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.

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)

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

International Nuclear Information System (INIS)

Holub, F.

1987-08-01

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

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)

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

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)

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.

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

Cold spray deposition of Ti{sub 2}AlC coatings for improved nuclear fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Maier, Benjamin R. [University of Wisconsin, Madison, WI (United States); Garcia-Diaz, Brenda L. [Savannah River National Laboratory, Aiken, SC (United States); Hauch, Benjamin [University of Wisconsin, Madison, WI (United States); Olson, Luke C.; Sindelar, Robert L. [Savannah River National Laboratory, Aiken, SC (United States); Sridharan, Kumar, E-mail: kumar@engr.wisc.edu [University of Wisconsin, Madison, WI (United States)

2015-11-15

Coatings of Ti{sub 2}AlC MAX phase compound have been successfully deposited on Zircaloy-4 (Zry-4) test flats, with the goal of enhancing the accident tolerance of LWR fuel cladding. Low temperature powder spray process, also known as cold spray, has been used to deposit coatings ∼90 μm in thickness using powder particles of <20 μm. X-ray diffraction analysis showed the phase-content of the deposited coatings to be identical to the powders indicating that no phase transformation or oxidation had occurred during the coating deposition process. The coating exhibited a high hardness of about 800 H{sub K} and pin-on-disk wear tests using abrasive ruby ball counter-surface showed the wear resistance of the coating to be significantly superior to the Zry-4 substrate. Scratch tests revealed the coatings to be well-adhered to the Zry-4 substrate. Such mechanical integrity is required for claddings from the standpoint of fretting wear resistance and resisting wear handling and insertion. Air oxidation tests at 700 °C and simulated LOCA tests at 1005 °C in steam environment showed the coatings to be significantly more oxidation resistant compared to Zry-4 suggesting that such coatings can potentially provide accident tolerance to nuclear fuel cladding. - Highlights: • Deposited Ti{sub 2}AlC coatings on Zircaloy-4 substrates with a low pressure powder spray process, also known as cold spray. • Coatings have high hardness and wear resistance for both damage resistance during rod insertion and fretting wear resistance. • The oxidation resistance of Ti{sub 2}AlC coated Zircaloy-4 at 700 °C and 1005 °C was significantly superior to uncoated Zircaloy. • Cold spray of Ti{sub 2}AlC demonstrates considerable promise as a near-term solution for accident tolerant Zr-alloy fuel claddings.

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

Energy Technology Data Exchange (ETDEWEB)

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

1963-07-01

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

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

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)

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.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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)

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

Microstructure and crystallographic texture evolution during TIG welding of zircaloy-2 material

International Nuclear Information System (INIS)

Jha, S.K.; Singh, R.P.; Singh, V.K.; Ramanathan, R.; Samjdar, I.; Srivastava, D.; Tewari, R.; Dey, G.K.

2005-01-01

Zirconium and its alloys are extensively used as structural materials in nuclear reactors, because of better neutron economy, good corrosion resistance in water and good mechanical properties at operating temperature. Zircaloy-2 and zircaloy-4 are widely used in both pressurized water reactors (PWR) and boiling water reactors (BWR) as fuel cladding materials and as calandria tube and pressure tube materials in pressurized heavy water reactors (PHWR). The satisfactory performance and the life of the reactor components depend mainly upon their mechanical properties, corrosion properties and dimensional stability in the reactor condition, which are strong function of metallurgical parameters such as microstructure and texture. Therefore, for best performance of the reactor components these parameters are optimized during their fabrication. The microstructure and texture of the zircaloy-2 components are expected to get modified during the welding of the components. In this study the evolution of the microstructure and texture has been investigated as a function of the welding parameters. Heat input was varied the current and welding time. A variety of analytical techniques have been applied for the study on microstructure and texture of the welds. Optical microscopy and electron microscopy were used to evaluate the detailed microstructure. X-ray diffraction (XRD) was used investigate the crystallographic textures among the base metal, heat affected zone and fusion zone. Particular attention was focused on the determination of microtexture in weld by using electron backscatter diffraction (EBSD) technique. After that, an effort was put to compare the results of X-ray macro-texture and EBS-microtexture. (author)

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.

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)

Contribuciones de Sir Roland Fisher a la Estadística Genética

Directory of Open Access Journals (Sweden)

Jaime Cuadros

2004-11-01

Full Text Available Sir Ronald Fisher (18901962 fue profesor de genética y muchas de sus innovaciones estadísticas encontraron expresión en el desarrollo de metodología en estadística genética. Sin embargo, mientras sus contribuciones en estadística matemática son fácilmente identificadas, en genética de poblaciones compartió su supremacía con Sewall Wright (1889 1988 y J. S. S. Haldane (1892 1965. Este documento muestra algunas de las mejores contribuciones de Fisher a las bases de la estadística genética, y sus interacciones con Wright y Haldane, los cuales contribuyeron al desarrollo del tema. Con la tecnología moderna, tanto la metodología la estadística como la información genética están cambiando. No obstante, muchos de los trabajos de Fisher permanecen relevantes, y pueden aun servir como una base para investigaciones futuras en el análisis estadístico de datos de DNA. El trabajo de este autor refleja su visión del papel de Ia estadística en Ia inferencia científica expresada en 1949

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

Science.gov (United States)

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

2017-05-01

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

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.)

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

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

Zircaloy-sheathed element rods fitted with thermo-couples

International Nuclear Information System (INIS)

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

1963-01-01

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

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

Diffusionless bonding of aluminum to Zircaloy-2

International Nuclear Information System (INIS)

Watson, R.D.

1965-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-08-01

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

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)

Effect of negative bias on TiAlSiN coating deposited on nitrided Zircaloy-4

Science.gov (United States)

Jun, Zhou; Zhendong, Feng; Xiangfang, Fan; Yanhong, Liu; Huanlin, Li

2018-01-01

TiAlSiN coatings were deposited on the nitrided Zircaloy-4 by multi-arc ion plating at -100 V, -200 V and -300 V. In this study, the high temperature oxidation behavior of coatings was tested by a box-type resistance furnace in air for 3 h at 800 °C; the macro-morphology of coatings was observed and analyzed by a zoom-stereo microscope; the micro-morphology of coatings was analyzed by a scanning electron microscopy (SEM), and the chemical elements of samples were analyzed by an energy dispersive spectroscopy(EDS); the adhesion strength of the coating to the substrate was measured by an automatic scratch tester; and the phases of coatings were analyzed by an X-ray diffractometer(XRD). Results show that the coating deposited at -100 V shows better high temperature oxidation resistance behavior, at the same time, Al elements contained in the coating is of the highest amount, meanwhile, the adhesion strength of the coating to the substrate is the highest, which is 33N. As the bias increases, high temperature oxidation resistance behavior of the coating weakens first and then increases, the amount of large particles on the surface of the coating increases first and then decreases whereas the density of the coating decreases first and then increases, and adhesion strength of the coating to the substrate increases first and then weakens. The coating's quality is relatively poor when the bias is -200 V.

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)

Performance of iron–chromium–aluminum alloy surface coatings on Zircaloy 2 under high-temperature steam and normal BWR operating conditions

Energy Technology Data Exchange (ETDEWEB)

Zhong, Weicheng; Mouche, Peter A.; Han, Xiaochun [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Heuser, Brent J., E-mail: bheuser@illinois.edu [University of Illinois, Department of Nuclear, Radiological, and Plasma Engineering, Urbana, IL 61801 (United States); Mandapaka, Kiran K.; Was, Gary S. [University of Michigan, Department of Nuclear Engineering and Radiological Sciences, Ann Arbor, MI 48109 (United States)

2016-03-15

Iron-chromium-aluminum (FeCrAl) coatings deposited on Zircaloy 2 (Zy2) and yttria-stabilized zirconia (YSZ) by magnetron sputtering have been tested with respect to oxidation weight gain in high-temperature steam. In addition, autoclave testing of FeCrAl-coated Zy2 coupons under pressure-temperature-dissolved oxygen coolant conditions representative of a boiling water reactor (BWR) environment has been performed. Four different FeCrAl compositions have been tested in 700 °C steam; compositions that promote alumina formation inhibited oxidation of the underlying Zy2. Parabolic growth kinetics of alumina on FeCrAl-coated Zy2 is quantified via elemental depth profiling. Autoclave testing under normal BWR operating conditions (288 °C, 9.5 MPa with normal water chemistry) up to 20 days demonstrates observable weight gain over uncoated Zy2 simultaneously exposed to the same environment. However, no FeCrAl film degradation was observed. The 900 °C eutectic in binary Fe–Zr is addressed with the FeCrAl-YSZ system. - Graphical abstract: Weight gain normalized to total sample surface area versus time during 700 °C steam exposure for FeCrAl samples with different composition (A) and Fe/Cr/Al:62/4/34 (B). In both cases, the responses of uncoated Zry2 (Zry2-13A and Zry2-19A) are shown for comparison. This uncoated Zry2 response shows the expected pre-transition quasi-cubic kinetic behavior and eventual breakaway (linear) kinetics. Highlights: • FeCrAl coatings deposited on Zy2 have been tested with respect to oxidation in high-temperature steam. • FeCrAl compositions promoting alumina formation inhibited oxidation of Zy2 and delay weight gain. • Autoclave testing to 20 days of coated Zy2 in a simulated BWR environment demonstrates minimal weight gain and no film degradation. • The 900 °C eutectic in binary Fe-Zr is addressed with the FeCrAl-YSZ system.

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

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

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

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)

The effect of neutron irradiation on the mechanical properties of welded zircaloy-2

Energy Technology Data Exchange (ETDEWEB)

Evans, D G

1962-07-15

Zircaloy-2 tensile specimens, subsize impact bars and representative spigot welds were subjected to three NRX cycles in the X-5 loop. Average loop temperature was 260{sup o}C over the three cycles. One group of tensile specimens was heat-treated in vacuum at 900{sup o}C for 40 minutes, another group contained welded areas in the centre of the gauge length and a third group was hydrided after welding. Notches of the impact specimens were located in the fusion zone of the weld, Spigot welds were made on autoclaved and unautoclaved simulated production assemblies. The transition temperature of Zircaloy-2 increased appreciably upon welding. This was accompanied by a decrease in absorbed energy values for all temperatures between 0{sup o} and 300{sup o}C. Neutron irradiation had no effect on the impact properties of welded. Zircaloy-2. Welding decreased the uniform and total elongation at room temperature and at 260{sup o}C, and increased the 260{sup o}C PL, YS and UTS. Hydriding to a nominal 100 ppm hydrogen had no effect on the unirradiated tensile properties at either test temperature. The heat treatment decreased the strength properties but did not affect the ductility. Neutron irradiation increased the YS of the welded and hydrided material by 20% and the heat treated YS by 40%. Irradiation also increased the 260{sup o}C strength properties of the as-welded material. It was found that the unautoclaved spigot welds had a generally higher tensile strength than the autoclaved and welded specimens. For specimens welded in either condition, the outer welds of the 19-element bundle had a lower average breaking load than the inner welds. Neutron irradiation had no effect on the tensile strength of these welds. It was also demonstrated that a cup-and-cone type of fracture could be produced in a bend test. These fractures were similar to those observed in irradiated fuel bundles which had been damaged during transfer operations. A large amount of scatter rendered some

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)

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.

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.

Characterization of electron beam welded Zircaloy-4

International Nuclear Information System (INIS)

Anishetty, Sharath; Manna, I.; Majumdar, J. Dutta

2015-01-01

Zirconium (Zr) alloys are the backbone materials for thermal reactors because of their low neutron absorption cross section and in addition have suitable properties like high temperature mechanical and corrosion properties. For various structural applications, different Zirconium based alloys are used. Zircaloy-4 (Zr-4) is most commonly used as channel boxes in boiling water reactors (BWRs), intermediate grid applications in pressurized water reactors (PWRs) and in fuel cladding. Zircaloy cladding acts as a barrier between the radioactive fuel and exterior coolants. Therefore, the structural integrity of the cladding tube is extremely important in the safe operation of reactors. Efforts are being made to produce Zircaloy-4 products with better mechanical properties. Different routes of processing are involved like forging, pilgering and extrusion are developed over years in fabricating components to improve in-reactor performance. In this study, microstructure and hardness properties of electron beam welded Zr-4 was evaluated

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)

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

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.

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)

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.

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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.

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)

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

Comparison study between GTWA and PAW welding techniques in zircaloy-4

International Nuclear Information System (INIS)

Martinez, R.L.; Boccanera, L.; Ortiz, L.; Fernandez, L.; Corso, H.

2003-01-01

The wide use of zirconium alloys in different structural parts of nuclear reactors mainly under severe environmental conditions has encouraged the study of Zircaloy-4 and specifically welded joints of this material.Many different factors affect mechanical properties, specifically hydrides, formed by absorbed hydrogen.Hydrogen solubility in Zircaloy-4 is low and because Zircaloy-4 picks up hydrogen during service the potential exist that zirconium hydrides phase precipitate causing loss of ductility, the most undesirable consequence. Therefore, the study and characterization of welded joint of nuclear materials assumes fundamental importance in the safety of nuclear reactors.This paper presents experimental results regarding of hardness and hydrogen concentration in Zircaloy-4 plates obtained by two different welding techniques GTWA (Gas Tungsten Arc Welding) and PAW (Plasma Arc Welding).In this work following these remarks the difference observed between these two techniques are presented and point out some aspects of PAW for further discussion

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)

Characterisation of metallic glass incorporated Zircaloy-2 weldments

International Nuclear Information System (INIS)

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

1995-01-01

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

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)

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.)

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.)

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

Linear Friction Welding of Dissimilar Materials 316L Stainless Steel to Zircaloy-4

Science.gov (United States)

Wanjara, P.; Naik, B. S.; Yang, Q.; Cao, X.; Gholipour, J.; Chen, D. L.

2018-02-01

In the nuclear industry, there are a number of applications where the transition of stainless steel to Zircaloy is of technological importance. However, due to the differences in their properties there are considerable challenges associated with developing a joining process that will sufficiently limit the heat input and welding time—so as to minimize the extent of interaction at the joint interface and the resulting formation of intermetallic compounds—but still render a functional metallurgical bond between these two alloys. As such, linear friction welding, a solid-state joining technology, was selected in the present study to assess the feasibility of welding 316L stainless steel to Zircaloy-4. The dissimilar alloy welds were examined to evaluate their microstructural characteristics, microhardness evolution across the joint interface, static tensile properties, and fatigue behavior. Microstructural observations revealed a central intermixed region and, on the Zircaloy-4 side, dynamically recrystallized and thermomechanically affected zones were present. By contrast, deformation on the 316L stainless steel side was limited. In the intermixed region a drastic change in the composition was observed along with a local increase in hardness, which was attributed to the presence of intermetallic compounds, such as FeZr3 and Cr2Zr. The average yield (316 MPa) and ultimate tensile (421 MPa) strengths met the minimum strength properties of Zircaloy-4, but the elongation was relatively low ( 2 pct). The tensile and fatigue fracture of the welds always occurred at the interface in the mode of partial cohesive failure.

