R.; Moteff, J.; Sikka, V.K.; McEnerney, J.W.

1983-01-01

Weldment sections from formed and welded type 316 stainless steel pipe are characterized with respect to some time-independent (tensile) and time-dependent (creep) mechanical properties at temperatures between 25 0 C and 649 0 C. The GTA weldment, welded with 16-8-2 filler metal, is sectioned from pipe in the formed + welded + solution annealed + straightened condition, as well as in the same condition with an additional re-solution treatment. Detailed room temperature microhardness measurements on these sections before and after reannealing enable a determination of the different recovery characteristics of weld and base metal. The observed stable weld metal solidification dislocation substructure in comparison with the base metal random dislocation structure, in fact, adequately explains weld/base metal elevated temperature mechanical behavior differences from this recovery characteristic standpoint. The weld metal substructure is the only parameter common to the variety of austenitic stainless steel welds exhibiting the consistent parent/weld metal deformation behavior differences described. As such, it must be considered the key to understanding weldment mechanical behavior

Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

International Nuclear Information System (INIS)

Stout, R.B.

2001-01-01

An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

Hydride precipitation crack propagation in zircaloy cladding during a decreasing temperature history

Energy Technology Data Exchange (ETDEWEB)

Stout, R.B. [California Univ., Livermore, CA (United States). Lawrence Livermore National Lab

2001-07-01

An assessment of safety, design, and cost tradeoff issues for short (ten to fifty years) and longer (fifty to hundreds of years) interim dry storage of spent nuclear fuel in Zircaloy rods shall address potential failures of the Zircaloy cladding caused by the precipitation response of zirconium hydride platelets. To perform such assessment analyses rigorously and conservatively will be necessarily complex and difficult. For Zircaloy cladding, a model for zirconium hydride induced crack propagation velocity was developed for a decreasing temperature field and for hydrogen, temperature, and stress dependent diffusive transport of hydrogen to a generic hydride platelet at a crack tip. The development of the quasi-steady model is based on extensions of existing models for hydride precipitation kinetics for an isolated hydride platelet at a crack tip. An instability analysis model of hydride-crack growth was developed using existing concepts in a kinematic equation for crack propagation at a constant thermodynamic crack potential subject to brittle fracture conditions. At the time an instability is initiated, the crack propagation is no longer limited by hydride growth rate kinetics, but is then limited by stress rates. The model for slow hydride-crack growth will be further evaluated using existing available data. (authors)

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.)

[Impact of introduction of O2 on the welding arc of gas pool coupled activating TIG].

Science.gov (United States)

Huang, Yong; Wang, Yan-Lei; Zhang, Zhi-Guo

2014-05-01

In the present paper, Boltzmann plot method was applied to analyze the temperature distributions of the are plasma when the gas pool coupled activating TIG welding was at different coupling degrees with the outer gas being O2. Based on this study of temperature distributions, the changing regularities of are voltage and are appearance were studied. The result shows that compared with traditional TIG welding, the introduction of O2 makes the welding arc constricted slightly, the temperature of the are center build up, and the are voltage increase. When argon being the inner gas, oxygen serving as the outer gas instead of argon makes the are constricted more obviously. When the coupling degree increases from 0 to 2, the temperature of the are center and the are voltage both increase slightly. In the gas pool coupled activating TIG welding the are is constricted not obviously, and the reason why the weld penetration is improved dramatically in the welding of stainless steel is not are constriction.

Inverse strain rate effect on cyclic stress response in annealed Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Sudhakar Rao, G.; Verma, Preeti [Center of Advanced Study, Department of Metallurgical Engineering, Indian Institute of Technology (Banaras Hindu University), Varanasi 221005 (India); Chakravartty, J.K. [Mechanical Metallurgy Group, Bhabha Atomic Research Center, Trombay 400 085, Mumbai (India); Nudurupati, Saibaba [Nuclear Fuel Complex, Hyderabad 500 062 (India); Mahobia, G.S.; Santhi Srinivas, N.C. [Center of Advanced Study, Department of Metallurgical Engineering, Indian Institute of Technology (Banaras Hindu University), Varanasi 221005 (India); Singh, Vakil, E-mail: vsingh.met@itbhu.ac.in [Center of Advanced Study, Department of Metallurgical Engineering, Indian Institute of Technology (Banaras Hindu University), Varanasi 221005 (India)

2015-02-15

Low cycle fatigue behavior of annealed Zircaloy-2 was investigated at 300 and 400 °C at different strain amplitudes and strain rates of 10{sup −2}, 10{sup −3}, and 10{sup −4} s{sup −1}. Cyclic stress response showed initial hardening with decreasing rate of hardening, followed by linear cyclic hardening and finally secondary hardening with increasing rate of hardening for low strain amplitudes at both the temperatures. The rate as well the degree of linear hardening and secondary hardening decreased with decrease in strain rate at 300 °C, however, there was inverse effect of strain rate on cyclic stress response at 400 °C and cyclic stress was increased with decrease in strain rate. The fatigue life decreased with decrease in strain rate at both the temperatures. The occurrence of linear cyclic hardening, inverse effect of strain rate on cyclic stress response and deterioration in fatigue life with decrease in strain rate may be attributed to dynamic strain aging phenomena resulting from enhanced interaction of dislocations with solutes. Fracture surfaces revealed distinct striations, secondary cracking, and oxidation with decrease in strain rate. Deformation substructure showed parallel dislocation lines and dislocation band structure at 300 °C. Persistent slip band wall structure and development of fine Corduroy structure was observed at 400 °C.

