Showing papers on "Zirconium alloy published in 1992"
TL;DR: In this paper, the authors used optical, scanning electron, and transmission electron microscopy data to show that the intergranular creep failure in Fe3Al is related to weak high-angle grain boundaries and to formation of subgrain boundary arrays, which reduce the ability of dislocations to glide or multiply to produce matrix plasticity.
Abstract: Recent alloy development efforts have shown that Fe3Al-based alloys can have room temperature tensile ductilities of 10–20% and yield strengths of 500 MPa at temperatures to 600 °C. These property improvements are important for enabling the use of iron-aluminides for structural applications that require their excellent corrosion resistance. New data are presented here from creep-rupture studies on Fe3Al and on Fe3Al-based alloys containing molybdenum or niobium plus zirconium. Binary Fe3Al alloys have low creep resistance, but the addition of 2 at. % Mo or 1% Nb plus 0.1% Zr increases the creep life and reduces the minimum creep rate, with the niobium-containing alloy being the strongest. The improvement in creep life is the result of a combination of factors which include grain boundary strengthening, resistance to dynamic recrystallization during stressing, precipitation strengthening, and changes in the formation and mobility of the dislocation network. Correlation of optical, scanning electron, and transmission electron microscopy data suggests that the intergranular creep failure found in Fe3Al after creep testing at 550–650 °C is related to weak high-angle grain boundaries and to formation of subgrain boundary arrays, which reduce the ability of dislocations to glide or multiply to produce matrix plasticity. The addition of niobium/zirconium results in solid solution strengthening effects, as well as the formation of fine MC precipitates (a small amount of carbon is present as a contaminant from the casting process) which strengthen both the matrix and grain boundaries. The result relative to the binary alloy is increased creep-rupture strength and life coupled with a change to a ductile-dimple transgranular failure mode. This suggests that the mechanisms that cause failure during creep can be controlled by macro- and microalloying effects.
101 citations
TL;DR: In this article, a study by transmission electron microscopy has been made of the incorporation and oxidation of intermetallic precipitates Zr(Cr,Fe)2 into the uniform oxide layer of Zircaloy-4.
100 citations
TL;DR: In this article, an out-of-pile corrosion test was conducted using Zircaloy-4 specimens cut from the channel box of a fuel assembly irradiated in the BWR (Monticello reactor) up to the neutron fluence of 1.53×1026 n/m2 (E>1MeV).
Abstract: In order to investigate irradiation effects on nodular corrosion resistance of Zircaloy-4, an out-of-pile corrosion test was conducted using Zircaloy-4 specimens cut from the channel box of a fuel assembly irradiated in the BWR (Monticello reactor) up to the neutron fluence of 1.53×1026 n/m2 (E>1MeV). The corrosion test was carried out in high pressure steam of 10.3 MPa at 783 K for 24 h. No nodules appeared on the specimens cut from welded areas of the channel box and nodular corrosion resistance tended to be better with increasing neutron fluence. Microstructural evolution in the form of irradiation-induced release of Fe atoms from Zr(Fe, Cr)2 type precipitates was detected by an analytical electron microscope. It was found that the higher the concentration of dissolved Fe and Cr in the grains of Zircaloy-4, the better the nodular corrosion resistance.
53 citations
TL;DR: In this article, a specific study was carried out to measure the influence of texture on the behavior of Zircaloy-4 under iodine-induced stress corrosion cracking, and the results showed that texture is a more important parameter than chemical composition, on condition that this composition remains within the ASTM specification.
47 citations
Patent•
04 Feb 1992TL;DR: In this article, a synthetic metallic heart valve fabricated of zirconium or zirconsium alloys that are coated with blue-black ZIRconium oxide or ZIRCONium nitride to provide a surface that is hard, biocompatible, resistant to impact, cavitation, and microfretting wear and exhibits improved hemocompatibility.
Abstract: The invention provides synthetic metallic heart valves fabricated of zirconium or zirconium alloys that are coated with blue-black zirconium oxide or zirconium nitride to provide a surface that is hard, biocompatible, resistant to impact, cavitation, and microfretting wear and exhibits improved hemocompatibility.
46 citations
TL;DR: In this article, atom probe studies have been performed on various specimens of Zircaloy-2 in order to measure the concentrations of the alloying elements Cr, Fe, and Ni in the α-Zr matrix.
33 citations
Patent•
24 Dec 1992
TL;DR: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing tin in a range of 0.45 to 0.75 wt. %.