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)

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

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.)

Fuel Performance Calculations for FeCrAl Cladding in BWRs

Energy Technology Data Exchange (ETDEWEB)

George, Nathan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Sweet, Ryan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Maldonado, G. Ivan [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Wirth, Brian D. [Univ. of Tennessee, Knoxville, TN (United States). Dept. of Nuclear Engineering; Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-01-01

This study expands upon previous neutronics analyses of the reactivity impact of alternate cladding concepts in boiling water reactor (BWR) cores and directs focus toward contrasting fuel performance characteristics of FeCrAl cladding against those of traditional Zircaloy. Using neutronics results from a modern version of the 3D nodal simulator NESTLE, linear power histories were generated and supplied to the BISON-CASL code for fuel performance evaluations. BISON-CASL (formerly Peregrine) expands on material libraries implemented in the BISON fuel performance code and the MOOSE framework by providing proprietary material data. By creating material libraries for Zircaloy and FeCrAl cladding, the thermomechanical behavior of the fuel rod (e.g., strains, centerline fuel temperature, and time to gap closure) were investigated and contrasted.

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)

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

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)

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)

Chemical and X-ray diffraction analysis on selected samples from the TMI-2 reactor core

International Nuclear Information System (INIS)

Kleykamp, H.; Pejsa, R.

1991-05-01

Selected samples from different positions of the damaged TMI-2 reactor core were investigated by X-ray microanalysis and X-ray diffraction. The measurements yield the following resolidified phases after cooling: Cd and In depleted Ag absorber material, intermetallic Zr-steel compounds, fully oxidized Zircaloy, UO 2 -ZrO 2 solid solutions and their decomposed phases, and Fe-Al-Cr-Zr spinels. The composition of the phases and their lattice parameters as well as the eutectic and monotectic character can serve as indicators of local temperatures of the core. The reaction sequences are estimated from the heterogeneous equilibria of these phases. The main conclusions are: (1) Liquefaction onset is locally possible by Inconel-Zircaloy and steel-Zircaloy reactions of spacers and absorber guide tubes at 930deg C. However, increased rates of dissolution occur above 1200deg C. (2) UO 2 dissolution in the Inconel-steel-Zircaloy melt starts at 1300deg C with increased rates above 1900deg C. (3) Fuel temperatures in the core centre are increased above 2550deg C, liquid (U,Zr)O 2 is generated. (4) Square UO 2 particles are reprecipitated from the Incoloy-steel-Zircaloy-UO 2 melt during cooling, the remaining metallic melt is oxygen poor; two types of intermetallic phases are formed. (5) Oxidized Fe and Zr and Al 2 O 3 from burnable absorber react to spinels which form a low melting eutectic with the fuel at 1500deg C. The spinel acts as lubricant for fuel transport to the lower reactor plenum above 1500deg C. (6) Ruthenium (Ru-106) is dissolved in the steel phase, antimony (Sb-125) in the α-Ag absorber during liquefaction. (7) Oxidation of the Zircaloy-steel phases takes place mainly in the reflood stage 3 of the accident scenario. (orig.) [de

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)

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

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

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

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

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)

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.

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

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

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.)

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.)

Contribuciones al estudiode los anfibios y reptiles de Méxicodurante el siglo XVIII y la Ilustración

Directory of Open Access Journals (Sweden)

Gustavo Casas Andreu

2008-01-01

Full Text Available Las bases de la herpetología moderna en general se establecieron en el siglo XVIII,particularmente durante la Ilustración y quienes hicieron la mayor contribución fueronfundamentalmente los naturalistas franceses. No obstante, en México se hicieron varias publicacionesde cierto relieve, en especial por los misioneros jesuitas y otros estudiosos de la Nueva España. Esinteresante mencionar que aun con la trascendencia de Linneo para la biología moderna, algunoshistoriadores de los anfibios y los reptiles señalan que por lo menos para la herpetofauna de Méxicoexistió una importante regresión, ya que era mucho mayor el conocimiento que había dejadoFrancisco Hernández en el siglo XVI. Las contribuciones de los autores de la escuela francesa comoBuffon y quienes lo sucedieron como Lacepéde y Daudin, fueron los grandes pilares de laherpetología o estudios de los anfibios y los reptiles de la manera en que la conocemos en laactualidad. Las bases establecidas en el siglo XVIII sirvieron para que con la apertura de México almundo a partir de su independencia, se entrara en una de las etapas de mayor relevancia para laherpetología del país.

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)

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.

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)

Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

Energy Technology Data Exchange (ETDEWEB)

Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wirth, B. D. [Univ. of Tennessee, Knoxville, TN (United States)

2017-07-26

Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. This allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to

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)

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).

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

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)

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

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)

303-K Storage Facility closure plan. Revision 2

Energy Technology Data Exchange (ETDEWEB)

1993-12-15

Recyclable scrap uranium with zircaloy-2 and copper silicon alloy, uranium-titanium alloy, beryllium/zircaloy-2 alloy, and zircaloy-2 chips and fines were secured in concrete billets (7.5-gallon containers) in the 303-K Storage Facility, located in the 300 Area. The beryllium/zircaloy-2 alloy and zircaloy-2 chips and fines are designated as mixed waste with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 303-K Storage Facility is now undergoing closure as defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and the Washington Administrative Code (WAC) Dangerous Waste Regulations, WAC 173-303-040. This closure plan presents a description of the 303-K Storage Facility, the history of materials and waste managed, and the procedures that will be followed to close the 303-K Storage Facility. The 303-K Storage Facility is located within the 300-FF-3 (source) and 300-FF-5 (groundwater) operable units, as designated in the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology et al. 1992). Contamination in the operable units 300-FF-3 and 300-FF-5 is scheduled to be addressed through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of 1980 remedial action process. Therefore, all soil remedial action at the 304 Facility will be conducted as part of the CERCLA remedial action of operable units 300-FF-3 and 300-FF-5.

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

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}.

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)

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

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)

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.)

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

Development of remote welding technology for nuclear fuel end capping (A study on the weldability of Zircaloy-4)

Energy Technology Data Exchange (ETDEWEB)

Kho, Jin Hyun; Sung, Ho Hyun; Hyun, Yong Kyu; Suh, Hee Kang [Korea University of Technology and Education, Cheonan (Korea)

1998-03-01

The integrity of nuclear fuel end cap welds is essential to the nuclear fuel performance and safety as well as the usability of power plant. The first aim of this project is to obtain experimental data on the nuclear fuel cladding materials of Zircaloy-4 with welding processes such as plasma arc, gas tungsten arc and laser beam welding. the data obtained in this study will be applicable to the nuclear fuel design, fabrication and nuclear fuel quality control. In addition, the welding processes applicable to the Zircaloy-4 welding were compared and contrasted. The weldability of Zircaloy-4 was evaluated from the metallurgical and mechanical standpoints. 88 refs., 57 figs., 16 tabs. (Author)

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

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

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)

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

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

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.)

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)

Combustion synthesis of AlB2-Al2O3 composite powders with AlB2 nanowire structures

Science.gov (United States)

Yang, Pan; Xiao, Guoqing; Ding, Donghai; Ren, Yun; Yang, Shoulei; Lv, Lihua; Hou, Xing

2018-05-01

Using of Al and B2O3 powders as starting materials, and Mg-Al alloy as additives, AlB2-Al2O3 composite powders with AlB2 nanowire structures were successfully fabricated via combustion synthesis method in Ar atmosphere at a pressure of 1.5 MPa. The effect of different amount of Mg-Al alloy on the phase compositions and morphology of the combustion products was investigated. The results revealed that AlB2 and Al2O3 increased, whereas Al decreased with the content of Mg-Al alloy increasing. The impurities MgAl2O4 and AlB12 would exist in the sample with adding of 18 wt% Mg-Al alloy. Interestingly, FESEM/TEM/EDS results showed that AlB2 nanowires were observed in the products when the content of Mg-Al alloy is 6 wt% and 12 wt%. The more AlB2 nanowires can be found as the content of Mg-Al alloy increased. And the yield of AlB2 nanowires with the diameter of about 200 nanometers (nm) and the length up to several tens of micrometers (μm) in the combustion product is highest when the content of Mg-Al alloy is 12 wt%. The vapor, such as Mg-Al (g), B2O2 (g), AlO (g) and Al2O (g), produced during the process of combustion synthesis, reacted with each other to yield AlB2 nanowires by vapor-solid (VS) mechanism and the corresponding model was also proposed.

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

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.)

Examen de auditoría integral a las contribuciones de mejoras por el período comprendido entre el 1 de Enero y el 31 de diciembre del 2012, en el Gobierno Municipal del Cantón Morona

OpenAIRE

Valencia Zabala, Liliam Margot

2014-01-01

En el presente trabajo de investigación trata sobre una auditoría integral al Gobierno Municipal del Cantón Morona, se han diseñado capítulos que se relacionan con el desarrollo de una auditoría integral a las contribuciones de mejoras, que ha sido desarrollado con el aporte de los funcionarios auditados y se ha evaluado con los conocimientos adquiridos en esta maestría. Se han utilizado conceptos claves, que son parte fundamental de la auditoría integral como son: auditoría de gesti...

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

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)

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.

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

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

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

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.

Contribuciones bibliográficas de la Facultad de Farmacia y Bioquímica

OpenAIRE

Bonilla Rivera, Pablo E.

2014-01-01

La contribución bibliográfica de la facultad de Farmacia y Bioquímica de San Marcos es muy importante, pero tal vez poco conocida en los claustros sanmarquinos. Por ello, a partir de este número, "Ciencia e Investigación" incluirá las listas de las contribuciones hechas por los docentes de la facultad y por los exalumnos a través de sus tesis de título y de grados académicos. La contribución bibliográfica de la facultad de Farmacia y Bioquímica de San Marcos es muy importante, pero tal vez...

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

International Nuclear Information System (INIS)

Maeng, Yunhwan; Lee, Jaeyoung; Park, Sanggil

2016-01-01

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

Interim report on the creepdown of Zircaloy fuel cladding

International Nuclear Information System (INIS)

Hobson, D.O.; Dodd, C.V.

1977-01-01

This report describes the creepdown phenomenon in Zircaloy fuel cladding and the methods by which it will be measured and analyzed. Instrumentation for monitoring radial deformation in the cladding is described in detail--in terms of theory, design, and stability. The programs that control the microcomputer are listed, both to document the level of sophistication of the instrumentation and to indicate the flexibility of the test equipment

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

International Nuclear Information System (INIS)

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

2010-01-01

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

Fabrication and post-irradiation examination of a zircaloy-2 clad UO2-1.5 wt% PuO2 fuel pin irradiated in PWL, CIRUS

International Nuclear Information System (INIS)

Sah, D.N.; Sahoo, K.C.; Chatterjee, S.; Majumdar, S.; Kamath, H.S.; Ramachandran, R.; Bahl, J.K.; Purushottam, D.S.C.; Ramakumar, M.S.; Sivaramakrishnan, K.S.; Roy, P.R.

1977-01-01

A zircaloy-2 clad UO 2 -1.5 wt% PuO 2 fuel pin was fabricated at the Radiometallurgy Section of the Bhabha Atomic Research Centre, Bombay, for irradiation in the pressurised water loop in CIRUS. Requisite development work related to powder conditioning, blending, pressing and sintering parameters was carried out to meet the exacting fuel pellet specifications of CANDU fuel. The fuel pin ruptured while being irradiated in the pressurised water loop in CIRUS, after experiencing a low burn-up of 507 MWD/MTM and was subsequently examined at the Radiometallurgy Hot Cells Facility. The results showed that internal clad hydriding led to primary failure of the fuel pin. Subsequent ingress of the coolant water caused excessive swelling of the thermal insulating magnesia pellets located at the ends of the fuel column. The swelling of magnesia pellets caused severe rupturing of the fuel pin at the two ends. The delayed rupturing of the fuel pin at the upper end, caused the fuel column to be displaced downwards by 5.85mm. (author)

A phenomenological model for iodine stress corrosion cracking of zircaloy

International Nuclear Information System (INIS)

Miller, A.K.; Tasooji, A.

1981-01-01

To predict the response of Zircaloy tubing in iodine environments under conditions where either crack initiation or crack propagation predominates, a unified model of the SCC process has been developed based on the local conditions (the local stress, local strain, and local iodine concentration) within a small volume of material at the cladding inner surface or the crack tip. The methodology used permits computation of these values from simple equations. A nonuniform distribution of local stress and strain results once a crack has initiated. The local stress can be increased due to plastic constraint and triaxiality at the crack tip. Iodine penetration is assumed to be a surface diffusion-controlled process. Experimental data are used to derive criteria for intergranular failure, transgranular failure, and ductile rupture in terms of the local conditions. The same failure criteria are used for both crack initiation and crack propagation. Irradiation effects are included in the model by changing the value of constants in the equation governing iodine penetration and by changing the values used to represent the mechanical properties of the Zircaloy. (orig./HP)

NIRVANA, a high-temperature creep model for Zircaloy fuel sheathing

International Nuclear Information System (INIS)

Sills, H.E.; Holt, R.A.