Inverse strain rate effect on cyclic stress response in annealed Zircaloy-2

International Nuclear Information System (INIS)

Sudhakar Rao, G.; Verma, Preeti; Chakravartty, J.K.; Nudurupati, Saibaba; Mahobia, G.S.; Santhi Srinivas, N.C.; Singh, Vakil

2015-01-01

Low cycle fatigue behavior of annealed Zircaloy-2 was investigated at 300 and 400 °C at different strain amplitudes and strain rates of 10 −2 , 10 −3 , and 10 −4 s −1 . Cyclic stress response showed initial hardening with decreasing rate of hardening, followed by linear cyclic hardening and finally secondary hardening with increasing rate of hardening for low strain amplitudes at both the temperatures. The rate as well the degree of linear hardening and secondary hardening decreased with decrease in strain rate at 300 °C, however, there was inverse effect of strain rate on cyclic stress response at 400 °C and cyclic stress was increased with decrease in strain rate. The fatigue life decreased with decrease in strain rate at both the temperatures. The occurrence of linear cyclic hardening, inverse effect of strain rate on cyclic stress response and deterioration in fatigue life with decrease in strain rate may be attributed to dynamic strain aging phenomena resulting from enhanced interaction of dislocations with solutes. Fracture surfaces revealed distinct striations, secondary cracking, and oxidation with decrease in strain rate. Deformation substructure showed parallel dislocation lines and dislocation band structure at 300 °C. Persistent slip band wall structure and development of fine Corduroy structure was observed at 400 °C

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

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)

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

Real weld geometry determining mechanical properties of high power laser welded medium plates

Science.gov (United States)

Liu, Sang; Mi, Gaoyang; Yan, Fei; Wang, Chunming; Li, Peigen

2018-06-01

Weld width is commonly used as one of main factors to assess joint performances in laser welding. However, it changes significantly through the thickness direction in conditions of medium or thick plates. In this study, high-power autogenous laser welding was conducted on 7 mm thickness 201 stainless steel to elucidate the factor of whole weld transverse shape critically affecting the mechanical properties with the aim of predicting the performance visually through the weld appearance. The results show that single variation of welding parameters could result in great changes of weld pool figures and subsequently weld transverse shapes. All the obtained welds are composed of austenite containing small amount of cellular dendritic δ-Ferrite. The 0.2% proof stresses of Nail- and Peanut-shaped joint reach 458 MPa and 454 MPa, 88.2% and 87.5% of the base material respectively, while that of Wedge-shaped joint only comes to 371 MPa, 71.5% of the base material. The deterioration effect is believed to be caused by the axial grain zone in the weld center. The fatigue strength of joint P is a bit lower than N, but much better than W. Significant deformation incompatibility through the whole thickness and microstructure resistance to crack initiation should be responsible for the poor performance of W-shaped joints.

Urinary β2 Microglobulin in Workers Exposed to Arc Welding Fumes

Directory of Open Access Journals (Sweden)

Khosro Sadeghniiat-Haghighi

2011-11-01

Full Text Available Welding is a process in which two or more metals are attached by the use of heat and, in some cases, pressure. Direct exposure and inhalation of welding fumes causes acute and chronic side effects in humans. Kidney damage is one of these important side effects. β2 microglobulin is an 11.8 kilodalton protein and levels increase in the case of some inflammatory and viral diseases, or kidney malfunction and autoimmune diseases. In this study measurements of β2 microglobulin were used as a criterion for assessing effects on the kidneys of workers exposed to welding fumes. The study population were electric arc welders in an industrial plant in Tehran, Iran. For control we selected workers who did not have any exposure to welding fumes. Both groups were selected on the basis of a questionnaire and the consideration of criteria for inclusion and exclusion. In the end 50 cases and 50 controls were chosen. A urine sample was collected from all participants and urinary pH was set to between 6-8 using NaOH (1M. Sample transportation to the laboratory complied with the related standards. The samples were assessed using the ORG 5BM kit. For quantitative assessment of β2 microglobulin we used the Enzyme-linked Immunosorbent Assay (ELISA method. The ages of the welders ranged from 21 to 48 years (mean=30.5±5.9 yrs and of controls from 23 to 56 years (mean=31.8±5.9 yrs. Mean employment duration was 7.86±5.01years (range 2 to 27 years for welders. Mean β2 microglobulin level was 0.10±0.096 μg/ml in welders and 0.11±0.06 in controls. This difference was not statistically significant (P=0.381. In conclusion we don't find that exposure to electric arc welding fumes cause a significant change in urinary β2 microglobulin compared to the control group.

Qualifying program on Non-Destructive Testing, Visual Inspection of the welding (level 2)

International Nuclear Information System (INIS)

Shafee, M. A.

2011-01-01

Nondestructive testing is a wide group of analysis technique used in science and industry to evaluate the properties of a material, component or system without causing damage. Common Non-Destructive Testing methods include ultrasonic, magnetic-particle, liquid penetrate, radiographic, visual inspection and eddy-current testing. AAEA put the new book of the Non-Destructive Testing publication series that focused on Q ualifying program on Non-Destructive Testing, visual inspection of welding-level 2 . This book was done in accordance with the Arab standard certification of Non-Destructive Testing (ARAB-NDT-CERT-002) which is agreeing with the ISO-9712 (2005) and IAEA- TEC-DOC-487. It includes twenty one chapters dealing with engineering materials used in industry, the mechanical behavior of metals, metal forming equipments, welding, metallurgy, testing of welds, introduction to Non-Destructive Testing, defects in metals, welding defects and discontinuities, introduction to visual inspection theory, properties and tools of visual testing, visual testing, quality control regulations, standards, codes and specifications, procedures of welding inspections, responsibility of welding test inspector, qualification of Non-Destructive Testing inspector and health safety during working.

Physics of zinc vaporization and plasma absorption during CO2 laser welding

International Nuclear Information System (INIS)

Dasgupta, A. K.; Mazumder, J.; Li, P.

2007-01-01

A number of mathematical models have been developed earlier for single-material laser welding processes considering one-, two-, and three-dimensional heat and mass transfers. However, modeling of laser welding of materials with multiple compositions has been a difficult problem. This paper addresses a specific case of this problem where CO 2 laser welding of zinc-coated steel, commonly used in automobile body manufacturing, is mathematically modeled. The physics of a low boiling point material, zinc, is combined with a single-material (steel) welding model, considering multiple physical phenomena such as keyhole formation, capillary and thermocapillary forces, recoil and vapor pressures, etc. The physics of laser beam-plasma interaction is modeled to understand the effect on the quality of laser processing. Also, an adaptive meshing scheme is incorporated in the model for improving the overall computational efficiency. The model, whose results are found to be in close agreement with the experimental observations, can be easily extended for studying zinc-coated steel welding using other high power, continuous wave lasers such as Nd:YAG and Yb:YAG

Using Taguchi method to optimize welding pool of dissimilar laser welded components

OpenAIRE

Anawa, E.; Olabi, Abdul-Ghani

2008-01-01

In the present work CO2 continuous laser welding process was successfully applied and optimized for joining a dissimilar AISI 316 stainless steel and AISI 1009 low carbon steel plates. Laser power, welding speed, and defocusing distance combinations were carefully selected with the objective of producing welded joint with complete penetration, minimum fusion zone size and acceptable welding profile. Fusion zone area and shape of dissimilar austenitic stainless steel with ferritic low carbon s...