Abstract: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing tin in a range of 0.45 to 0.75 wt. %, and typically 0.6 wt. %; iron in a range of 0.4 to 0.53 wt. %, and typically 0.45 percent; chromium in a range of 0.2 to 0.3 wt. %, and typically 0.25 percent; niobium in a range of 0.3 to 0.5 wt. %, and typically 0.45 wt. %; nickel in a range of 0.012 to 0.03 wt. %, and typically 0.02 wt. %; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 1,000 to 2,000 ppm, and typically 1,600 ppm, with the balance zirconium. The addition of iron and niobium improves mechanical properties of the alloy with its lower level of tin, while corrosion resistance is addressed by having an iron level of 0.45 wt. % and an iron/chromium ratio on the order of 1.5. The addition of niobium also counters the effect of higher iron on the hydrogen absorption characteristics of the alloy. The addition of nickel, silicon, carbon, and oxygen as alloying elements provide desired corrosion resistance and strength.
31 citations
Patent•
30 Apr 1992
TL;DR: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance under neutron irradiation environment in a zirconium alloy containing tin in a range of 8 to 12 percent, iron in a ranges of 2 to 5 percent, and typically 035 percent, chromium in a ranging of 0.25 to 0.30 percent, niobium in range of from a measurable amount up to 06 percent, Niobium, silicon, and oxygen were added as an alloying element to reduce hydrogen absorption by the alloy and to reduce variations in the corrosion resistance with
Abstract: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance under neutron irradiation environment in a zirconium alloy containing tin in a range of 08 to 12 percent; iron in a range of 02 to 05 percent, and typically 035 percent; chromium in a range of 01 to 04 percent, and typically 025 percent; niobium in a range of from a measurable amount up to 06 percent, and typically 030 percent; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 900 to 1800 ppm, typically 1600 ppm The silicon is added as an alloying element to reduce hydrogen absorption by the alloy and to reduce variations in the corrosion resistance with variations in the processing history of the alloy
31 citations
TL;DR: In this paper, the contact angles of Zr-Ni, ZrCu and ZrCo alloys against PSZ were measured by the sessile drop method, and the fracture shear strength of this joint was 55 MPa.
Abstract: The contact angles of Zr-Ni, Zr-Cu and Zr-Co alloys against PSZ were measured by the sessile drop method. Each alloy wetted PSZ very well. Zr-Co alloys showed a different behaviour. Joints of PSZ plates were obtained using Zr-17Ni alloy. At the joint interface, internal oxidation of zirconium occurred. The fracture shear strength of this joint was 55 MPa.
28 citations
Patent•
11 Aug 1992TL;DR: An electrochemical hydrogen storage alloy having decreased hydrogen overpressure as well as other desirable operational parameters, comprising on an atomic percent basis: 14 to 22 percent vanadium; 28 to 39 percent nickel; 7 to 15 percent titanium; 15 to 34 percent zirconium; and at least one member selected from the group consisting of 0.26 to 1:0.68.
Abstract: An electrochemical hydrogen storage alloy having decreased hydrogen overpressure as well as other desirable operational parameters, comprising on an atomic percent basis: 14 to 22 percent vanadium; 28 to 39 percent nickel; 7 to 15 percent titanium; 15 to 34 percent zirconium; and at least one member selected from the group consisting of 0.001 to 7 percent chromium, 0.001 to 7 percent cobalt, 0.001 to 7 percent iron, 0.001 to 3.6 percent manganese, and 0.001 to 2.7 percent aluminum, wherein the atomic ratio of the vanadium to zirconium is in the range of 1:2.26 to 1:0.68. An electrochemical hydrogen storage alloy having a reduced self-discharge rate comprising an alloy having a heterogeneous, disordered microstructure resulting from changes in the mutual solubility of the elements of the alloy, wherein hydrogen in a particular phase is not easily discharged either though low surface area, or an oxide of limited porosity or catalytic property.
27 citations
TL;DR: In this paper, Zirconium alloys were laser surface melted (LSM) using a continuous wave CO2 laser at energy densities of 4,7 and 10 kJ cm−2.
Abstract: Zirconium alloys were laser surface melted (LSM) using a continuous wave CO2 laser at energy densities of 4,7 and 10 kJ cm−2. LSM samples examined using SEM and optical microscopy exhibited resolidified regions with several different microstructures, including ultrafine martensite. Corrosion performance was obtained by steam autoclave tests and immersion tests in 10% FeCl3 at room temperature. Coarser microstructures performed better than fine microstructures in autoclave tests, while fine microstructures performed better than coarse microstructures in 10% FeCl3 immersion tests. Accelerated corrosion in the autoclave and immersion tests was observed to occur near the laser beam overlap region. The surface chemistry was examined for alloy segregation using secondary iron mass spectroscopy. Tin and iron alloy elements segregated near the periphery of each melt pool. Segregated regions containing increased iron concentrations associated with each laser pass were responsible for accelerated corrosion.
TL;DR: In this article, an analysis of the various cases of local enhancement of the corrosion rate of zirconium alloys under irradiation was carried out and it was observed that in most cases a strong emission of energetic β− is present leading to a local energy desorption rate higher than the core average.
TL;DR: In this paper, an X-ray diffraction profile analysis of hydrided Zircaloy-4 has been performed to understand the mechanism of rupture and to predict the threshold stresses for hydride stress orientation, especially the microstrain caused by crystalline lattice misfit.