1979-05-01

We have developed a multi-component model to describe the transient plastic deformation of Zircaloy fuel sheathing during high-temperature transients. From deformation maps we identify three deformation mechanisms which, in principle, occur in all three phase fields of Zircaloy (α, α+β, β): diffusional creep, dislocation creep, and athermal strian. A strain component occurring during the α → β transformation is also identified. Microstructural changes which alter deformation rates -grain structure, recrystallization, phase transformation -are accounted for. The individual components of the model represent known metallurgical phenomena. The combined model gives excellent agreement with transient test data from 700-1800 K, a range of heating rates from 0-100 K.s -1 , and a range of strain rates from 10 -5 to 10 -1 .s -1 . To enable comparison with available data the transient creep model was combined with an axially uniform, thin-walled tube representation having anisotropic material properties. The resulting computer code, NIRVANA provides facilities for simulating uniaxial and biaxial tube tests over specified stress/temperature histories. (author)

Temperature escalation in PWR fuel rod simulator bundles due to the zircaloy/steam reaction: Post test investigations of bundle test ESBU-2A

International Nuclear Information System (INIS)

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

1986-11-01

This KfK report describes the post test investigation of bundle experiment ESBU-2a. ESBU-2a was the second of two bundle tests on the temperature escalation of zircaloy clad fuel rods. The investigation of the temperature escalation is part of the program of out-of-pile experiments performed within the frame work of the PNS-Severe Fuel Damage program. The bundle was composed of a 3x3 fuel rod array of our fuel rod simulators (central tungsten heater, UO 2 -ring pellet and zircaloy cladding). The length was 0.4 meter. The bundle was heated to a maximum temperature of 2175 0 C. Molten cladding which dissolved part of the UO 2 pellets and slumped away from the already oxidized cladding formed a lump in the lower part of the bundle. After the test the bundle was embedded in epoxy and sectioned with a diamand saw, in the region of the refrozen melt. The cross sections were investigated by metallographic examination. The refrozen (U,Zr,O) melt consists variously of three phases with increasing oxygen content (metallic α-Zry, metallic (U,Zr) alloy and a (U,Zr)O 2 mixed oxide), two phases (α-Zry, (U,Zr)O 2 mixed oxide), or one phase ((U,Zr)O 2 mixed oxide). The cross sections show the increasing oxidation of the cladding with increasing elevation (temperature). A strong azimuthal dependency of the oxidation is found. In regions where the initial oxidized cladding is contacted by the melt one can recognize the interaction between the metallic melt and ZrO 2 of the cladding. Oxygen is taken away from the ZrO 2 . If the melt is in direct contact with steam a relatively well defined oxide layer is formed. (orig.) [de

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-04-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Influence of deformation history on texture change and subsequent yield locus of zircaloy-2 tubing

International Nuclear Information System (INIS)

Nagai, Nobuyuki; Kakuma, Tsutomu; Miyamoto, Yoshiyuki

1981-01-01

Fully-annealed Zircaloy-2 tubing was strained by balanced axial stress σsub(z) and circumferential stress σsub(theta) (stress ratio: α = σsub(z)/σsub(theta)). Then, texture and subsequent yield loci of these prestrained materials were measured. Results of texture measurement after prestraining showed that (0002) poles tend to move toward the radial tube direction under α = 0, 0.5 and 1, but toward the circumferential tube direction under α = 2 and infinity. Specimens highly prestrained under α = 0 and 0.5 have extremely concentrated texture. Such texture changes can be explained by a deformation model in which type slip system was assumed as one of the deformation system. The yield strength of most prestrained materials is higher than that of starting material, however, the material prestrained under α = infinity shows lower yield strength than starting material under test condition of α = 0. It was observed that the texture change had an important influence on subsequent yield behavior. Typically, the material highly prestrained under α = 0.5, which had concentrated basal poles, gave the yield locus characterized by remarkable ''texture hardening''. (author)

Rethinking the Zircaloy Embrittlement Criteria and Its Impact on Safety Margin

Energy Technology Data Exchange (ETDEWEB)

Lee, You Ho; Kim, Bo Kyung; No, Hee Cheon [KAIST, Daejeon (Korea, Republic of)

2016-05-15

These fuel rod failure modes include integral thermal shock fracture, and impact tests. It is quite remarkable to see that the proposed Zircaloy embrittlemt criteria attained from ring compression tests, in general, successfully assure structural integrity of fuel rods subject to relevant failure modes in accidents. This fact demonstrates that ductility of Zircaloy is the key metric to structural integrity of fuel rods. However, the Zircaloy embrittlement criteria set in 1970s inevitably pose limitations that have become increasingly important for today's nuclear fuel and reactor operations. In particular, the criteria do not take into account the steady-state hydrogen embrittlement with burnup. This may be understandable considering the markedly lower discharge burnup in 1970s compared to that of today. The revision of the rule has been already conducted by the U.S NRC to account for high burnup effects on ECR while the temperature limit remains unchanged. The newly proposed rule of the U.S NRC stick to the similar ring compression tests conducted in the early 1970s. In the monumental experimental investigation of Hobson and Rittenhouse in 1972 and 1973, the experimental evidence for the current 1204oC was first addressed. The study found a reasonably accurate correlation between zero ductility temperature and the sum of alpha and oxide layer thickness for the specimens oxidized below 2200oF (1204 .deg. C). However, in spite of the similar oxidation degree, specimens oxidized at 2400 .deg. F (1315 deg. C) were markedly more brittle than specimens oxidized at 2200 .deg. F (1204 .deg. C). The study explained this by the increase in solid-solution hardening due to a higher oxygen solubility at a higher temperature. Such a nice experimental correlation attained between the nil ductility temperature and the remaining beta layer thickness fraction below 1204 .deg. C has become a critical basis for the current temperature limit; at 1315 .deg. C- thecorrelation

State Environmental Policy Act (SEPA) environmental checklist forms for 304 Concretion Facility Closure Plan. Revision 2

Energy Technology Data Exchange (ETDEWEB)

1993-11-01

The 300 Area of the Hanford Site contains reactor fuel manufacturing facilities and several research and development laboratories. Recyclable scrap uranium with zircaloy-2 and copper silicon alloy, uranium-titanium alloy, beryllium/zircaloy-2 alloy, and zircaloy-2 chips and fines were secured in concrete billets (7.5-gallon containers) in the 304 Facility, located in the 300 Area. The beryllium/zircaloy-2 alloy and zircaloy-2 chips and fines are designated as mixed waste with the characteristic of ignitability. The concretion process reduced the ignitability of the fines and chips for safe storage and shipment. This process has been discontinued and the 304 Facility is now undergoing closure as defined in the Resource Conservation and Recovery Act (RCRA) of 1976 and the Washington Administrative Code (WAC) Dangerous Waste Regulations, WAC 173-303-040. This closure plan presents a description of the 304 Facility, the history of materials and waste managed, and the procedures that will be followed to close the 304 Facility. The 304 Facility is located within the 300-FF-3 (source) and 300-FF-5 (groundwater) operable units, as designated in the Hanford Federal Facility Agreement and Consent Order (Tri-Party Agreement) (Ecology et al. 1992). Contamination in the operable units 300-FF-3 and 300-FF-5 is scheduled to be addressed through the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA) of 1980 remedial action process. Therefore, all soil remedial action at the 304 Facility will be conducted as part of the CERCLA remedial action of operable units 300-FF-3 and 300-FF-5.

Temperature estimates from the zircaloy oxidation kinetics in the α plus β phase region

International Nuclear Information System (INIS)

Olsen, C.S.

1981-01-01

Oxidation rates of zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near thermocouple locations

Temperature estimates from the Zircaloy oxidation kinetics in the α plus β phase region

International Nuclear Information System (INIS)

Olsen, C.S.

1981-01-01

Oxidation rates of Zircaloy in steam were measured at temperatures between 961 and 1264 K and for duration times between 25 and 1900 seconds in order to calculate, in conjunction with measurements from postirradiation metallographic examination, the prior peak temperatures of Zircaloy fuel rod cladding. These temperature estimates will be used in light water reactor research programs to assess (a) the accuracy of temperature measurements of fuel rod cladding peak temperatures from thermocouples attached to the surface during loss-of-coolant experiments (LOCEs), (b) the perturbation of the fuel rod cladding LOCE temperature history caused by the presence of thermocouples, and (c) the measurements of cladding azimuthal temperature gradients near the thermocouple locations

Thermal-Hydraulic Aspects of Changing the Nuclear Fuel-Cladding Materials from Zircaloy to Silicon Carbides

International Nuclear Information System (INIS)

Niceno, Bojan; Pouchon, Manuel

2014-01-01

The accident in Fukushima has drastically shown the drawbacks of Zircaloy claddings despite their beneficial properties in normal use. The effect of the lack of cooling and the production of hydrogen would not have been so strong if the fuel cladding had not consisted of a zirconium (or metal) alloy. International activities have been started to search for an alternative to Zircaloy, however, still on a limited basis. A project sponsored by Swissnuclear has been conducted at Paul Scherrer Institute (PSI) with the aim to close the gap in knowledge on application of silicon carbides (SiC) as potential replacement for Zircaloys as material for nuclear fuel cladding. The work was interdisciplinary, result of collaboration between different laboratories at PSI, and has focused on SiC cladding material properties, implication of its usage on neutronics and on thermal-hydraulics. This paper summarizes thermal-hydraulic aspects of changing Zircaloy for SiC as the cladding material. The change of cladding material inevitably changes the surface properties thus making a significant impact on boiling curve, and critical heat flux (CHF). Low chemical reactivity of SiC means fewer particles in the flow (less crud), which leads to fewer failures, but also decreases the CHF. Due to differences in physical properties between SiC and Zircaloys, higher brittleness of SiC in particular, might have impact on fuel-rod assembly design, which has direct influence on flow patterns and heat transfer in the fuel assembly. Higher melting (i.e. decomposition) point for SiC means that severe accident management guidelines (SAMG) should have to be re-assessed. Not only would the core degrade later than in the case of conventional fuels, but the production of hydrogen would be quite different as well. All these issues are explored in this work in two steps; first the SiC properties which may have influence on thermal-hydraulics are outlined, then each thermal-hydraulic issues is explained from

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

International Nuclear Information System (INIS)

Gomez Sanchez, Andrea V.

2006-01-01

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

For the world's best cladding tubes, ten years of progress by Zircaloy Special Committee of JAPCO

International Nuclear Information System (INIS)

Mishima, Yoshitsugu

1982-01-01

The zircaloy special committee was organized in 1971 for the purpose of planning the trial use of two nuclear fuel assemblies for which Japan-made cladding tubes were to be used, for a BWR. Now, seven years later, these two fuel assemblies have completed their service life, and have been submitted to post-irradiation examination after cooling for a year. Zircaloy tubes have been produced by Sumitomo Metal Industries, Ltd., and Kobe Steel, Ltd., and more than ten years have elapsed since wholly Japan-made zircaloy cladding tubes were used for reloading fuel elements for the Japan Power Demonstration Reactor. In this report, the history, progress and significance of the works performed by the committee are summarized. The LWR fuel elements made in Japan have attained the highest performance in the world as the leak has been scarce, and the works of the committee is one of the pioneering activities in the development of LWR fuel technology. The situation for starting the committee, the activity of the committee during ten years, the significance and outcome of the committee activity are reported. (Kako, I.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-03-15

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

Mechanistic considerations used in the development of the probability of failure in transient increases in power (PROFIT) pellet-zircaloy cladding (thermo-mechanical-chemical) interactions (pci) fuel failure model

International Nuclear Information System (INIS)

Pankaskie, P.J.

1980-05-01

A fuel Pellet-Zircaloy Cladding (thermo-mechanical-chemical) interactions (PCI) failure model for estimating the Probability of Failure in Transient Increases in Power (PROFIT) was developed. PROFIT is based on (1) standard statistical methods applied to available PCI fuel failure data and (2) a mechanistic analysis of the environmental and strain-rate-dependent stress versus strain characteristics of Zircaloy cladding. The statistical analysis of fuel failures attributable to PCI suggested that parameters in addition to power, transient increase in power, and burnup are needed to define PCI fuel failures in terms of probability estimates with known confidence limits. The PROFIT model, therefore, introduces an environmental and strain-rate dependent Strain Energy Absorption to Failure (SEAF) concept to account for the stress versus strain anomalies attributable to interstitial-dislocation interaction effects in the Zircaloy cladding

TEX-2 - experimental reports 1996

International Nuclear Information System (INIS)

Brokmeier, H.-G.; Witassek, B.

1997-01-01

In 1996 the FRG-1 has been operated during 262 days. Hence, numerous investigations on about 26 different projects were carried out at TEX-2. This report gives a short description of these projects. Beam time was given at about 50% for geological projects and about 50% for materials science projects. The different samples measured at TEX-2 represent a broad spectrum: Magnetite, hematite, gypsum, calcite, halite, galena, YBaCuO-superconductures, niobium, copper, titanium aluminides, intermetallic NiAl, composites (Fe-Cu, Al-Mg, Al-Nb, Cu-Mg, Ag-Ni) and zircaloy. Instrumental improvements were carried out at the loading device, at special sample holders and at the slit system. (orig.) [de

Zircaloy oxidation and cladding deformation in PWR-specific CORA experiments

International Nuclear Information System (INIS)

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

1991-07-01

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

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

International Nuclear Information System (INIS)

Costa Viana, C.S. da

1980-01-01

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

Irradiation of a CANDU UOsub(2) fuel element with twenty-three machined slits cut through the zircaloy sheath

International Nuclear Information System (INIS)

DaSilva, R.L.

1984-09-01

A CANDU fuel element was purposely defected, exposing a minimum UOsub(2) fuel stack area of 272 mmsup(2), by machining 23 longitudinal slits through the Zircaloy-4 sheathing. The element was then irradiated in the X-2 loop of the NRX reactor for a period of 14.64 effective full power days at a linear heat rating of 48 kW/m to investigate the relationship between fission product release and UOsub(2) oxidation behaviour in an element with minimal fuel-to-gap fission gas trapping. The fission product releases, as measured by on-line gamma-ray spectroscopy, revealed that the noble gases and radioiodines are both released from the UOsub(2) fuel matrix directly to the coolant via simple diffusion kinetics, and that their diffusivities in hyperstoichiometric UOsub(2) are approximately equal. The oxidation of UOsub(2) to the higher states UOsub(2+x), Usub(4)Osub(9) and Usub(3)Osub(8), was accompanied by substantial fuel swelling and sheath deformation preferentially located in the lower powered end of the element. The spalling and erosion behaviour of the fuel pellets was correlated to the rate of fuel oxidation

a Study on the Fretting Fatigue Life of Zircaloy Alloys

Science.gov (United States)

Kwon, Jae-Do; Park, Dae-Kyu; Woo, Seung-Wan; Chai, Young-Suck

Studies on the strength and fatigue life of machines and structures have been conducted in accordance with the development of modern industries. In particular, fine and repetitive cyclic damage occurring in contact regions has been known to have an impact on fretting fatigue fractures. The main component of zircaloy alloy is Zr, and it possesses good mechanical characteristics at high temperatures. This alloy is used in the fuel rod material of nuclear power plants because of its excellent resistance. In this paper, the effect of the fretting damage on the fatigue behavior of the zircaloy alloy is studied. Further, various types of mechanical tests such as tension and plain fatigue tests are performed. Fretting fatigue tests are performed with a flat-flat contact configuration using a bridge-type contact pad and plate-type specimen. Through these experiments, it is found that the fretting fatigue strength decreases by about 80% as compared to the plain fatigue strength. Oblique cracks are observed in the initial stage of the fretting fatigue, in which damaged areas are found. These results can be used as the basic data for the structural integrity evaluation of corrosion-resisting alloys considering the fretting damages.

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

International Nuclear Information System (INIS)

Derep, J.L.

1981-09-01

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

Experimental study on coolability of particulate core-metal debris bed with oxidization, (2). Fragmentation and enhanced heat transfer in zircaloy debris bed

International Nuclear Information System (INIS)

Su, Guanghui; Sugiyama, Ken-ichiro; Aoki, Hiroomi; Kimura, Iichi

2006-01-01

The oxidization and coolability characteristics of the particulate Zircaloy debris bed, which is deposited under the hard debris and through which first vapor penetrates and then water penetrates, are studied in the present paper. In the vapor penetration experiments, it is found that Zircaloy debris particles are effectively broken into small pieces after making thick oxidized layer with deep clacks by rapid oxidization under the condition that vapor with 20 cm/s penetrates for 30 to 70 min at an initial debris bed temperature of 1,030degC. It is also confirmed in the water penetration experiments that the oxidized particle debris bed has potentially of high coolability when water penetrates through the fully oxidized particle bed because of a high capillary force originating from those particles with deep cracks on their surfaces. Based on the present study, a new scenario for the appearance and disappearance of the hot spot in the TMI-2 accident is possible. The particulate core-metal core-metal debris bed is first heated up by rapid oxidization with heat generation when vapor can penetrate through the debris bed with porosities. This corresponds to the appearance of the hot spot. The resultant oxidized particulate debris bed causes a high coolability due to its high capillary force when the water can touch the debris bed at wet condition. This corresponds to the disappearance of the hot spot. (author)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

Chauvy, C.