MAG narrow gap welding - an economic way to minimize welding expenses

International Nuclear Information System (INIS)

Kast, W.; Scholz, E.; Weyland, F.

1982-01-01

The thicker structural components are, the more important it is to take measures to reduce the volume of the weld. The welding process requiring the smallest possible weld section is the so-called narrow gap process. In submerged arc narrow gap welding as well as in MAG narrow gap welding different variants are imaginable, some of them already in practical use. With regard to efficiency and weld quality an optimum variant of the MAG narrow gap welding process is described. It constitutes a two wire system in which two wire electrodes of 1.2 mm diameter are arranged one behind the other. In order to avoid lack of fusion, the wire guides are slightly pointed towards each groove face. Thus, by inclining the two arcs burning one behind the other in the direction of weld progress, it is achieved that two separately solidifying weld pools and two beads per layer are simultaneously formed. Welding parameters are selected in such a way that a heat input of 16-20 kJ/cm and a deposition rate of 11-16 kgs/h are obtained. In spite of this comparatively high deposition rate, good impact values are found both in the weld and HAZ (largely reduced coarse-grain zone) which is due to an optimum weld build-up. With the available welding equipment the process can be applied to structural members having a thickness of 40-400 mm. The width of gap is 13 mm (root section) with a bevel angle of 1 0 . As filler metal, basic flux-cored wires are used which, depending on the base metal to be welded and the required tensile properties, can be of the Mn-, MnMo-, MnCrMo-, MnNi-, or MnNiMo-alloyed types. (orig.)

Embedded cladding surface thermocouples on Zircaloy-sheathed heater rods

International Nuclear Information System (INIS)

Wilkins, S.C.

1977-06-01

Titanium-sheathed Type K thermocouples embedded in the cladding wall of zircaloy-sheathed heater rods are described. These thermocouples constitute part of a program intended to characterize the uncertainty of measurements made by surface-mounted cladding thermocouples on nuclear fuel rods. Fabrication and installation detail, and laboratory testing of sample thermocouple installations are included

Capabilities of infrared weld monitor

Energy Technology Data Exchange (ETDEWEB)

Sanders, P.G.; Keske, J.S.; Leong, K.H.; Kornecki, G.

1997-11-01

A non-obtrusive pre-aligned, solid-state device has been developed to monitor the primary infrared emissions during laser welding. The weld monitor output is a 100-1000 mV signal that depends on the beam power and weld characteristics. The DC level of this signal is related to weld penetration, while AC portions of the output can be correlated with surface irregularities and part misalignment or contamination. Changes in DC behavior are also noted for both full and deep penetration welds. Full penetration welds are signified by an abrupt reduction in the weld monitor output. Bead on plate welds were made on steel, aluminum, and magnesium with both a CW CO{sub 2} laser and a pulsed Nd:YAG laser to explore the relationships between the weld characteristics and the weld monitor output.

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

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

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

Analysis of the Corrosion Behavior of an A-TIG Welded SS 409 Weld Fusion Zone

Science.gov (United States)

Vidyarthy, R. S.; Dwivedi, D. K.

2017-11-01

AISI 409 (SS 409) ferritic stainless steel is generally used as the thick gauge section in freight train wagons, in ocean containers, and in sugar refinery equipment. Activating the flux tungsten inert gas (A-TIG) welding process can reduce the welding cost during fabrication of thick sections. However, corrosion behavior of the A-TIG weld fusion zone is a prime concern for this type of steel. In the present work, the effect of the A-TIG welding process parameters on the corrosion behavior of a weld fusion zone made of 8-mm-thick AISI 409 ferritic stainless-steel plate has been analyzed. Potentiodynamic polarization tests were performed to evaluate the corrosion behavior. The maximum corrosion potential ( E corr) was shown by the weld made using a welding current of 215 A, a welding speed of 95 mm/min, and a flux coating density of 0.81 mg/cm2. The minimum E corr was observed in the weld made using a welding current of 190 A, a welding speed of 120 mm/min, and a flux coating density of 1.40 mg/cm2. The current study also presents the inclusive microstructure-corrosion property relationships using the collective techniques of scanning electron microscopy, energy-dispersive x-ray spectroscopy, and x-ray diffraction.

On the Generalized Correlation Equation of Welding Current for the Tig Welding Machine Used in Nuclear Fuel Fabrication

International Nuclear Information System (INIS)

Umar, Efrizon

1995-01-01

In nuclear fuel fabrication, welding plays a very important role to join the end cap to the tube. In order to determine the welding current in TIG welding process for various materials, weld geometries and welding rates, the correlation between the welding current and the other parameters are needed. This paper presents the correlation of those parameters mentioned above. The proposed correlation was tested and produced satisfactory results. (author). 8 refs., 2 tabs., 2 figs

Welding wires for high-tensile steels

International Nuclear Information System (INIS)

Laz'ko, V.E.; Starova, L.L.; Koval'chuk, V.G.; Maksimovich, T.L.; Labzina, I.E.; Yadrov, V.M.

1993-01-01

Strength of welded joints in arc welding of high-tensile steels of mean and high thickness by welding wires is equal to approximately 1300 MPa in thermohardened state and approximately 600 MPa without heat treatment. Sv-15Kh2NMTsRA-VI (EhK44-VI) -Sv-30Kh2NMTsRA-VI (EkK47-VI) welding wires are suggested for welding of medium-carbon alloyed steels. These wires provide monotonous growth of ultimate strength of weld metal in 1250-1900 MPa range with increase of C content in heat-treated state

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 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)

Sub-arc narrow gap welding of Atucha 2 RPV closure head

International Nuclear Information System (INIS)

Hantsch, H.; Million, K.; Zimmermann, H.

1982-01-01

Narrow gap technology was used for reasons of design and fabrication when welding the closure-head dome to its flange. Preliminary tests had yielded the necessary improvements of the well-proven sub-arc practice. New facilities had to be developed for welding proper and for the accompanying machining work (finishing in the narrow gap). Special measures were adopted for monitoring the welding process and for recording the welding parameters. The new method was tried out on several large test coupons before welding of the final product was started. No difficulties were encountered during the welding job. Fabrication of the closure head is shown in a short film sequence. (orig.)