2004-09-01

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

Programme of CABRI start-up measurements with the Zircaloy loop

International Nuclear Information System (INIS)

Kussmaul, G.; Rongier, P.

1981-06-01

After installation and operational tests of the CABRI Zircaloy loop, a start-up test programme will be carried out to determine the new coupling value between the driver core and the test pin and the reactivity dependent driver core energy release for transients from different power levels and modified injection rates. The purpose of the tests and the test programme itself are described in the report

Post test investigation of the single rod tests ESSI 1-11 on temperature escalation in PWR fuel rod simulators due to the Zircaloy/steam reaction

International Nuclear Information System (INIS)

Hagen, S.; Kapulla, H.; Malauschek, H.; Katanishi, S.

1987-03-01

This KfK-report describes the posttest investigation of the single rod tests ESSI-1 to ESSI-11. The objective of these tests was to investigate the temperature escalation behaviour of Zircaloy clad PWR-fuel rods in steam. The investigation of the temperature escalation is part of the program of out-of-pile experiments (CORA) performed within the frame work of the PNS Severe Fuel Damage Program. The experimental arrangement consisted of fuel rod simulator (central tungsten heater, UO 2 ring pellets and Zircaloy cladding), Zircaloy shroud and fiber ceramic insulation. The introductory test ESSI-1 to ESSI-3 were scoping tests designed to obtain information on the temperature escalation of zircaloy in steam. ESSI-4 to ESSI-8 were run with increasing heating rates to investigate the influence of the oxide layer thickness at the start of the escalation. ESSI-9 to ESSI-11 were performed to investigate the influence of the insulation thickness on the escalation behaviour. In these tests we also learned that the gap between removed shroud and insulation has a remarkable influence due to heat removal by convection in the gap. After the test the fuel rod simulator was embedded into epoxy and cut by a diamond saw. The cross sections were photographed and investigated by metalograph microscope, SEM and EMP examinations. (orig./GL) [de

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

International Nuclear Information System (INIS)

Scheuer, A.; Gutsmiedl, E.

1999-01-01

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

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

International Nuclear Information System (INIS)

Gutsmiedl, Erwin

2001-01-01

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

Temperature effect on Zircaloy-4 stress corrosion cracking

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Microstructure in Zircaloy Creep Tested in the R2 Reactor

International Nuclear Information System (INIS)

Pettersson, Kjell

2004-12-01

Tubular specimens of Zircaloy-4 have been creep tested in bending in the R2 reactor in Studsvik. The creep deformation in the reactor core is accelerated in comparison with creep deformation outside the reactor core. The possible mechanisms behind this behaviour are described briefly. In order to determine which the actual mechanism is, the microstructure of the material creep tested in the R2 reactor has been examined by transmission electron microscopy. Due to the bending, material subjected to both tensile and compressive stress during creep was available. Since some of the proposed mechanisms might give microstructures which are different when the material is subjected to compressive or tensile stress it was assumed that examination of both types of material would give valuable information with regard to the operating mechanism. The result of the examination was that in the as-irradiated condition there were no obvious differences detected between materials which had been deformed in tension or compression. After a heat treatment to coarsen the irradiation induced microstructure there were still no significant differences between the two types of material. However it was now observed that in addition to dislocation loops the microstructure also contained network dislocations which presumably had been invisible in the electron microscope before heat treatment due to the high density of small dislocation loops in this state. It is therefore concluded that the most probable mechanism for irradiation creep in this case is climb and glide of the network dislocations. The role of irradiation is two-fold: It accelerates climb due to the production of point defects of which more interstitials than vacancies arrive to the network dislocations stopped at an obstacles. This leads to a net climb after which a dislocation is released from the obstacle and an amount of glide takes place. The second effect is the production of loops which serve as an increasing density of

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Some observations on pitting corrosion in the zircaloy cladding of fuel pins irradiated in a PWR loop

International Nuclear Information System (INIS)

Linde, A. van der; Letsch, A.C.; Hornsveld, E.M.

1978-11-01

A three-pins, zircaloy-4 clad, sphere-pac bundle was irradiated in a 280 0 C PWR loop in the HFR at Petten during 131 effective full power days to a bundle average burnup of 0.84 % FIMA. The pins contained a mixture of 61.5 w/o of 1050 μm (U,Pu) 0 2 spheres, 18.5 w/o of 115 μm UO 2 spheres and 20.0 w/o of 2 spheres. The as-fabricated smear density of the vibratory compacted mixture was 81-85 % T.D. The pressure of the pin filling gas was 1 bar helium for pin 306 and 25 bar helium for the pins 308 and 309. The cladding was zircaloy-4 tubing, stress relieved for 4 hours at 540 0 C, with an inner diameter of 9.30 mm and a wall thickness of 0.73 mm. Exposure of the pins in the loop started in the as-pickled, degreased surface condition. The pins operated at an average heat rating of 335 W/cm and at a peak rating of 620 W/cm. The end-of -life peak rating was 425 W/cm. Unfavourable water chemistry conditions of the coolant during the last weeks of the irradition, in particular low NH 3 concentrations resulting in low pH values, caused the deposition of heavy crud layers on the pin surfaces. This crud layer caused a small cladding defect in pin 306 at the axial position of the peak heat rating. The zircaloy-4 wall failed by complete oxidation, which started at and progressed from the outer, coolant side, surface. Immediately after the detection of fission product activity in the loop water, the irradiation of the bundle was terminated. Microscopic investigations on cross sections of the pins 306 and 309 revealed the presence of oxide pits at the outer surface of the zircalloy-4 wall

Solid-phase zirconium and fluoride species in alkaline zircaloy cladding waste at Hanford.

Science.gov (United States)

Reynolds, Jacob G; Huber, Heinz J; Cooke, Gary A; Pestovich, John A

2014-08-15

The United States Department of Energy Hanford Site, near Richland, Washington, USA, processed plutonium between 1944 and 1987. Fifty-six million gallons of waste of various origins remain, including waste from removing zircaloy fuel cladding using the so-called Zirflex process. The speciation of zirconium and fluoride in this waste is important because of the corrosivity and reactivity of fluoride as well as the (potentially) high density of Zr-phases. This study evaluates the solid-phase speciation of zirconium and fluoride using X-ray diffraction (XRD) and scanning electron microscopy with energy dispersive spectroscopy (SEM-EDS). Two waste samples were analyzed: one waste sample that is relatively pure zirconium cladding waste from tank 241-AW-105 and another that is a blend of zirconium cladding wastes and other high-level wastes from tank 241-C-104. Villiaumite (NaF) was found to be the dominant fluoride species in the cladding waste and natrophosphate (Na7F[PO4]2 · 19H2O) was the dominant species in the blended waste. Most zirconium was present as a sub-micron amorphous Na-Zr-O phase in the cladding waste and a Na-Al-Zr-O phase in the blended waste. Some zirconium was present in both tanks as either rounded or elongated crystalline needles of Na-bearing ZrO2 that are up to 200 μm in length. These results provide waste process planners the speciation data needed to develop disposal processes for this waste. Copyright © 2014 Elsevier B.V. All rights reserved.

Threshold values characterizing iodine-induced SCC of zircaloys

International Nuclear Information System (INIS)

Une, K.

1984-01-01

Threshold values of stress, stress intensity factor, strain, strain rate and iodine concentration for SCC of unirradiated and irradiated Zircaloys are reviewed. The ratio of σsub(th)/σsub(y) adequately represents the effects of cold work and irradiation damage on the SCC susceptibility, where threshold stress σsub(th) is defined as the minimum stress to cause SCC to failure after -6 and 10 -3 min -1 . A comparison of SCC data between constant strain rate and constant stress tests is presented in order to examine the validity of a cumulative-damage concept under SCC conditions. (author)

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

Directory of Open Access Journals (Sweden)

Viplav Duth Shukla

2011-01-01

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

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

International Nuclear Information System (INIS)

Costa Viana, C.S. da

1980-01-01

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

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

International Nuclear Information System (INIS)

Santos, A.M.M. dos.

1981-07-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Treatment of stainless steels and zircaloy cladding hulls

International Nuclear Information System (INIS)

Jenkins, I.L.; Taylor, R.F.

1978-01-01

Results are reported on the fissile material content and the distribution of alpha and beta-gamma emitters in both types of cladding. Apart from very small amounts of residual fuel, fissile material is present as a deposit formed during the dissolution of fuel and also as material driven into the cladding by fission recoil. Alpha-emitters penetrate to depths of 1-2 μm into both S.S. and Zircaloy claddings. The surface deposits on individual hulls can be effectively removed by refluxing with nitric acid or by cleaning with nitric acid in an ultrasonic bath. The physical structural and handling behavior of hull assemblies are examined as being of key importance to the establishment of an efficient cleaning process. The reference leaching target is to extract residual fuel fragments and to remove surface deposits. Preferred routes for compaction, drumming, and encapsulation are briefly reviewed with regard to achieving a final package volume half that of the original hulls with associated hardware

Structural and corrosive properties of ZrO2 thin films on zircaloy-4 by RF reactive magnetron sputtering

International Nuclear Information System (INIS)

Kim, Soo Ho; Lee, Kwang Hoon; Ko, Jae Hwan; Yoon, Young Soo; Baek, Jong Hyuk; Lee, Sang Jin

2006-01-01

Zirconium-oxide (ZrO 2 ) thin films as protective layers were grown on a Zircaloy-4 (Z-4) cladding material as a substrate by RF reactive magnetron sputtering at room temperature. To investigate the effect of plasma immersion on the structural and the corrosive properties of the as-grown ZrO 2 thin film, we immersed Z-4 in plasma during the deposition process. X-ray diffraction (XRD) measurements showed that the as-grown ZrO 2 thin films immersed in plasma had cubic, well as monoclinic and tetragonal, phases whereas those immersed in the plasma had monoclinic and tetragonal phases only. Atomic force microscopy (AFM) measurements of the surface morphology showed that the surface roughness of the as-grown ZrO 2 thin films immersed in plasma was larger than that of the films not immersed in plasma. In addition, the corrosive property of the as-grown ZrO 2 thin films immersed in the plasma was characterized using the weight gains of Z-4 after the corrosion test. Compared with the non-immersed films, the weight gains of the immersed films were larger. These results indicate that the ZrO 2 films immersed in plasma cannot protect Z-4 from corrosive phenomena.

Status of Zircaloy deformation and oxidation research at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Chapman, R.H.; Cathcart, J.V.; Hobson, D.O.

1976-01-01

The U.S. Nuclear Regulatory Commission sponsors a broad range of research on the response of nuclear fuel assemblies to normal, off-normal, and accident conditions in light-water reactors. The paper reviews the current status of three Zircaloy cladding research programs in progress at the Oak Ridge National Laboratory and presents some preliminary results from each

Propagation of stress-corrosion cracks in unirradiated zircaloy

International Nuclear Information System (INIS)

Norring, K.; Haag, Y.; Wikstroem, C.

1982-01-01

Propagation of iodine-induced stress-corrosion cracks in Zircaloy was studied using pre-cracked and internally pressurized cladding tubes. These were recrystallized at different temperatures, to obtain grain sizes between 4 μm and 10 μm. No statistically significant difference in propagation rate due to the difference in grain size was observed. If the obtained data, with Ksub(I) values ranging from 4 to 11 MNmsup(-3/2), were log-log plotted (da/dt = CKsub(I)sup(N)), as usual, they fell within the scatter-band of data reported earlier. But from this plot it could also be seen that the Ksub(I) interval can be divided into two separate parts having different da/dt-Ksub(I) relations. The transition takes place at a Ksub(I) value of about 8 MNmsup(-3/2). The region with lower Ksub(I) values shows a substantially lower n value than the upper region (2.4 and 9.8 respectively), and earlier reported values (n = 7 to 10). This transition is in good agreement with a transition from an intergranular to a transgranular propagation mode of the stress-corrosion crack. (orig.)

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

Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

Energy Technology Data Exchange (ETDEWEB)

Chollet, Mélanie, E-mail: melanie.chollet@psi.ch [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland); Valance, Stéphane; Abolhassani, Sousan; Stein, Gene [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland); Grolimund, Daniel [Paul Scherrer Institute, SLS, 5232 Villigen (Switzerland); Martin, Matthias; Bertsch, Johannes [Paul Scherrer Institute, NES, 5232 Villigen (Switzerland)

2017-05-15

For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO{sub 2} are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components. - Highlights: •A Zircaloy-2 cladding irradiated 9 cycles was investigated thanks to synchrotron X-ray diffraction. •Microstructure and uniform strain through the oxide layer is revealed. •The m-ZrO{sub 2} uniform strain is oriented presenting compression along the (−111) plane. •Virtual tensor is built based on reflecting planes of families of grains. •Tensor components vary from tensile to compressive along the oxide layer.

Synchrotron X-ray diffraction investigations on strains in the oxide layer of an irradiated Zircaloy fuel cladding

International Nuclear Information System (INIS)

Chollet, Mélanie; Valance, Stéphane; Abolhassani, Sousan; Stein, Gene; Grolimund, Daniel; Martin, Matthias; Bertsch, Johannes

2017-01-01

For the first time the microstructure of the oxide layer of a Zircaloy-2 cladding after 9 cycles of irradiation in a boiling water reactor has been analyzed with synchrotron micro-X-ray diffraction. Crystallographic strains of the monoclinic and to some extent of the tetragonal ZrO 2 are depicted through the thick oxide layer. Thin layers of sub-oxide at the oxide-metal interface as found for autoclave-tested samples and described in the literature, have not been observed in this material maybe resulting from irradiation damage. Shifts of selected diffraction peaks of the monoclinic oxide show that the uniform strain produced during oxidation is orientated in the lattice and displays variations along the oxide layer. Diffraction peaks and their shifts from families of diffracting planes could be translated into a virtual tensor. This virtual tensor exhibits changes through the oxide layer passing by tensile or compressive components. - Highlights: •A Zircaloy-2 cladding irradiated 9 cycles was investigated thanks to synchrotron X-ray diffraction. •Microstructure and uniform strain through the oxide layer is revealed. •The m-ZrO 2 uniform strain is oriented presenting compression along the (−111) plane. •Virtual tensor is built based on reflecting planes of families of grains. •Tensor components vary from tensile to compressive along the oxide layer.