Zircaloy cladding corrosion degradation in a Tuff repository: initial experimental plan

International Nuclear Information System (INIS)

Smith, H.D.

1984-07-01

The projected environmental history of a Tuff repository sited in an unsaturated hydrologic setting is evaluated to identify the potentially most severe corrosion conditions for Zircaloy spent fuel cladding. Three distinct corrosion periods are identified over the projected history. In two of those, liquid water may be present which is believed to produce the most severe corrosive environment for Zircaloy spent fuel cladding. In the time interval 100 to 1000 years after emplacement in the repository, the most severe condition is exposure to 170 0 C water at about 100 psi in an unbreached canister. This condition will be reproduced experimentally in an autoclave. For times after 1000 years, the most severe condition is exposure to 90 0 C water that is equilibrated with the tuff and invades breached canisters. This condition will be reproduced with a water bath system

Prediction of Weld Residual Stress of Narrow Gap Welds

International Nuclear Information System (INIS)

Yang, Jun Seog; Huh, Nam Su

2010-01-01

The conventional welding technique such as shield metal arc welding has been mostly applied to the piping system of the nuclear power plants. It is well known that this welding technique causes the overheating and welding defects due to the large groove angle of weld. On the other hand, the narrow gap welding(NGW) technique has many merits, for instance, the reduction of welding time, the shrinkage of weld and the small deformation of the weld due to the small groove angle and welding bead width comparing with the conventional welds. These characteristics of NGW affect the deformation behavior and the distribution of welding residual stress of NGW, thus it is believed that the residual stress results obtained from conventional welding procedure may not be applied to structural integrity evaluation of NGW. In this paper, the welding residual stress of NGW was predicted using the nonlinear finite element analysis to simulate the thermal and mechanical effects of the NGW. The present results can be used as the important information to perform the flaw evaluation and to improve the weld procedure of NGW

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

Performance of mesh seam welds in tailor welded blanks; Terado blank yo mash seam yosetsubu no tokusei

Energy Technology Data Exchange (ETDEWEB)

Uchihara, M; Takahashi, M; Kurita, M; Hirose, Y; Fukui, K [Sumitomo Metal Industries, Ltd., Osaka (Japan)

1997-10-01

Formability, fatigue properties and corrosion behavior of mash seam welded steel sheets were investigated and the results were compared with laser weld. The stretch formability of mash seam weld and laser weld were same level. Mash seam weld however, showed slightly smaller formability in hole expansion test. The fatigue strength of mash seam welds was lower than that of laser welds in case of differential thickness joints. Corrosion was apt to initiate at weld in both mash seam and laser weld with E-coat. The corrosion resistance of welds was improved by using zinc coated steel. 3 refs., 14 figs., 2 tabs.

Residual stresses and their mechanisms of production at circumferential weld by heat-sink welding

International Nuclear Information System (INIS)

Ueda, Yukio; Nakacho, Keiji; Ohkubo, Katsumi; Shimizu, Tsubasa.

1983-01-01

In the previous report, the authors showed effectiveness of the heat-sink welding (water cooling) to accomplish this end by conducting theoretical analysis and an experiment on residual stresses in the 4B pipe of SUS 304 by the conventional welding and the heat-sink welding at a certain standard heat-input condition. In this research, different pipe sizes and varied heat-input are applied. The welding residual stresses by the conventional welding and the heat-sink welding are obtained by the theoretical analysis and their production mechanisms are clarified. Hence the influence of the above changes of conditions on effectiveness of the heat-sink welding is investigated. The main results are summarized as follow. (1) In case of this pipes such as 2B and 4B pipes, it is important to minimize heat-input per one pass (especially for latter half passes) in order to improve the effectiveness of the heat-sink welding. The effectiveness can be predicted either by theoretical analysis of the temperature distribution history with consideration of the characteristic of heat transfer under spray-watering or by experimental measurement. (2) In case of 24B pipes, thick pipes, it is desirable to minimize heat-input for the first half passes, by which the heat-sink welding becomes more effective. In addition, no matter whether the conventional welding or the heat-sink welding, it is important to prevent angular distorsion which produces tensile axial stresses on the inner surface of the pipe in the weld zone. Possible measures to meet these requirements are to apply restraining jigs, to minimize the section area of the groove (ex. application of the narrow gap arc welding), and to change continuous welding to skip one. (J.P.N.)

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.)

Numerical weld modeling - a method for calculating weld-induced residual stresses

International Nuclear Information System (INIS)

Fricke, S.; Keim, E.; Schmidt, J.

2001-01-01

In the past, weld-induced residual stresses caused damage to numerous (power) plant parts, components and systems (Erve, M., Wesseling, U., Kilian, R., Hardt, R., Bruemmer, G., Maier, V., Ilg, U., 1994. Cracking in Stabilized Austenitic Stainless Steel Piping of German Boiling Water Reactors - Characteristic Features and Root Causes. 20. MPA-Seminar 1994, vol. 2, paper 29, pp.29.1-29.21). In the case of BWR nuclear power plants, this damage can be caused by the mechanism of intergranular stress corrosion cracking in austenitic piping or the core shroud in the reactor pressure vessel and is triggered chiefly by weld-induced residual stresses. One solution of this problem that has been used in the past involves experimental measurements of residual stresses in conjunction with weld optimization testing. However, the experimental analysis of all relevant parameters is an extremely tedious process. Numerical simulation using the finite element method (FEM) not only supplements this method but, in view of modern computer capacities, is also an equally valid alternative in its own right. This paper will demonstrate that the technique developed for numerical simulation of the welding process has not only been properly verified and validated on austenitic pipe welds, but that it also permits making selective statements on improvements to the welding process. For instance, numerical simulation can provide information on the starting point of welding for every weld bead, the effect of interpass cooling as far as a possible sensitization of the heat affected zone (HAZ) is concerned, the effect of gap width on the resultant weld residual stresses, or the effect of the 'last pass heat sink welding' (welding of the final passes while simultaneously cooling the inner surface with water) producing compressive stresses in the root area of a circumferential weld in an austenitic pipe. The computer program FERESA (finite element residual stress analysis) was based on a commercially

Resistance Spot Welding of dissimilar Steels

Directory of Open Access Journals (Sweden)

Ladislav Kolařík

2012-01-01

Full Text Available This paper presents an analysis of the properties of resistance spot welds between low carbon steel and austenitic CrNi stainless steel. The thickness of the welded dissimilar materials was 2 mm. A DeltaSpot welding gun with a process tape was used for welding the dissimilar steels. Resistance spot welds were produced with various welding parameters (welding currents ranging from 7 to 8 kA. Light microscopy, microhardness measurements across the welded joints, and EDX analysis were used to evaluate the quality of the resistance spot welds. The results confirm the applicability of DeltaSpot welding for this combination of materials.