Release of indigenous gases from LWR fuel and the reaction kinetics with Zircaloy cladding

International Nuclear Information System (INIS)

Beyer, C.E.; Hann, C.R.

1977-04-01

The objective of this study was to evaluate the open literature data to estimate: the rate of gaseous impurity release from oxide fuel, the amount and composition of the gaseous impurities, and their subsequent rate of reaction with the fuel or Zircaloy

NORA-2, a model for creep deformation and rupture of zircaloy at high temperatures

International Nuclear Information System (INIS)

Raff, S.; Meyder, R.

1983-01-01

A model has been developed to describe Zircaloy cladding behaviour under LOCA and small leak conditions within specified temperature range and strain rates. The deformation model consists of a strain rate equation with two components representing strain rate controlled contributions from different deformation mechanisms. Transition from one mechanism to the other produces the strain rate dependence of the stress exponent of steady state creep. During transient creep the change of creep mechanisms produces a flow softening behaviour which induces unstable creep. Together with a strain hardening model, the strain history can be described for low and high strain values. The influence of oxidation is taken into account by modelling hardening due to solid solution of oxygen, cracking of the brittle oxide and oxygen stabilised α-phase layers, and by an oxidation-induced creep component in steam atmosphere. The rupture criterion is based on a strain fraction rule whose variables are temperature, strain rate or applied stress, and oxygen content. (author)

Contribuciones al conocimiento de las Magnoliácea de Colombia, V

Directory of Open Access Journals (Sweden)

Lozano Contreras Gustavo

1978-12-01

Full Text Available La consecución de material adicional de Magnoliaceae del sur de Colombia permite establecer dos nuevos taxa, ubicados en los géneros Dugandiodendron y Talauma, respectivamente.Al igual que Talauma dixonii Little y Dugandiodendron striatifolium (Little G. Lozano-c., conocidas de Esmeraldas, Ecuador, las nuevas especies reciben el nombre vernáculo de "Cucharillo". Como ocurre con otras magnoliáceas, la madera de estas especies es muy apreciada, por lo cual se les explota a pesar de la prohibición vigente en cuanto a tala; más aún, las especies que se describen están en peligro de extincion yes difícil obtener material completo de las mismas.  Los "cucharillos" han sido elementos importantes en la composición de la selva pluvial macrotérmica de la región tropical occidental ubicada en el Departamento de Nariño.

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

International Nuclear Information System (INIS)

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

2008-01-01

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

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

Science.gov (United States)

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

2008-12-01

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

Static strain aging of Zircaloy-2: the effect of dislocation dynamics on yielding behaviour

International Nuclear Information System (INIS)

Thorpe, W.R.; Smith, I.O.

1981-01-01

The static strain-aging response of Zircaloy-2 was determined in the temperature range 293-723 K. A modified Hahn yielding model was found to provide a satisfactory description of the magnitude and shape of the yield points after aging, thereby providing information about the mobile dislocation density and the dislocation generation rate. For example, the characteristic double peak in the temperature dependence of strain aging was simplified to a single broad minimum in the mobile dislocation density over the temperature interval 500-700 K. The shape of the yield point was also found to be temperature dependent; the yield drop became less sharp at test temperatures above 648 K. This was ascribed to the inhibition of dislocation multiplication by dynamic strain aging. A kinetic law was developed by applying Snoek ordering kinetics to the process of dislocation locking and the resultant change in mobile dislocation density was then used to predict the strain-aging response as a function of aging time. The stress dependence of strain aging at 573 K was investigated at aging stresses of between 0.07 and 0.975 of the flow stress sigmasub(f). The strain-aging response increased for aging at stresses between 0.07sigmassub(f) and 0.8sigmasub(f), whereafter it declined steeply to the limit of zero at the flow stress. (Auth.)

Quenching behaviour of hot zircaloy tube

International Nuclear Information System (INIS)

Chinchole, A.S.; Kulkarni, P.P.; Nayak, A.K.; Vijayan, P.K.

2015-01-01

The quenching process plays a very important role in case of safety of nuclear reactors. During large break Loss of Coolant Accident in a nuclear reactor, the cooling water from the system is lost. Under this condition, cold water is injected from emergency core cooling system. Quenching behaviour of such heated rod bundle is really complex. It is well known that nanofluids have better heat removal capability and high heat transfer coefficient owing to enhanced thermal properties. Alumina nano-particles result in better cooling abilities compared with the traditionally used quenching media. In this paper, the authors have carried out experiments on quenching behaviour of hot zircaloy tube with demineralized water and nanofluids. It was observed that, the tube got quenched within few seconds even with the presence of decay heat and shows slightly reduced quenching time compared with DM water. (author)

Luis Agustín García Moreno, Esther Sánchez Medina (eds.), Del Nilo al Guadalquivir. II Estudios sobre las fuentes de la conqui

OpenAIRE

Meouak, Mohamed

2015-01-01

El libro que el lector tiene entre sus manos constituye una seria apuesta por sacar definitivamente de la obscuridad los estudios dedicados al Mediterráneo centro-occidental y el norte de África entre el final de la Antigüedad y la alta Edad Media dentro del ámbito académico español. Entre mundos distintos y culturas en contacto, dicho volumen ofrece contribuciones científicas de calidad, basadas en investigaciones relativas al norte de África que permiten aumentar y matizar considerablemente...

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

International Nuclear Information System (INIS)

Costa Viana, C.S. da

1982-01-01

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

Fabrication characteristics of zircaloy tubes for nuclear reactors

International Nuclear Information System (INIS)

Haydt, H.M.

1980-11-01

The production sequence for zircaloy cladding tubes to be used in nuclear reactors is described, with emphasis on the texture after reduction and on the variation in the hydrides orientation. The qualities requested for the cladding tubes are presented and reference is made to the quality control applied in the process. The destructive tests as well as the final inspection to which those tubes are subjected are related. A Fabrication Quality Project is requested from the manufacturers by reason of what Quality Control Plans are submitted to be clients. At last an evaluation of the quality to be obtained and of the control performed is mentioned. (Author) [pt

Analysis of atomic distribution in as-fabricated Zircaloy-2 claddings by atom probe tomography under high-energy pulsed laser

Energy Technology Data Exchange (ETDEWEB)

Sawabe, T., E-mail: sawabe@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Sonoda, T.; Kitajima, S. [Central Research Institute of Electric Power Industry (CRIEPI), Iwado Kita 2-11-1, Komae, Tokyo 201-8511 (Japan); Kameyama, T. [Tokai University, Department of Nuclear Engineering, Kitakaname 4-1-1, Hiratsuka, Kanagawa 259-1292 (Japan)

2013-11-15

The properties of second-phase particles (SPPs) in Zircaloy-2 claddings are key factors influencing the corrosion resistance of the alloy. The chemical compositions of Zr (Fe, Cr){sub 2} and Zr{sub 2}(Fe, Ni) SPPs were investigated by means of pulsed laser atom probe tomography. In order to prevent specimen fracture and to analyse wide regions of the specimen, the pulsed laser energy was increased to 2.0 nJ. This gave a high yield of average of 3 × 10{sup 7} ions per specimen. The Zr (Fe, Cr){sub 2} SPPs contained small amounts of Ni and Si atoms, while in Zr{sub 2}(Fe, Ni) SPPs almost all the Si was concentrated and the ratio of Zr: (Fe + Ni + Si) was 2:1. Atomic concentrations of the Zr-matrix and the SPPs were identified by two approaches: the first by using all the visible peaks of the mass spectrum and the second using the representative peaks with the natural abundance of the corresponding atoms. It was found that the change in the concentration between the Zr-matrix and the SPPs can be estimated more accurately by the second method, although Sn concentration in the Zr{sub 2}(Fe, Ni) SPPs is slightly overestimated.

Measurement of dose rate and estimation of beta activity in zircaloy hull drum

International Nuclear Information System (INIS)

Pandey, J.P.N.; Kumar, Pankaj; Shinde, A.M.; Purohit, R.G.; Sarkar, P.K.

2012-01-01

Fuel Reprocessing Plant is designed for the processing of spent fuel from reactor for the recovery of plutonium and uranium as PuO 2 and U 3 O 8 respectively. Zircaloy is used as cladding material of natural uranium fuel pins used in the reactors. In reprocessing plants chop and leach method is used to remove the zircaloy clad from the fuel matrix during Head End Treatment. Initially spent fuel bundles are chopped into pieces and collected in perforated baskets kept in dissolvers. All chopped pieces are dissolved in HNO 3 in the dissolvers followed by heating and boiling. Dissolved solutions are transferred to Filtrate Tank (FT) leaving behind un-dissolved zircoloy hull pieces in the dissolver baskets. Un-dissolved and almost dry hull pieces are transferred in hull drum from the dissolver baskets using the Hull Tilting Facility. Hull drums are made of stainless steel having 500 litre capacity and two third of its volume is filled with zircoloy pieces. Hull drums filled with hull pieces are loaded in Hull Removal Cask (HRC) and transported to SWMF (Solid Waste Management Facility) site for interim storage/disposal in tile holes. Hull pieces are high active solid wastes which contain significant amount of fission products. Radiation levels on hull drums are in the range of few hundreds of mGy/h which has high potential of external hazards if not handled properly. Therefore hull drums are handled remotely in specially designed lead shielded cask

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

Determination of 93Zr, 107Pd and 135Cs in zircaloy hulls analytical development on inactive samples

International Nuclear Information System (INIS)

Excoffier, E.; Bienvenu, Ph.; Combes, C.; Pontremoli, S.; Delteil, N.; Ferrini, R.

2000-01-01

A study involving the participation of three laboratories of the Direction of the Fuel Cycle has been undertaken within the framework of a common interest program existing between the COGEMA and the CEA. Its purpose is to develop analytical methods for the determination of long-lived radionuclides in zircaloy hulls coming from spent fuel reprocessing operations. Acting as a complement to work carried out at the DRRV in ATALANTE concerning zircaloy dissolution and direct analysis of hull solutions, a study is now being conducted at the DESD/SCCD/LARC in Cadarache on three of these radionuclides, namely: zirconium 93, palladium 107 and caesium 135. It concerns three radioisotopes having very long periods (∼10 6 y), and which stabilize mainly through emission of β particles. The analytical technique chosen for the final measurement is inductively coupled plasma mass spectrometry (ICP/MS). Prior to the measurement, chemical separation processes are used to extract the radionuclides from the matrix and separate them from interfering elements and β emitters. The method developed initially on inactive solutions is being validated on irradiated samples coming from UP2/800 - UP3 reprocessing plants. (authors)

The relative axial expansions under irradiation of stacks of UO{sub 2} pellets in zircaloy sheaths

Energy Technology Data Exchange (ETDEWEB)

Notley, M. J.F.

1962-08-15

An experiment was performed to measure the relative axial movement of UO{sub 2} fuel pellets inside a Zircaloy sheath. Although the results must be treated with some reservation, the inferences are that there is very little relative movement between the fuel and the sheath when the two are in firm pressure contact. Relative movement was in the range 0.075 {+-} 0.03 cm for a 30 cm fuel length, and was not greatly affected by the power output, profile of the pellet end-faces or the diametral clearance left between the fuel and the sheath on assembly. However in two elements that had thick sheaths to withstand the coolant pressure and that were assembled with large diametral clearances (2% of the diameter) the available axial clearance for relative fuel/ sheath movement (1%) was fully taken up. The thin sheathed elements showed residual axial expansions of up to 0.17 cm, indicating that the pellets move relative to the sheath only until frictional forces are sufficient for the sheath to grip the fuel; thereafter the sheath is extended. The measurements also indicate that sheath elongation is governed by the temperature at the contact points between adjacent pellets, eg. at the inner edge of a pellet shoulder, as long as that temperature is below approximately 1000{sup o}C. At higher temperatures, the UO{sub 2} is too plastic to exert sufficient force to strain the sheath. (author)

Electronic Structures and Bonding Properties of Ti2AlC and Ti3AlC2

Institute of Scientific and Technical Information of China (English)

MIN Xinmin; REN Yi

2007-01-01

The relation among electronic structure, chemical bond and property of Ti2AlC, Ti3AlC2 and doping Si into Ti2AlC was studied by density function and the discrete variation (DFT-DVM) method. After adding Si into Ti2AlC, the interaction between Si and Ti is weaker than that between Al and Ti, and the strengths of ionic and covalent bonds decrease both. The ionic and covalent bonds in Ti3AlC2, especially in Ti-Al, are stronger than those in Ti2AlC. Therefore, in synthesis of Ti2AlC, the addition of Si enhances the Ti3AlC2 content instead of Ti2AlC. The density of state (DOS) shows that there is mixed conductor characteristic in Ti2AlC and Ti3AlC2. The DOS of Ti3AlC2 is much like that of Ti2AlC. Ti2SixAl1-x C has more obvious tendency to form a semiconductor than Ti2AlC, which is seen from the obvious difference of partial DOS between Si and Al3p.

Algunas contribuciones al debate sobre la clínica de las psicosis

Directory of Open Access Journals (Sweden)

Juan Manuel Rodríguez

Full Text Available La clínica de las psicosis plantea la necesidad de establecer una modalidad de tratamiento diferente al tratamiento de las neurosis. Esta practica implica un trabajo multidisciplinario cercano y frecuente en donde se establecen algunas coordenadas fundamentales en el devenir del tratamiento. Asimismo, en el interior del grupo de trabajo se presentan algunos fenómenos similares a la dinámica propia del paciente psicotico, por ello se hace necesario considerar la locura del grupo como un factor dinámico del tratamiento. En este sentido es fundamental partir de la singularidad del paciente para desarrollar el modelo único en cada caso, en el cual la intervención es un elemento que pretende sostener aquello que falla en la dimensión delirante del paciente.

Delayed hydride cracking of Zircaloy-4 fuel cladding

International Nuclear Information System (INIS)

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

2007-01-01

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

Irradiation-induced growth of zircaloy and its effects on the mechanical design of fuel assemblies

International Nuclear Information System (INIS)

Yao Pu

1991-01-01

Zircaloy growth could be induced due to irradiation. The ammount of growth is described as a function of texture, irradiation temperature, fast neutron fluence and the reduction of cold work, and it should be given great attention in the mechanical design of fuel assemblies

Consideraciones sobre afectividad en las relaciones de enseñanza: las contribuciones de Vigotski

Directory of Open Access Journals (Sweden)

Raimundo Nonato de Oliveira Falabelo

2015-10-01

Full Text Available Este artigo tiene como objetivo presentar conclusiones de un estudio sobre la cuestión afectiva en una perspectiva inter-relacional con la cognición, como una misma unidad, que es la vida psíquica humana, lo que significa concebir que el afecto está presente y es constitutivo de toda y cualquier acción humana. Tiene como base teórica los estudios realizados por L. S. Vigotski. Echamos mano, aún, de las contribuciones de algunos de sus comentadores o interpretadores que tratan específicamente de este tema. El texto hace parte de los fundamentos teóricos de la Tesis de Doctorado sobre esta problemática, en contexto empírico, de clase, en un aula de educación de jóvenes y adultos.

Microstructure in welding zone of a zircaloy 4 tube welded by TIG process

International Nuclear Information System (INIS)

Bolfarini, C.; Domingues Filho, H.

1982-01-01

The details concerned with the welding of seamless zircaloy 4 tubes for nuclear application and the earlier welding tests made in the tubes that will be used for the construction of the Argonautas' Reactor fuel element, are described. Based on the references the microestructure changes in the heat affected zone were analyzed in respect to the material's performance in operation. (Author) [pt

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

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

International Nuclear Information System (INIS)

You, Gil Sung

1992-02-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

Understanding thermally activated plastic deformation behavior of Zircaloy-4

Science.gov (United States)

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

2018-06-01

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

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Characterization of the Microstructure in Recrystallized Zircaloy-2 Cladding Irradiated to a High Neutron Dose

International Nuclear Information System (INIS)

Pettersson, Kjell

2003-04-01

The objectives of the present project were to determine if there is anything in the microstructure of highly irradiated Zircaloy-2 which may make the material fracture in a brittle manner. Samples were taken from three different locations on a fuel rod which had been irradiated for 12 years. The displacement doses were estimated to be 1.4, 9 and 28 dpa. Specimens for electron microscopy were prepared with two different orientation called axial and radial. In the axial orientation the electron beam goes parallel with the basal plane and diffraction conditions can be arranged so that dislocations with a Burgers' vectors become invisible. In the low dose specimen only a-component damage was present and all second phase particles were crystalline. In both the high and intermediate dose samples there was c-component damage present with a slightly higher amount in the high dose sample. The particles of the Zr(Cr,Fe) 2 type were generally amorphous in these samples and the Fe-content of the particles was highly reduced. The hydride structures were similar in all samples. The hydrides were often precipitated in parallel in the same grain and chains of hydrides were seen which ran from grain to grain. No population of small hydrides were observed except from surface hydrides formed during specimen preparation. It was concluded from the investigation that there is nothing in the microstructure which may make the material in the high dose state subject to a purely mechanically induced fast brittle cracking

Coupled growth of Al-Al2Cu eutectics in Al-Cu-Ag alloys

International Nuclear Information System (INIS)

Hecht, U; Witusiewicz, V; Drevermann, A

2012-01-01

Coupled eutectic growth of Al and Al 2 Cu was investigated in univariant Al-Cu-Ag alloys during solidification with planar and cellular morphology. Experiments reveal the dynamic selection of small spacings, below the minimum undercooling spacing and show that distinct morphological features pertain to nearly isotropic or anisotropic Al-Al 2 Cu interfaces.

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

Science.gov (United States)

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

2013-10-01

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

Oxiding and hydriding properties of Zr-1Nb cladding material in comparison with zircaloys

Energy Technology Data Exchange (ETDEWEB)

Vrtilkova, V; Molin, L [Nuclear Fuel Inst., Zbraslav (Czech Republic); Valach, M [Nuclear Research Inst., Rez plc (Czech Republic)

1997-02-01

This paper presents an overview of experimental research related to the Zr-1Nb corrosion behaviour in water and steam environment performed in the Czech Republic. Presented work is focused on the differences between Zr1Nb and Zircaloy corrosion performance. The effects of steam pressure, temperature transients and preoxidation are discussed. (author). 14 refs, 15 figs.

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

International Nuclear Information System (INIS)

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

2012-01-01

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

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

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

International Nuclear Information System (INIS)

Dellaretti Filho, O.

1974-01-01

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

Analysis of zircaloy oxide thickness data from PWRs

International Nuclear Information System (INIS)

Sheppard, K.D.; Speyer, D.M.; Chan, Y.Y.; Frankl, I.; Strasser, A.A.

1990-02-01

Prior EPRI funded research (Project 1250-1) resulted in a set of Zircaloy waterside corrosion models. These models were based principally on KWU reactor data. The objective of this study was to evaluate the ability of the KWU corrosion models to predict available domestic USA data for all domestic PWR vendors in order to further validate the models and to provide a consistent basis to judge the corrosion data of the domestic plants. A methodology for analyzing the large amount of data was developed and implemented in a single channel model. This model includes the capability, by a method described herein, of accounting for open core related effects (crossflow) and the effect of the immediately adjacent fuel rods, guide tubes, etc., on the coolant conditions around the fuel rods that were measured for oxide thickness. Data from the Arkansas Unit number-sign 2 (ANO-2) Combustion Engineering (C-E), Oconee Units 1 and 2 built by Babcock ampersand Wilcox (B ampersand W), and the Trojan reactor built by Westinghouse (W) were used in this study. The corrosion models previously developed, and the present single channel model methodology, were able to predict the corrosion data quite well. The maximum corrosion thickness was on the order of 20 to 40 microns in all plants studied. 13 refs., 11 figs., 5 tabs

PRECIPITATION HARDENING IN B2-ORDERED NiAl BY Ni2AlTiCOMPOUND

Institute of Scientific and Technical Information of China (English)

W.H. Tian; K. Ohishi; M. Nemoto

2001-01-01

Microstructural variations and correlated hardness changes in B2-ordered NiAl containing fine precipitation of Ni2AlTi have been investigated by means of transmission electron microscopy (TEM) and hardness tests. The amount of age hardening is not large as compared to the large microstructural variations during aging. TEM observations have revealed that the L21-type Ni2AlTi precipitates keep a lattice coherency with the NiAl matrix at the beginning of aging. By longer periods of aging Ni2AlTi precipitates lose their coherency and change their morphology to the globular ones surrounded by misfit dislocations. The temperature dependence of the yield strength of precipitate-containing B2-ordered NiAl was investigated by compression tests over the temperature range of 873-1273K. The fine precipitation of Ni2AlTi was found to enhance greatly the yield strength and the high-temperature strength is comparison with that of superalloy Mar-M200.``

Fundamental metallurgical aspects of axial splitting in zircaloy cladding

International Nuclear Information System (INIS)

Chung, H. M.

2000-01-01

Fundamental metallurgical aspects of axial splitting in irradiated Zircaloy cladding have been investigated by microstructural characterization and analytical modeling, with emphasis on application of the results to understand high-burnup fuel failure under RIA situations. Optical microscopy, SEM, and TEM were conducted on BWR and PWR fuel cladding tubes that were irradiated to fluence levels of 3.3 x 10 21 n cm -2 to 5.9 x 10 21 n cm -2 (E > 1 MeV) and tested in hot cell at 292--325 C in Ar. The morphology, distribution, and habit planes of macroscopic and microscopic hydrides in as-irradiated and posttest cladding were determined by stereo-TEM. The type and magnitude of the residual stress produced in association with oxide-layer growth and dense hydride precipitation, and several synergistic factors that strongly influence axial-splitting behavior were analyzed. The results of the microstructural characterization and stress analyses were then correlated with axial-splitting behavior of high-burnup PWR cladding reported for simulated-RIA conditions. The effects of key test procedures and their implications for the interpretation of RIA test results are discussed

Irradiation growth of Zircaloy (LWBR) development program

International Nuclear Information System (INIS)

Williard, H.J.

1984-01-01

Irradiation growth of recrystallized annealed (RXA) Zircaloy is divided into four stages and a model is presented to account for each stage. Stage I is a short time, low-strain transient caused by the accumulation of point defects, small interstitial loops, and vacancy clusters. Stage II is a quasi-steady-state region of relatively low strain rate during which the loops grow and intrinsic dislocations climb. Stage III is a transient during which the strain rate increases due to the production and motion of irradiation-induced dislocation lines. Stage IV is a high-strain-rate, steady-state region during which nonrecoverable strain is caused predominantly by glide of the irradiationinduced dislocations. The proposed model is based on two new mechanisms: (1) direct production of an interstitial dislocation loop accompanied by a vacancy cluster in the primary damage event, and (2) production of dislocations due to the activation of Frank-Read sources by internal stresses caused by interaction of the loops with themselves and with intrinsic (cold work) dislocations. Nonconservative, recoverable strain is due to climb of all dislocations, whereas conservative, nonrecoverable strain is caused by glide of irradiation-induced and intrinsic dislocations under the action of the internal stress. The conservative strain follows a (1-3f) texture dependence

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

International Nuclear Information System (INIS)

Pshenichnikov, Anton; Stuckert, Juri; Walter, Mario

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Superstructure formation in PrNi_2Al_3 and ErPd_2Al_3

International Nuclear Information System (INIS)

Eustermann, Fabian; Hoffmann, Rolf-Dieter; Janka, Oliver; Oldenburg Univ.

2017-01-01

The intermetallic phase ErPd_2Al_3 was obtained by arc-melting of the elements and subsequent annealing for crystal growth. The sample was studied by X-ray diffraction on powders and single crystals. The structure of ErPd_2Al_3 was refined from X-ray diffraction data and revealed a superstructure of PrNi_2Al_3 - a CaCu_5 derivative (P6/m, a=1414.3(1), c=418.87(3) pm wR=0.0820, 1060 F"2 values, 48 variables). The same superstructure was subsequently found for PrNi_2Al_3 (P6/m, a=1407.87(4), c=406.19(2) pm, wR=0.0499, 904 F"2 values, 47 variables). In the crystal structure, the aluminium and transition metal atoms form a polyanionic network according to [T_2Al_3]"δ"-, while rare earth atoms fill cavities within the networks. They are coordinated by six transition metal and twelve aluminum atoms. In contrast to the PrNi_2Al_3 type structure reported so far, two crystallographic independent rare-earth sites are found of which one (1b) is shifted by 1/2 z, causing a distortion in the structure along with a recoloring of the T and Al atoms in the network.

High pressure studies of A2Mo3O12 negative thermal expansion materials (A2=Al2, Fe2, FeAl, AlGa)

International Nuclear Information System (INIS)

Young, Lindsay; Gadient, Jennifer; Gao, Xiaodong; Lind, Cora

2016-01-01

High pressure powder X-ray diffraction studies of several A 2 Mo 3 O 12 materials (A 2 =Al 2 , Fe 2 , FeAl, and AlGa) were conducted up to 6–7 GPa. All materials adopted a monoclinic structure under ambient conditions, and displayed similar phase transition behavior upon compression. The initial isotropic compressibility first became anisotropic, followed by a small but distinct drop in cell volume. These patterns could be described by a distorted variant of the ambient pressure polymorph. At higher pressures, a distinct high pressure phase formed. Indexing results confirmed that all materials adopted the same high pressure phase. All changes were reversible on decompression, although some hysteresis was observed. The similarity of the high pressure cells to previously reported Ga 2 Mo 3 O 12 suggested that this material undergoes the same sequence of transitions as all materials investigated in this paper. It was found that the transition pressures for all phase changes increased with decreasing radius of the A-site cations. - Graphical abstract: Overlay of variable pressure X-ray diffraction data of Al 2 Mo 3 O 12 collected in a diamond anvil cell. Both subtle and discontinuous phase transitions are clearly observed. - Highlights: • The high pressure behavior of A 2 Mo 3 O 12 (A=Al, Fe, (AlGa), (AlFe)) was studied. • All compounds undergo the same sequence of pressure-induced phase transitions. • The phase transition pressures correlate with the average size of the A-site cation. • All transitions were reversible with hysteresis. • Previously studied Ga 2 Mo 3 O 12 undergoes the same sequence of transitions.

Modelling of oxidation and hydriding behaviour of Zircaloy-2 pressure tubes in PHWR

International Nuclear Information System (INIS)

Sah, D.N.; Sunil Kumar; Khan, K.B.

2002-01-01

A computer model named DOCTOR (Deuteriding of Coolant Tubes during Operation of Reactor) has been developed for predicting the axial profile of oxide thickness and hydrogen (Deuterium) concentration in PHWR pressure tubes. This model is applicable to single channel or full core analysis. The main source of hydrogen is considered to be oxidation of pressure tube on the i.d. surface by high temperature coolant water. Three stages of oxidation is considered namely, pre- transition, post transition and accelerated. Oxidation rate is considered to be dependent on channel power, axial power/flux distribution, coolant temperature and pre-existing oxide thickness at the location. The kinetics parameters for oxidation model are derived from the actual measurement of oxide thickness on a number of pressure tubes examined in PIE Division. The input data required for the model are: channel power, channel power factor, axial flux distribution, coolant inlet temperature, critical oxide thickness, hydrogen pick up fraction, initial hydrogen in the material and time of operation (efpy). The model calculates the oxide layer thickness on the inside surface of the pressure tube along the length. The amount of hydrogen picked up by the pressure tube is calculated from the oxide thickness using hydrogen pick up fraction determined from the PIE data. The pressure tube length is divided into a number of axial segments for calculation. The temperature and fast neutron flux assumed to be constant in a given segment. The axial temperature profile calculated from the axial power profile in the channel is used for calculating the oxidation rate at various locations in the pressure tube. The model has been validated with PIE data of hydrogen equivalent measurement on a number of irradiated Zircaloy-2 pressure tubes of various PHWRs. The performance of the model in predicting the axial profile of hydrogen in the pressure tubes has been found to be good. (author)

Chemical inhomogeneity populations in various zircaloy claddings and their association with SCC and corrosion resistance

International Nuclear Information System (INIS)

Tasooji, A.; Miller, A.K.; Cheung, T.Y.; Brooks, M.; Santucci, J.

1987-01-01

A technique has been developed that permits detection and characterization of sparsely distributed chemical inhomogeneities in Zircaloy. These inhomogeneities have previously been observed at the origins of iodine stress-corrosion cracks but are not detectable by, for example, simple scanning electron microscopy (SEM) examination. The technique uses radioactive iodine to ''label'' the chemical inhomogeneities, autoradiography to detect their locations, and SEM and energy-dispersive X-ray analysis (EDAX) to further characterize them. Large areas of surface have been surveyed and statistically meaningful populations of chemical inhomogeneities measured for five different lots of Zircaloy cladding. Inner surfaces and cladding cross-sectional surfaces have been studied. There are clear differences in chemical inhomogeneity size distribution and composition between the various claddings. For three of the claddings characterized in this work, the previously measured stress-corrosion cracking (SCC) threshold stresses correlate well (inversely) with the new data on their average chemical inhomogeneity sizes. Of special interest is the fact that the most SCC-resistant cladding contains far fewer iron-bearing inhomogeneities than the other claddings

Theoretical and experimental studies for selective removal of antimony from zircaloy using thiourea grafted polystyrene adsorbent. Contributed Paper MS-01

International Nuclear Information System (INIS)

Arora, Jyotsna S.; Gaikar, Vilas G.