Mechanical properties of dissimilar friction welded steel bars in relation to post weld heat treatment

Energy Technology Data Exchange (ETDEWEB)

Kong, Yu Sik; Kim, Seon Jin [Pukyong National University, Busan (Korea, Republic of)

2006-04-15

Dissimilar friction welding were produced using 15(mm) diameter solid bar in chrome molybedenum steel(KS SCM440) to carbon steel(KS S45C) to investigate their mechanical properties. The main friction welding parameters were selected to endure good quality welds on the basis of visual examination, tensile tests, Vickers hardness surveys of the bond of area and H.A.Z and microstructure investigations. The specimens were tested as-welded and Post-Weld Heat Treated(PWHT). The tensile strength of the friction welded steel bars was increased up to 100% of the S45C base metal under the condition of all heating time. Optimal welding conditions were n=2,000(rpm), P{sub 1}=60(MPa), P{sub 2}=100(MPa), t{sub 1}=4(s), t{sub 2}=5(s) when the total upset length is 5.4 and 5.7(mm), respectively. The peak of hardness distribution of the friction welded joints can be eliminated by PWHT. Two different kinds of materials are strongly mixed to show a well-combined structure of macro-particles without any molten material and particle growth or any defects.

Thermal gradient effects on the oxidation of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Klein, A.C.; Reyes, J.N. Jr.; Maguire, M.A.

1990-01-01

A Thermal Gradient Test Facility (TGTF) has been designed and constructed to measure the thermal gradient effect on pressurized water reactor (PWR) fuel rod cladding. The TGTF includes a heat flux simulator assembly capable of producing a wide range of PWR operating conditions including water flow velocities and temperatures, water chemistry conditions, cladding temperatures, and heat fluxes ranging to 160 W/cm 2 . It is fully instrumented including a large number of thermocouples both inside the water flow channel and inside the cladding. Two test programs are in progress. First, cladding specimens are pre-oxidized in air at 500 deg. C and in 400 deg. C steam for various lengths of time to develop a range of uniform oxide thicknesses from 1 to 60 micrometers. The pre-oxidized specimens are placed in the TGTF to characterize the oxide thermal conductivity under a variety of water flow and heat flux conditions. Second, to overcome the long exposure times required under typical PWR conditions a series of tests with the addition of high concentrations of lithium hydroxide to the water are being considered. Static autoclave tests have been conducted with lithium hydroxide concentrations ranging from 0 to 2 moles per liter at 300, 330, and 360 deg. C for up to 36 hours. Results for zircaloy-4 show a considerable increase in the weight gain for the exposed samples with oxidation rate enhancement factors as high as 70 times that of pure water. Operation of the TGTF with elevated lithium hydroxide levels will yield real-time information concerning the effects of a heat flux on the oxidation kinetics of zircaloy fuel rod cladding. (author). 5 refs, 5 figs, 2 tabs

Effect of dynamic strain aging on cyclic stress response and deformation behavior of Zircaloy-2

International Nuclear Information System (INIS)

Sudhakar Rao, G.; Verma, Preeti; Mahobia, G.S.; Santhi Srinivasa, N.C.; Singh, Vakil; Chakravartty, J.K.; Nudurupatic, Saibaba

2016-01-01

The effect of strain rate and temperature was studied on cyclic stress response and deformation behavior of annealed Zircaloy-2. Dynamic strain aging was exhibited under some test conditions. The cyclic stress response was found to be dependent on temperature and strain rate. At 300 °C, with decrease in strain rate, there was decrease in the rate as well as the degree of cyclic hardening. However, at 400°C, there was opposite trend and with decrease in strain rate both the rate as well as the degree of hardening increased. The deformation substructure showed dislocation bands, dislocation vein structure, PSB wall structure at both the temperatures. Irrespective of the temperature, there was dislocation loop structure, known as corduroy structure, at both the test temperatures. Based on the dislocation structure, the initial linear hardening is attributed to development of veins and PSB wall structure and the secondary hardening to the Corduroy structure. (author)

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

Laser welding to expand the allowable gap in bore welding for ITER blanket hydraulic connection

Energy Technology Data Exchange (ETDEWEB)

Tanigawa, Hisashi, E-mail: tanigawa.hisashi@jaea.go.jp; Maruyama, Takahito; Noguchi, Yuto; Takeda, Nobukazu; Kakudate, Satoshi

2015-10-15

For application to bore welding of hydraulic connection in the ITER blanket module, laser welding presents the following benefits: low weld heat input is preferred for re-welding of the irradiated material. Its contactless process can intrinsically avoid a failure mode of the tool sticking on the weld. The exact requirements for pipe alignment were assessed in comparison with the assembly tolerance. The groove geometry was modified to expand the allowable initial gap. The groove was machined to be partially thick to obviate the filler wire. First, plates with partially thick grooves were welded to elucidate the preferred groove geometry and welding conditions. With the modified groove, the plates were welded for the initial gap of 1.0 mm. Then the groove geometry and welding conditions were adjusted based on results of pipe welding tests. By application of the additional 0.5-mm-thick and 2.5-mm-wide metal in the groove, pipes with an initial gap of 0.7 mm were welded successfully.

Investigation of a weld defect, reactor vessel head Ringhals 2

International Nuclear Information System (INIS)

Embring, G.; Pers-Anderson, E.B.