2014-01-01

During the dissolution step in nuclear fuel reprocessing, hulls consisting of essentially zircaloy clad are produced as high active solid waste. For recovery and reuse of zircaloy from this solid waste, 58 Co and 125 Sb which are present as the activation products of cobalt and tin in zircaloy tubes need to be separated. The present work involves selective sorption of antimony on thiourea grafted polymeric adsorbent in the presence of cobalt and zirconium. The effect of pH for the optimum uptake of antimony ions was studied. Since the variation in pH influences the antimony species formed in the solution, density functional theoretical (DFT) studies were performed in order to understand the complexation of the metal species with the grafted adsorbent at the molecular level. The highest occupied molecular orbital (HOMO) of the adsorbent which is located on S atom of loaded thiourea interacts with lowest unoccupied molecular orbital (LUMO) of Sb(V). The grafted adsorbent exhibits higher interaction with antimony species as compared to cobalt and zirconium. The metal-S bond distances are in good agreement with the XRD values for similar systems. Including the effect of solvation model helps in validation of simulation results with experimental adsorption data suggesting the application of thiourea grafted adsorbent for antimony separation. (author)

Creep modeling of textured zircaloy under biaxial stressing

International Nuclear Information System (INIS)

Adams, B.L.; Murty, K.L.

1984-01-01

Anisotropic biaxial creep behavior of textured Zircaloy tubing was modeled using a crystal-plastic uniform strain-rate upper-bound and a uniform stress lower-bound approach. Power-law steady-state creep is considered to occur on each crystallite glide system by fixing the slip rate to be proportional to the resolved shear stress raised to a power. Prismatic, basal, and pyramidal slip modes were considered. The crystallographic texture is characterized using the orientation distribution function determined from a set of three pole-figures. This method is contrasted with a Von-Mises-Hill phenomenological model in comparison with experimental data obtained at 673 deg K. The resulting creep-dissipative loci show the importance of the basal slip mode on creep in heavily cold-worked cladding, whereas prismatic slip is more important for the recrystallized materials. (author)

A new high temperature deformation model for zircaloy clad ballooning under hypothetical LOCA conditions

International Nuclear Information System (INIS)

Brzoska, B.; Cheliotis, G.; Kunick, A.; Senski, G.

1977-01-01

Assuming Zircaloy clad ballooning occurs predominantly by thermal activated secondary creep, generally a power law is applied to describe the creep rate analytically. According to Norton the creep rate is taken as a power function of the cladding hoop stress multiplied by a numerical constant which is determined by the cladding structural properties and a Boltzmann factor including the creep activation energy, the gas constant and the cladding temperature respectively. As is well known, the stress exponent is not a constant value in the total range of LOCA stresses, but increases steadily with stress. This difficulty is avoided by introducing into the Norton law a plastic flow-factor including a limiting stress, which was derived by G. Senski using plastic crack models from Dugdale and Irwin. For LOCA applications the limiting stress is identified with the burst stress, which is experimentally determined. A total number of about 280 directly heated KWU burst tests including two types of experiments: (i) controlled temperature transient tests, (ii) creep rupture tests, are used to fit the burst stress of KWU zircaloy tubes simulating the whole range of LOCA temperatur

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

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

Proposition of a modification to the VAR process and its application in the consolidation of pressed zircaloy chips and the evaluation of the dynamical system of the electric arc; Proposicao de um processo alternativo a fusao via forno VAR para a consolidacao de cavacos prensados de zircaloy e estudo do sistema dinamico do arco eletrico

Energy Technology Data Exchange (ETDEWEB)

Mucsi, Cristiano Stefano

2005-07-01

The objective of this work is the investigation of a new process as an alternative to the Vacuum Arc Remelting technology in the consolidation of Zircaloy chips. A procedure is proposed for the recycling of primary Zircaloy scraps by means of a modified VAR furnace. The performed studies were made in order to optimise the low cost new devices added to existing VAR furnace prototype, find ideal operational conditions, evaluate data acquisition system and the electric arc dynamical system in order to made viable the automated control of the modified VAR prototype. A funnel-crucible special device was developed and installed in a VAR prototype furnace allowing ingots to be obtained from pressed chips. This indicated the viability of creation of a new process for the consolidation of Zircaloy chips. The voltage of the electric arc during the melting runs was digitally recorded allowing the evaluation of the electric arc dynamics by using the topological invariant of the system: correlation dimension and the higher Liapunov exponent. (author)

The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

International Nuclear Information System (INIS)

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

1981-01-01

The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

Tailoring ultrafine grained and dispersion-strengthened Ti 2 AlC/TiAl ...

Indian Academy of Sciences (India)

In situ Ti 2 AlC/TiAl composite was fabricated by hot-pressing method via the reaction system of Ti 3 AlC 2 and Ti-Al pre-alloyed powders at low temperature of 1150 ∘ C. The composite mainly consisted of TiAl, Ti 3 Al and Ti 2 AlC phases. Fine Ti 2 AlC particles were homogeneously distributed and dispersed in the matrix.

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

In situ synthesis of Ti{sub 2}AlC–Al{sub 2}O{sub 3}/TiAl composite by vacuum sintering mechanically alloyed TiAl powder coated with CNTs

Energy Technology Data Exchange (ETDEWEB)

Wang, Jian [Department of Materials Science and Engineering of Tianjin University, Tianjin Key Laboratory of Composite and Functional Materials, Tianjin 300072 (China); Zhao, Naiqin, E-mail: nqzhao@tju.edu.cn [State Key Laboratory of Hydraulic Engineering Simulation and Safety, Tianjin (China); Department of Materials Science and Engineering of Tianjin University, Tianjin Key Laboratory of Composite and Functional Materials, Tianjin 300072 (China); Nash, Philip [Thermal Processing Technology Center, Illinois Institute of Technology, IL (United States); Liu, Enzuo; He, Chunnian; Shi, Chunsheng; Li, Jiajun [Department of Materials Science and Engineering of Tianjin University, Tianjin Key Laboratory of Composite and Functional Materials, Tianjin 300072 (China)

2013-11-25

Highlights: •Using zwitterionic surfactant to enhance the dispersion of the CNTs on the powder surface. •CNTs as carbon source decreased the formation temperature of Ti{sub 2}AlC. •Al{sub 2}O{sub 3} was generated in situ from the oxygen atoms introduced in the drying procedure. •Nanosized Ti{sub 3}Al was precipitated at 1250 °C and distribute in the TiAl matrix homogeneously. •Ti{sub 2}AlC–Al{sub 2}O{sub 3}/TiAl composite was synthesized in situ by sintering pre-alloy Ti–Al coated with CNTs. -- Abstract: Bulk Ti{sub 2}AlC–Al{sub 2}O{sub 3}/TiAl composites were in situ synthesized by vacuum sintering mechanically alloyed Ti–50 at.% Al powders coated with carbon nanotubes (CNTs). The pre-alloyed Ti–50 at.% Al powder was obtained by ball milling Ti and Al powders. The multi-walled carbon nanotubes as the carbon resource were covered on the surface of the pre-alloyed powders by immersing them into a water solution containing the CNTs. A zwitterionic surfactant was used to enhance the dispersion of the CNTs on the powder surface. The samples were cold pressed and sintered in vacuum at temperatures from 950 to 1250 °C, respectively. The results show that the reaction of forming Ti{sub 2}AlC can be achieved below 950 °C, which is 150 °C lower than in the Ti–Al–TiC system and 250 °C lower than for the Ti–Al–C system due to the addition of CNTs. Additionally, the reinforcement of Al{sub 2}O{sub 3} particles was introduced in situ in Ti{sub 2}AlC/TiAl by the drying process and subsequent sintering of the composite powders. Dense Ti{sub 2}AlC–Al{sub 2}O{sub 3}/TiAl composites were obtained by sintering at 1250 °C and exhibited a homogeneous distribution of Ti{sub 2}AlC, Al{sub 2}O{sub 3} and precipitated Ti{sub 3}Al particles and a resulting high hardness.

27Al NMR studies of NpPd5Al2

International Nuclear Information System (INIS)

Chudo, H.; Sakai, H.; Tokunaga, Y.; Kambe, S.; Aoki, D.; Homma, Y.; Shiokawa, Y.; Haga, Y.; Ikeda, S.; Matsuda, T.D.; Onuki, Y.; Yasuoka, H.

2009-01-01

We present 27 Al NMR studies for a single crystal of the Np-based superconductor NpPd 5 Al 2 (T c =4.9K). We have observed a five-line 27 Al NMR spectrum with a center line and four satellite lines separated by first-order nuclear quadrupole splittings. The Knight shift clearly drops below T c . The temperature dependence of the 27 Al nuclear spin-lattice relaxation rate shows no coherence peak below T c , indicating that NpPd 5 Al 2 is an unconventional superconductor with an anisotropic gap. The analysis of the present NMR data provides evidence for strong-coupling d-wave superconductivity in NpPd 5 Al 2 .

Al2O3 adherence on CoCrAl alloys

International Nuclear Information System (INIS)

Kingsley, L.M.

1980-04-01

Adherence of protective oxides on NiCrAl and CoCrAl superalloys has been promoted by a dispersion of a highly oxygen reactive element or its oxide being produced within the protection system. Two aspects of this subject are investigated here: the use of Al 2 O 3 as both the dispersion and protective oxide; and the production of an HfO 2 dispersion while simultaneously aluminizing the alloy. It was found that an Al 2 O 3 dispersion will act to promote the adherence of an external scale of Al 2 O 3 to a degree comparable to previously tested dispersions and an HfO 2 dispersion comparable to that produced by a Rhines pack treatment is produced during aluminization

Hyperfine fields and spin relaxation of Ce in GdAl2 and DyAl2

International Nuclear Information System (INIS)

Waeckelgaard, E.; Karlsson, E.; Lindgren, B.; Mayer, A.

1987-04-01

We have investigated the ferromagnetic state of the cubic intermetallic compounds GdAl 2 and DyAl 2 with the 140 Ce nuclei using DPAC. The local fields of Ce are for the lowest measured temperatures B eff (30 K) = 54(2) T for GdAl 2 and B eff (12.5 K) = 27(1) T for DyAl 2 which are considerably lower than the hyperfine field measured for a free Ce ion (183 T). By introducing a crystal field, of cubic symmetry, a lower hyperfine field is obtained which is in quantitative agreement with the local field of Ce in GdAl 2 . For DyAl 2 an additional effect, represented by a non-magnetic level below the lowest magnetic level, may explain a further reduction of the hyperfine field. Two models relating to a Kondo non-magnetic state of Ce are discussed. Spin relaxation in the paramagnetic state are also studied and compared with observations of the same systems measured with μSR. (authors)

On-line ultrasonic inside-diameter control system for Zircaloy

International Nuclear Information System (INIS)

Tanaka, Y.; Fujii, N.; Komatsu, M.; Kubota, H.

1984-01-01

An ultrasonic inside-diameter (ID) control system was used during the final etching process for producing Zircaloy nuclear fuel cladding tubes. This results in establishing automatic inside-diameter control during etching with an automatic etching system. In this system, the inside-diameter at the center point in the length of each tube is continuously measured with the ultrasonic inside-diameter measuring equipment during the etching process and the etching is automatically stopped by a signal from the control equipment when the inside-diameter reaches the target value. This made the final etching process economical and suitable for large-scale production, having an equal or better level at the inside-diameter of tubes etched with this system than those made by a process controlled by an air-micrometer

Creep behavior of Zircaloy cladding under variable conditions

International Nuclear Information System (INIS)

Matsuo, Y.

1989-01-01

Various creep tests of Zircaloy cladding tubes under variable conditions were conducted to investigate which hardening rule can be applicable for the creep behavior associated with condition changes. The results show that the strain-hardening rule is applicable in general when either the stress or temperature conditions change, provided that a certain amount of creep strain recovery is observed in case of stress drop. In stress reversal conditions, however, softening of the material was observed. Strain rate after stress reversal is much higher than that predicted by the strain-hardening rule. In this case, the modified strain-hardening model, considering a recoverable creep-hardening range together with the strain recovery, predicts the creep behavior well. The applicability of the model is ascertained through a verification test that includes stress reversal, strain recovery, stress changes, and temperature changes

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

International Nuclear Information System (INIS)

Costa, I.; Ramanathan, L.V.

1986-01-01

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

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

Science.gov (United States)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

La relación entre creatividad y expertise: contribuciones teóricas y empíricas

Directory of Open Access Journals (Sweden)

Afonso C. T. Galvao

2009-06-01

Full Text Available La creatividad y el expertise son temas que están recibiendo mayor atención de los inves­tigadores en distintos campos del conocimiento. Atributos personales asociados a la alta producción creativa, factores que favorecen el desarrollo, la expresión de la creatividad y el expertise, así como los procesos implicados en el desarrollo de la alta competencia y la creatividad, son temas que han sido objeto de numerosos estudios. El presente estudio pre­senta contribuciones teóricas y estudios empíricos acerca de la creatividad y el expertise. Se describen los elementos comunes para el desarrollo de la creatividad y el expertise, así como las características de los ambientes educativos que los promueven, los modelos teóricos y estudios empíricos.

Psicología comunitaria y expresiones psicosociales de la pobreza: contribuciones para la intervención en políticas públicas

OpenAIRE

Morais Ximenes, Verônica; Camurça Cidade, Elívia; Barbosa Nepomuceno, Bárbara

2015-01-01

El objetivo de este trabajo es analizar, desde la psicología comunitaria, las expresiones psicosociales de la pobreza y sus contribuciones para la intervención en políticas públicas. La psicología comunitaria critica los factores perpetuadores de aspectos materiales y simbólicos que interfieren en la constitución subjetiva de los pobres. La investigación exploratoria, de carácter cuantitativo y cualitativo, fue realizada con 417 sujetos adultos de dos comunidades de Brasil, una rural y otra u...