1994-01-01

During the summer-outage 1993 Ringhals unit 2 vessel head was inspected at weld-area of Alloy 182. One major defect was found Two plus two ''boat-samples'' were taken out from the zone between the weld and the stainless cladding. All samples were sent to Studsviks laboratories for detailed investigations. The metallographic and fractographic investigations revealed that the major weld-defect had been there from manufacturing. The defect was located between the Alloy 182-buttering and the pressure vessel steel SA 533 grB cl 1. No indications of PWSCC or IDSCC were found. An inspection programme was defined. Different types of reference blocks were provided by Ringhals in cooperation with ABB TRC. Reference reflectors of type flat bottom hole (FBH) and eroded notches (EDM), with different sizes and separation were manufactured. One weld sample with manufacturing defects -lack of fusion and slag was inclusions- was present. ABB TRC performed UT inspection in the gap between the penetration and the thermal sleeve. Inspection results like defect identification, defect separation and defect sizing accuracy were compared with result from the destructive inspection. No relevant additional defects were found. An analysing and repair program was performed. A special designed disc sealed off the defect area. (authors). 5 figs., 3 refs

Comparison of welding induced residual stresses austenitic and ferritic steel weld joints

International Nuclear Information System (INIS)

Rajkumar, K.V.; Arun Kumar, S.; Mahadevan, S.; Manojkumar, R.; Rao, B. Purna Chandra; Albert, Shaju K.; Murugan, S.

2015-01-01

X-ray diffraction (XRD) is a well established technique for measurement of residual stresses in components and is being widely used. In XRD technique, the distance between the crystallographic planes (d spacing) is measured from peak position (2è) at various ø angles, where ø is the angle between the normal to the sample and the bisector of the incident and diffracted beam. From the slope of sin2ø vs. d spacing plot, the residual stresses are arrived by assuming a plane stress model. Welding induced residual stresses is of high importance as it is a major cause of failure in components. Surface compressive stresses improve the fatigue strength, whereas tensile residual stresses tend to decrease the fatigue strength. The present study compares the residual stresses that develop in 3 mm thick SS 316 and P91 TIG weld joints using the XRD technique. This study is aimed at understanding the influence of shrinkage during cooling and the effect of phase transformation induced volume changes on residual stress development in these two steels. While the first effect is predominant in the SS 316 weld, both the effects are present in the P91 welds. Stress measurements on SS 316 and P91 were carried out using Cr Kâ (λ-2.0840 Å) and Cr Ká (λ-2.2896 Å) radiations respectively. Typical 'M' type stress profile was observed across the weld centre line in both the welds. The variation and similarities between the longitudinal stress profiles observed in these two weld joints would be discussed. (author)

The effect of plastic strain on the evolution of crystallographic texture in Zircaloy-2

International Nuclear Information System (INIS)

Ballinger, R.G.; Lucas, G.E.; Pelloux, R.M.

1984-01-01

The evolution of crystallographic texture during plastic deformation was investigated in Zircaloy-2 using X-ray and metallographic techniques. Inverse pole figures, the resolved fraction of basal poles, and the volume fraction of twinned material, were determined as a function of plastic strain for several strain paths and initial textures at 298 K and 623 K. Incremental transverse platic strain ratios (R) were measured as a function of plastic strain. Texture rotation occurs early in the deformation process, after as little as 1.5% plastic strain. For compressive plastic strains, the resolved fraction of basal poles increases in the direction parallel to the strain axis. For tensile plastic strains, the resolved fraction of basal poles decreases in the direction parallel to the strain axis. The rate of change of the resolved fraction of basal poles with plastic strain is a function of the initial resolved fraction of basal poles. The texture rotation can be explained by considering the operating of the principal tensile twinning systems, [10anti 12], . (orig.)

Fracture of Zircaloy cladding by interactions with uranium dioxide pellets in LWR fuel rods. Technical report 10

International Nuclear Information System (INIS)

Smith, E.; Ranjan, G.V.; Cipolla, R.C.

1976-11-01

Power reactor fuel rod failures can be caused by uranium dioxide fuel pellet-Zircaloy cladding interactions. The report summarizes the current position attained in a detailed theoretical study of Zircaloy cladding fracture caused by the growth of stress corrosion cracks which form near fuel pellet cracks as a consequence of a power increase after a sufficiently high burn-up. It is shown that stress corrosion crack growth in irradiated Zircaloy must be able to proceed at very low stress intensifications if uniform friction effects are operative at the fuel-cladding interface, when the interfacial friction coefficient is less than unity, when a symmetric distribution of fuel cracks exists, and when symmetric interfacial slippage occurs (i.e., ''uniform'' conditions). Otherwise, the observed fuel rod failures must be due to departures from ''uniform'' conditions, and a very high interfacial friction coefficient and particularly fuel-cladding bonding, are means of providing sufficient stess intensification at a cladding crack tip to explain the occurrence of cladding fractures. The results of the investigation focus attention on the necessity for reliable experimental data on the stress corrosion crack growth behavior of irradiated Zircaloy, and for further investigations on the correlation between local fuel-cladding bonding and stress corrosion cracking

Oxide particle size distribution from shearing irradiated and unirradiated LWR fuels in Zircaloy and stainless steel cladding: significance for risk assessment

Energy Technology Data Exchange (ETDEWEB)

Davis, W. Jr.; West, G.A.; Stacy, R.G.

1979-03-22

Sieve fractionation was performed with oxide particles dislodged during shearing of unirradiated or irradiated fuel bundles or single rods of UO/sub 2/ or 96 to 97% ThO/sub 2/--3 to 4% UO/sub 2/. Analyses of these data by nonlinear least-squares techniques demonstrated that the particle size distribution is lognormal. Variables involved in the numerical analyses include lognormal median size, lognormal standard deviation, and shear cut length. Sieve-fractionation data are presented for unirradiated bundles of stainless-steel-clad or Zircaloy-2-clad UO/sub 2/ or ThO/sub 2/--UO/sub 2/ sheared into lengths from 0.5 to 2.0 in. Data are also presented for irradiated single rods (sheared into lengths of 0.25 to 2.0 in.) of Zircaloy-2-clad UO/sub 2/ from BWRs and of Zircaloy-4-clad UO/sub 2/ from PWRs. Median particle sizes of UO/sub 2/ from shearing irradiated stainless-steel-clad fuel ranged from 103 to 182 ..mu..m; particle sizes of ThO/sub 2/--UO/sub 2/, under these same conditions, ranged from 137 to 202 ..mu..m. Similarly, median particle sizes of UO/sub 2/ from shearing unirradiated Zircaloy-2-clad fuel ranged from 230 to 957 ..mu..m. Irradiation levels of fuels from reactors ranged from 9,000 to 28,000 MWd/MTU. In general, particle sizes from shearing these irradiated fuels are larger than those from the unirradiated fuels; however, unirradiated fuel from vendors was not available for performing comparative shearing experiments. In addition, variations in particle size parameters pertaining to samples of a single vendor varied as much as those between different vendors. The fraction of fuel dislodged from the cladding is nearly proportional to the reciprocal of the shear cut length, until the cut length attains some minimum value below which all fuel is dislodged. Particles of fuel are generally elongated with a long-to-short axis ratio usually less than 3. Using parameters of the lognormal distribution estimates can be made of fractions of dislodged fuel having