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Engineered zircaloy cladding modifications for improved accident tolerance of LWR fuel: US DOE NEUP Integrated Research Project

International Nuclear Information System (INIS)

Heuser, Brent

2013-01-01

An integrated research project (IRP) to fabricate and evaluate modified zircaloy LWR cladding under normal BWR/PWR operation and off-normal events has been funded by the US DOE. The IRP involves three US academic institutions, a US national laboratory, an intermediate stock industrial cladding supplier, and an international academic institution. A combination of computational and experimental protocols will be employed to design and test modified zircaloy cladding with respect to corrosion and accelerated oxide growth, the former associated with normal operation, the latter associated with steam exposure during loss of coolant accidents (LOCAs) and low-pressure core re-floods. Efforts will be made to go beyond design-base accident (DBA) scenarios (cladding temperature equal to or less than 1204 deg. C) during the experimental phase of modified zircaloy performance characterisation. The project anticipates the use of the facilities at ORNL to achieve steam exposure beyond DBA scenarios. In addition, irradiation of down-selected modified cladding candidates in the ATR may be performed. Cladding performance evaluation will be incorporated into a reactor system modelling effort of fuel performance, neutronics, and thermal hydraulics, thereby providing a holistic approach to accident-tolerant nuclear fuel. The proposed IRP brings together personnel, facilities, and capabilities across a wide range of technical areas relevant to the study of modified nuclear fuel and LWR performance during normal operation and off-normal scenarios. Two pathways towards accident-tolerant LWR fuel are envisioned, both based on the modification of existing zircaloy cladding. The first is the modification of the cladding surface by the application of a coating layer designed to shift the M + O→MO reaction away from oxide growth during steam exposure at elevated temperatures. This pathway is referred to as the 'surface coating' solution. The second is the modification of the bulk

When does ALS start? ADAR2-GluA2 hypothesis for the etiology of sporadic ALS

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Takuto eHideyama

2011-11-01

Full Text Available Amyotrophic lateral sclerosis (ALS is the most common adult-onset motor neuron disease. More than 90% of ALS cases are sporadic, and the majority of sporadic ALS patients do not carry mutations in genes causative of familial ALS; therefore, investigation specifically targeting sporadic ALS is needed to discover the pathogenesis. The motor neurons of sporadic ALS patients express unedited GluA2 mRNA at the Q/R site in a disease-specific and motor neuron-selective manner. GluA2 is a subunit of the AMPA receptor, and it has a regulatory role in the Ca2+-permeability of the AMPA receptor after the genomic Q codon is replaced with the R codon in mRNA by adenosine-inosine conversion, which is mediated by adenosine deaminase acting on RNA 2 (ADAR2. Therefore, ADAR2 activity may not be sufficient to edit all GluA2 mRNA expressed in the motor neurons of ALS patients. To investigate whether deficient ADAR2 activity plays pathogenic roles in sporadic ALS, we generated genetically modified mice (AR2 in which the ADAR2 gene was conditionally knocked out in the motor neurons. AR2 mice showed an ALS-like phenotype with the death of ADAR2-lacking motor neurons. Notably, the motor neurons deficient in ADAR2 survived when they expressed only edited GluA2 in AR2/GluR-BR/R (AR2res mice, in which the endogenous GluA2 alleles were replaced by the GluR-BR allele that encoded edited GluA2. In heterozygous AR2 mice with only one ADAR2 allele, approximately 20% of the spinal motor neurons expressed unedited GluA2 and underwent degeneration, indicating that half-normal ADAR2 activity is not sufficient to edit all GluA2 expressed in motor neurons. It is likely therefore that the expression of unedited GluA2 causes the death of motor neurons in sporadic ALS. We hypothesize that a progressive downregulation of ADAR2 activity plays a critical role in the pathogenesis of sporadic ALS and that the pathological process commences when motor neurons express unedited GluA2.

Influence of specimen design on the ductility of zircaloy cladding: Experiment and analysis

International Nuclear Information System (INIS)

Bates, D. W.; Majumdar, S.; Koss, D. A.; Motta, A. T.

1999-01-01

In a reactivity-initiated accident (RIA), a control rod ejection or drop causes a sudden increase in reactor power, which in turn deposits a large amount of energy into the fuel. The resulting thermal expansion and fission gas release loads the cladding into the plastic regime and may cause it to fail. In order to predict cladding survivability, there has been considerable interest and effort in supplementing integral WA tests with separate-effects ring tests of cladding tubes. Such tests can give one insight into failure mechanisms and measure relevant mechanical properties (such as yield strength, uniform elongation, uniaxial stress-strain curve, etc.), for use in computer codes that attempt to predict cladding response during an RIA. The accuracy of such model predictions obviously depends on appropriate and accurate failure data. This study concerns itself with the proper development of ring tensile tests that (i) are similar to the loading conditions present in an RIA, (ii) measure the relevant mechanical properties and (iii) provide insight regarding the influence of the strain paths on the failure mechanisms present if Zircaloy cladding. Based on both experiments and computational modeling, the authors investigate the failure of Zircaloy tubing as a function of specimen geometry, and discuss the limitations of certain ring-test geometries in yielding failure ductility data that are applicable to RIA situations

Characterization of Zircaloy-4 oxide layers by impedance spectroscopy

International Nuclear Information System (INIS)

Barberis, P.

1999-01-01

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

A fracture mechanics model for iodine stress corrosion crack propagation in Zircaloy tubing

International Nuclear Information System (INIS)

Crescimanno, P.J.; Campbell, W.R.; Goldberg, I.

1984-01-01

A fracture mechanics model is presented for iodine-induced stress corrosion cracking in Zircaloy tubing. The model utilizes a power law to relate crack extension velocity to stress intensity factor, a hyperbolic tangent function for the influence of iodine concentration, and an exponential function for the influence of temperature and material strength. Comparisons of predicted to measured failure times show that predicted times are within a factor of two of the measured times for a majority of the specimens considered

Explosive destruction of "2"6Al

International Nuclear Information System (INIS)

Kahl, D.; Yamaguchi, H.; Shimizu, H.

2016-01-01

The γ-ray emission associated with the radioactive decay of "2"6Al is one of the key pieces of observational evidence indicating stellar nucleosynthesis is an ongoing process in our Galaxy, and it was the first such radioactivity to be detected. Despite numerous efforts in stellar modeling, observation, nuclear theory, and nuclear experiment over the past four decades, the precise sites and origin of Galactic "2"6Al remain elusive. We explore the present experimental knowledge concerning the destruction of "2"6Al in massive stars. The precise stellar rates of neutron-induced reactions on "2"6Al, such as (n,p) and (n,α), have among the largest impacts on the total "2"6Al yield. Meanwhile, reactions involving the short-lived isomeric state of "2"6Al such as radiative proton capture are highly-uncertain at present. Although we presented on-going experimental work from n TOF at CERN with an "2"6Al target, the present proceeding focuses only on the "2"6Al isomeric radioactive beam production aspect and the first experimental results at CRIB.

Proporcionalidad y equidad en las contribuciones. El amparo fiscal en México, 1917-1968

Directory of Open Access Journals (Sweden)

Carlos de Jesús Becerril Hernández

2015-01-01

Full Text Available El juicio de amparo ha sido comúnmente identificado como parte del proceso de la modernización jurídica liberal de la segunda mitad del siglo XIX. No obstante, el amparo fiscal, que tiene que ver directamente con el reclamo por la proporcionalidad y equidad de las contribuciones contenidas en el artículo 31, fracción IV, de la Constitución Política de los Estados Unidos Mexicanos de 1917, no fue reconocido como jurisprudencia inmediatamente aplicable; el criterio del ministro Ignacio L. Vallarta (1830-1893 sobre la incapacidad de la Corte para conocer de este recurso fue vigente hasta bien entrado el siglo XX. El amplio arco temporal (1917-1968 cubierto por este ensayo tratará de explicar, a la luz del derecho constitucional y del desempeño económico, por qué la justicia federal se negó a amparar y proteger a los causantes, bajo el argumento de que la Corte no era juez ni órgano revisor de las leyes expedidas por el Congreso.

JAEA's research on the effects of seawater and radiation on corrosion of Zircaloy and PCV/RPV steels

International Nuclear Information System (INIS)

Tsukada, Takashi; Motooka, Takafumi; Nakano, Junichi

2014-01-01

In order to implement successfully a lot of work for the extraction of fuel assemblies from spent fuel pool (SFP) and also for the removal of fuel debris from reactor pressure vessel (RPV) and primary containment vessel (PCV) at the Fukushima Daiichi Nuclear Power Station (NPS) of Tokyo Electric Power Co., it is necessary to investigate and to prevent the degradation of structural materials of the fuel assemblies and PCV/RPV which are exposed to the gamma radiation and water containing seawater ingredient, because those factors are influencing and possibly accelerating corrosion of the materials. Therefore, at the Japan Atomic Energy Agency (JAEA), we are carrying out the research related to the corrosion issues which may affect the integrity of fuel assemblies and reactor vessels, i.e. PCV and reactor pressure vessel (RPV), from a viewpoint of the effect of gamma radiation and diluted seawater on corrosion behavior as described in this review. In SFP, hydrazine (N_2H_4) was added to salt-containing water in order to reduce dissolved oxygen (DO). Therefore, deoxygenation behavior by N_2H_4 addition was investigated at the ambient temperature. To evaluate the effects of radiolysis on the initiation of pitting corrosion on Zircaloy-2 in water containing sea salt, the pitting potentials of Zircaloy-2 were evaluated. The experimental results showed that the pitting potential under irradiation was slightly higher than that under conditions in which no radiation was present. Corrosion tests of PCV/RPV steels were conducted in diluted seawater at 50degC under gamma-ray irradiation of dose rates of 4.4 and 0.2 kGy/h. To assess the effect of N_2H_4 as an oxygen scavenger under gamma-ray irradiation in PCV condition, 10 and 100 mg/L N_2H_4 were added to the diluted seawater. When gas phase in test flask was replaced with N_2, corrosion weight loss of the steels decreased remarkably. (author)

CuAlO2 and CuAl2O4 thin films obtained by stacking Cu and Al films using physical vapor deposition

Science.gov (United States)

Castillo-Hernández, G.; Mayén-Hernández, S.; Castaño-Tostado, E.; DeMoure-Flores, F.; Campos-González, E.; Martínez-Alonso, C.; Santos-Cruz, J.

2018-06-01

CuAlO2 and CuAl2O4 thin films were synthesized by the deposition of the precursor metals using the physical vapor deposition technique and subsequent annealing. Annealing was carried out for 4-6 h in open and nitrogen atmospheres respectively at temperatures of 900-1000 °C with control of heating and cooling ramps. The band gap measurements ranged from 3.3 to 4.5 eV. Electrical properties were measured using the van der Pauw technique. The preferred orientations of CuAlO2 and CuAl2O4 were found to be along the (1 1 2) and (3 1 1) planes, respectively. The phase percentages were quantified using a Rietveld refinement simulation and the energy dispersive X-ray spectroscopy indicated that the composition is very close to the stoichiometry of CuAlO2 samples and with excess of aluminum and deficiency of copper for CuAl2O4 respectively. High resolution transmission electron microscopy identified the principal planes in CuAlO2 and in CuAl2O4. Higher purities were achieved in nitrogen atmosphere with the control of the cooling ramps.

High-pressure modifications of CaZn2, SrZn2, SrAl2, and BaAl2: Implications for Laves phase structural trends

International Nuclear Information System (INIS)

Kal, Subhadeep; Stoyanov, Emil; Belieres, Jean-Philippe; Groy, Thomas L.; Norrestam, Rolf; Haeussermann, Ulrich

2008-01-01

High-pressure forms of intermetallic compounds with the composition CaZn 2 , SrZn 2 , SrAl 2 , and BaAl 2 were synthesized from CeCu 2 -type precursors (CaZn 2 , SrZn 2 , SrAl 2 ) and Ba 21 Al 40 by multi-anvil techniques and investigated by X-ray powder diffraction (SrAl 2 and BaAl 2 ), X-ray single-crystal diffraction (CaZn 2 ), and electron microscopy (SrZn 2 ). Their structures correspond to that of Laves phases. Whereas the dialuminides crystallize in the cubic MgCu 2 (C15) structure, the dizincides adopt the hexagonal MgZn 2 (C14) structure. This trend is in agreement with the structural relationship displayed by sp bonded Laves phase systems at ambient conditions. - Graphical abstract: CeCu 2 -type polar intermetallics can be transformed to Laves phases upon simultaneous application of pressure and temperature. The observed structures are controlled by the valence electron concentration

Contribuciones al desarrollo de la investigación en enfermería: retos y perspectivas

Directory of Open Access Journals (Sweden)

Myriam Durán Parra

2012-12-01

Full Text Available Desde la etimología del término “investigación” la palabra proviene del latín in (en y vestigare (hallar, inquirir, indagar, seguir vestigios. De ahí el uso más elemental del término en el sentido de “averiguar o describir alguna cosa”. Desde el momento en que el hombre se enfrentó a problemas y frente a ellos empezó a interrogarse sobre el porqué, como y para que, con esta indagación sobre las cosas, de una manera embrionaria, comenzó lo que hoy llamamos investigación. Con el transcurso del tiempo el ser humano ha hecho un esfuerzo importante por sobrevivir y la mejor forma de lograrlo ha sido experimentando aportando al desarrollo del conocimiento científico.Donaldson y Bottorff (1978 expresan frente a la investigación en enfermería que “es la fuente de desarrollo de conocimiento que le da la característica de disciplina, que ha de estar presente tanto en la práctica clínica como en el desarrollo de la administración y la educación, que ha de dar respuesta a las necesidades sociales que están relacionadas con procesos de salud-enfermedad, condiciones y calidad de vida de las personas, familias y comunidades, en todos los grupos etarios” (1. Por su parte Borrero-Cabal (1994, afirma que “la disciplina conlleva al sentido de rigor, de dedicación, de entrenamiento y ejercicio de los hábitos científicos de la persona para elaborar, transmitir y aprender una ciencia” (2. En enfermería esto se traduce que para poder llevar el cuidado a los pacientes de una manera técnica, científica , analítica, debe tener unas bases como la dedicación, el amor por la profesión y el gusto por la investigación reconociendo que la enfermería avanza y se fortalece a medida que crece científicamente, mientras conoce la cultura, la riqueza de la naturaleza, forma parte del país y sobre todo del mundo como una profesión que demuestra su interés en la búsqueda de su propia identidad. 

Development of advanced neutron radiography for inspection on irradiated fuels and materials (2). Observation of hydride and oxide film on zircaloy cladding by using neutron radiography

International Nuclear Information System (INIS)

Yasuda, Ryou; Nakata, Masahito; Mastubayashi, Masahito; Harada, Katsuya

2001-02-01

Neutron radiography has been used as available diagnosis method of integrity on irradiated fuels, and has not been employed for estimating hydride and oxide film, which are influenced on integrity of Zircaloy cladding. Preliminary tests for PIE were carried out to assess possibility of neutron radiography as evaluation tool for hydrided and oxide film on the cladding. In these experiments, Zircaloy claddings with controlled amount of hydrogen absorption (200, 500, and 1000ppm) and thickness of oxide film were radiographed in center axis and in side directions of cladding tube by neutron imaging plate method. It is noted that thickness of oxide film was formed range from 7 μ m to 70 μ m at various temperatures (973, 1173, and 1323K) under steam atmosphere on the Zircaloy claddings. CT (Computed Tomography) restructure calculation was carried out to obtain cross section image of the claddings non-destructively. The Radiographs were qualitatively investigated about structure formation area and dependence of hydrogen absorption amount on PSL (Photo Simulated Luminescence) and CT values using by image analysis processor. At the results of imaging plate test, obvious difference was not found out between hydride formation (except for 1000ppm cladding) and standard claddings in side direction image. However, on the center axis direction image, outer circumference in the cladding cross-section that corresponded with hydride segregation area became blacker. In the case of oxide film formed cladding images, although oxide film could not find out on all speciments in the radiographs taken at the center axis and side directions, cross-section of claddings heat-processed at 973K showed appreciable blackness increasing with oxide film thickness on the radiographs. On the other hand, there is no effective difference between images of oxide film formed claddings processed at 1173K and 1323K and that of standard cladding. In CT image of 1000ppm hydrogen absorbed cladding, it is