IMPROVEMENT OF WELDED CONNECTIONS WITH SIDE LAP WELDS BY REDISTRIBUTION OF ALL-WELD METAL ALONG LENGTHS AND CROSS-SECTIONS THEREOF USING MECHANIZED AND ROBOTIC WELDING SYSTEMS

Directory of Open Access Journals (Sweden)

Pavlov Evgeniy Igorevich

2017-05-01

Full Text Available Experimental study of bearing capacity of samples of two series performed by semiautomatic welding in CO2 on the axis, and by robotic welding machine in mixture (CO2 + Ar, is presented. Welds of constant cross section, welds with extended leg on end sections, and welds in the form of two dowels on end sections were performed. Efficiency of pilot samples of the first series (with extended leg on end sections by way of a smooth transition defined by the ratio of weld metal volume to a crushing load reaches 28 % relative to samples with a leg constant as per length. Samples of the first series with an extended leg on end sections also showed efficiency increased to 17 %. According to the second series samples test results, the exceeding of bearing capacity of the samples performed with an extended leg on end sections by 24 % in comparison with the samples with a leg of constant cross section was determined. Samples of the second series performed in the form of two dowels on end sections demonstrated the exceeding of the relative bearing capacity by 42 % in comparison with the samples with a continuous leg of constant cross-section.

Investigation on mechanical properties of welded material under different types of welding filler (shielded metal arc welding)

Science.gov (United States)

Tahir, Abdullah Mohd; Lair, Noor Ajian Mohd; Wei, Foo Jun

2018-05-01

The Shielded Metal Arc Welding (SMAW) is (or the Stick welding) defined as a welding process, which melts and joins metals with an arc between a welding filler (electrode rod) and the workpieces. The main objective was to study the mechanical properties of welded metal under different types of welding fillers and current for SMAW. This project utilized the Design of Experiment (DOE) by adopting the Full Factorial Design. The independent variables were the types of welding filler and welding current, whereas the other welding parameters were fixed at the optimum value. The levels for types of welding filler were by the models of welding filler (E6013, E7016 and E7018) used and the levels for welding current were 80A and 90A. The responses were the mechanical properties of welded material, which include tensile strength and hardness. The experiment was analyzed using the two way ANOVA. The results prove that there are significant effects of welding filler types and current levels on the tensile strength and hardness of the welded metal. At the same time, the ANOVA results and interaction plot indicate that there are significant interactions between the welding filler types and the welding current on both the hardness and tensile strength of the welded metals, which has never been reported before. This project found that when the amount of heat input with increase, the mechanical properties such as tensile strength and hardness decrease. The optimum tensile strength for welded metal is produced by the welding filler E7016 and the optimum of hardness of welded metal is produced by the welding filler E7018 at welding current of 80A.

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)

Effects of alloying element on weld characterization of laser-arc hybrid welding of pure copper

Science.gov (United States)

Hao, Kangda; Gong, Mengcheng; Xie, Yong; Gao, Ming; Zeng, Xiaoyan

2018-06-01

Effects of alloying elements of Si and Sn on weld characterizations of laser-arc hybrid welded pure copper (Cu) with thickness of 2 mm was studied in detail by using different wires. The weld microstructure was analyzed, and the mechanical properties (micro-hardness and tensile property), conductivity and corrosion resistance were tested. The results showed that the alloying elements benefit the growth of column grains within weld fusion zone (FZ), increase the ultimate tensile strength (UTS) of the FZ and weld corrosion resistance, and decrease weld conductivity. The mechanisms were discussed according to the results.

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)

Effects of δ-hydride precipitation at a crack tip on crack propagation in delayed hydride cracking of Zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Kubo, T., E-mail: kubo@nfd.co.jp [Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki 311-1313 (Japan); Kobayashi, Y. [M.O.X. Co., Ltd., 1828-520 Hirasu-cho, Mito, Ibaraki 311-0853 (Japan)

2013-08-15

Highlights: • Steady state crack velocity of delayed hydride cracking in Zircaloy-2 was analyzed. • A large stress peak is induced at an end of hydride by volume expansion of hydride. • Hydrogen diffuses to the stress peak, thereby accelerating steady hydride growth. • Crack velocity was estimated from the calculated hydrogen flux into the stress peak. • There was good agreement between calculation results and experimental data. -- Abstract: Delayed hydride cracking (DHC) of Zircaloy-2 is one possible mechanism for the failure of boiling water reactor fuel rods in ramp tests at high burnup. Analyses were made for hydrogen diffusion around a crack tip to estimate the crack velocity of DHC in zirconium alloys, placing importance on effects of precipitation of δ-hydride. The stress distribution around the crack tip is significantly altered by precipitation of hydride, which was strictly analyzed using a finite element computer code. Then, stress-driven hydrogen diffusion under the altered stress distribution was analyzed by a differential method. Overlapping of external stress and hydride precipitation at a crack tip induces two stress peaks; one at a crack tip and the other at the front end of the hydride precipitate. Since the latter is larger than the former, more hydrogen diffuses to the front end of the hydride precipitate, thereby accelerating hydride growth compared with that in the absence of the hydride. These results indicated that, after hydride was formed in front of the crack tip, it grew almost steadily accompanying the interaction of hydrogen diffusion, hydride growth and the stress alteration by hydride precipitation. Finally, crack velocity was estimated from the calculated hydrogen flux into the crack tip as a function of temperature, stress intensity factor and material strength. There was qualitatively good agreement between calculation results and experimental data.

Waste canister closure welding using the inertia friction welding process

International Nuclear Information System (INIS)

Klein, R.F.; Siemens, D.H.; Kuruzar, D.L.

1986-02-01

Liquid radioactive waste presently stored in underground tanks is to undergo a vitrifying process which will immobilize it in a solid form. This solid waste will be contained in a stainless steel canister. The canister opening requires a positive seal weld, the properties and thickness of which are at least equal to those of the canister material. This paper describes the inertia friction welding process and a proposed equipment design concept that will provide a positive, reliable, inspectable, and full thickness seal weld while providing easily maintainable equipment, even though the weld is made in a highly contaminated hot cell. All studies and tests performed have shown the concept to be highly feasible. 2 refs., 6 figs

Critical element development of standard components for pipe welding/cutting by CO{sub 2} laser

Energy Technology Data Exchange (ETDEWEB)

Oka, Kiyoshi; Kakudate, Satoshi; Nakahira, Masataka [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

1994-11-01

In D-T burning reactors such as International Thermonuclear Experimental Reactor(ITER), an internal access is inevitable for welding/cutting of cooling pipes of in-vessel components, because of spatial constraint due to a narrow port opening space. An internal-access pipe welding/cutting equipment is being developed in JAERI. Internal access is to approach through inside a pipe to a welding/cutting position, to use 10kW CO{sub 2} laser beam, and to be applicable to both welding and cutting with using a same processing head. A welding/cutting processing head with 10kW CO{sub 2} laser beam has been fabricated and the basic feasibility has been successfully demonstrated for studies of the internal-access pipe welding/cutting concept using 100-A stainless steel pipe with a thickness of 6.3mm. In this study, the optimum focal point of laser beam, laser power and traveling speed of the head have been investigated together with an adjusting mechanism of a relative distance between the head and the pipe wall. In addition, the radiation resistance of critical elements such as optical lens has been investigated. (author).

Welding for the CRBRP steam generators

International Nuclear Information System (INIS)

Spalaris, C.N.; Ring, P.J.; Durand, R.E.; Wright, E.A.

1979-01-01

The rationale for selecting weld design, welding procedures and inspection methods was based upon the desire to obtain the highest reliability welds for the CRBRP steam generators. To assure the highest weld reliability, heavy emphasis was placed on the control of material cleanliness and composition substantially exceeding the requirements of the ASME Code for 2-1/4Cr--1Mo. The high tube/tubesheet weld quality was achieved through close material control, an extensive weld development program and the selection of high reliability welding equipment. Shell and nozzle weld fabrication using TIG, MIG, and submerged arc procedures are also being controlled through precise specifications, including preheat and postheat programs, together with radiography and ultrasonic inspection to ascertain the weld quality desired. Details of the tube/tubesheet welding and shell welding are described and results from the weld testing program are discussed

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

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.

Spot Welding Characterizations With Time Variable

International Nuclear Information System (INIS)

Abdul Hafid; Pinitoyo, A.; History; Paidjo, Andryansyah; Sagino, Sudarmin; Tamzil, M.

2001-01-01

For obtain spot welding used effective data, this research is made, so that time operational of machine increasing. Welding parameters are material classification, electrical current, and weld time. All of the factors are determined welding quality. If the plate more thick, the time must be longer when the current constant. Another factor as determined welding quality are surface condition of electrode, surface condition of weld material, and material classifications. In this research, the weld machine type IP32A2 VI (110 V), Rivoira trademark is characterized

Creep and rupture behavior of weld-deposited Type 16-8-2 stainless steel at 5930C

International Nuclear Information System (INIS)

Ward, A.L.; Blackburn, L.D.

1976-03-01

The creep and rupture behavior of weld-deposited Type 16-8-2 stainless steel at 593 0 C was investigated over the time range from 3.6 x 10 4 s to 2.5 x 10 7 s. Equations relating stress to the time to rupture, the time to the onset of tertiary creep, and the time to produce a given creep strain were obtained. The experimental results indicate that the control of welding parameters (e.g. current, voltage and travel speed) within reasonable ranges can yield weld deposits with consistent time-dependent properties. Limited data suggest that high temperature (1065 0 C) post-weld annealing significantly alters only the flow curve for plastic deformation, while long-term thermal exposure at an intermediate temperature (565 0 C) produces only minor changes in either the plastic deformation or creep behavior of the weld materials

Handbook of welding engineering. Vol. 1 and Vol. 2. 2. rev. ed.

International Nuclear Information System (INIS)

Ruge, J.

1980-01-01

This second edition of the handbook still has been guided by the principle of presenting as comprehensive information as possible on the whole subject field of welding engineering as concisely as seems adequate. The task of completely revising the first edition has not been restricted to up-dating the standards, guidelines and instruction sheets. It rather also seemed appropriate to amend the text in many cases in order to incorporate the latest results of research in science and technology. This inevitably enlarged the material to an extent recommending a publication in two volumes. Volume I deals with materials problems, and the sections discussing technical aspects of fracture mechanics and the welding of high-alloy steels have been enlarged. The section on nonmetals has been supplemented by a more detailed treatment of plastics and by chapters on other nonmetals such as glass, ceramics, graphite, and biological substances. Volume II deals with welding techniques, fabrication and quality assurance. Apart from the methods of welding, cutting, soldering, bonding (adhesives), and thermal spray coating, methods of improving the efficiency of fabrication by means of numerically controlled welding and process control are discussed in detail. Taking into account the growing importance of quality assurance, new chapters on modern control methods have been incorporated, methods such as control by neutron radiation, xeroradiography and acoustic and optical holography, as well as a section on distortion and buckling. The chapters on welding and cutting under water in marine technology, on occupational safety and economic aspects have been considerably enlarged. (orig./IHOE) [de

Study of liquid phase formation kinetics due to solid/solid chemical interaction and its model. Application to the Zircaloy/Inconel

International Nuclear Information System (INIS)

Garcia, E.A.; Denis, A.

1990-01-01

A description is made of the chemical interaction between Inconel spacing grids and the Zircaloy of the sheaths. Experiments performed at 1000, 1100 and 1200 deg C with base Zircaloy and with a previously formed layer of ZrO 2 , show that the kinetics is parabolic. The difference between both types of experiments is that the oxide layer delays the initiation of the Inconel-Zry interaction. A model is presented, for the description of the solid/solid interaction, which leads to the formation of eutectic that is liquid at the experiment temperature. Also a model, which represents the oxide layer dissolution and predicts the instant in which it disappears completely, is presented. (Author) [es

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)

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.

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