Abstract: Zircaloy-4, used as cladding tube material in the nuclear reactors, may become brittle due to the precipitation of hydrides. During hydride formation, the anisotropic misfit strains between hydrides and the hexagonal-close-packed zirconium matrix results in a preferred orientation of the hydride platelets in the anisotropic stress field caused by non-relieved fabrication residual stresses and misfit stresses. To understand the mechanism of rupture and to predict the threshold stresses for hydride stress orientation, it is necessary to study the residual stresses, especially the microstrain caused by crystalline lattice misfit, in a hydrided specimen. The X-ray diffraction profile analysis is very sensitive to all the microstructure evolution in metallic materials. It is a non-destructive and voluminal technique compared with transmission electron microscope observation. The XRD peak broadening can be related closely with the microstrain in case of hydrided Zircaloy-4, because the hydride formation creates in general a great number of dislocations which contributes especially to the diminution of coherent domain size and to the increase of microstrain. To calibrate the internal microstrain due to precipitation effect of hydrided specimens, XRD profile analysis has also been realized on the non-hydrided specimens deformed by uniaxial tension. In this paper the authors restrict to analyzing more » the results about the recrystallized state, because more informations about the anisotropic elasticity, plasticity, thermal expansion, neutron diffraction measurement and the crystallographic texture results are available. « less
Patent•
30 Dec 1992
TL;DR: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing on a weight percentage basis tin in a range of 0.4 to 1.0 percent and typically 0.46 percent; chromium and silicon in the range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a ranges 1200 to 2500 ppm, typically 1800 to 2200 ppm as discussed by the authors.
Abstract: A stabilized alpha metal matrix provides an improved ductility, creep strength, and corrosion resistance against irradiation in a zirconium alloy containing on a weight percentage basis tin in a range of 0.4 to 1.0 percent and typically 0.5; iron in a range of 0.3 to 0.6 percent, and typically 0.46 percent; chromium in a range of 0.2 to 0.4 percent, and typically 0.23 percent; silicon in a range of 50 to 200 ppm, and typically 100 ppm; and oxygen in a range 1200 to 2500 ppm, typically 1800 to 2200 ppm. The high oxygen level assists in reducing hydrogen uptake of the alloy compared to Zircaloy-4, for example.
TL;DR: In this article, the tensile deformation of the Zr-1% Nb alloy was investigated in the temperature interval 295 to 773 K. The activation energy of 217.6 kj mol−1 was obtained from the shift of elongation minimum temperature.
Patent•
13 Oct 1992
TL;DR: A corrosion resistant metallic coating (60) of zirconium nitride is applied to the cladding tube (40) of a nuclear fuel rod (20) as mentioned in this paper.
Abstract: A corrosion resistant metallic coating (60) of zirconium nitride is applied to the cladding tube (40) of a nuclear fuel rod (20). The zirconium nitride is reactively deposited on a zirconium-alloy cladding tube by a cathodic arc plasma deposition process. The zirconium nitride coating provides superior wear test results and enhances the corrosion resistance of the cladding tube.
Patent•
21 Feb 1992
TL;DR: In this paper, a structural part for a nuclear reactor fuel assembly includes a zirconium alloy material having at least one alloy ingredient selected from the group consisting of oxygen and silicon, a tin alloy ingredient, a zinc, chromium, and a remainder of ZIRCONIUM and unavoidable contaminants.
Abstract: A structural part for a nuclear reactor fuel assembly includes a zirconium alloy material having at least one alloy ingredient selected from the group consisting of oxygen and silicon, a tin alloy ingredient, at least one alloy ingredient selected from the group consisting of iron, chromium and nickel, and a remainder of zirconium and unavoidable contaminants. The zirconium alloy material has a content of the oxygen in a range of substantially from 700 to 2000 ppm, a content of the silicon of substantially up to 150 ppm, a content of the iron in a range of substantially from 0.07 to 0.5% by weight, a content of the chromium in a range of substantially from 0.05 to 0.35% by weight, a content of the nickel of substantially up to 0.1% by weight, and a content of the tin in a range of substantially from 0.8 to 1.7% by weight. The alloy ingredients selected from the group consisting of iron, chromium and nickel are precipitated out of a matrix of the zirconium alloy as secondary phases, having a diameter with a geometric mean value in a range of substantially from 0.1 to 0.3 μm. The degree of recrystallization of the zirconium alloy is less than or equal to 10% and a sample of the zirconium alloy, after a recrystallization annealing with a degree of recrystallization of 97±2%, has a mean grain diameter less than or equal to 3 μm.
TL;DR: The oxide-metal interface formed during the aqueous corrosion of a Zr-2.5 wt% Nb alloy has been studied by SEM, using a specimen preparation method which also has potential application to other materials containing an oxidemetal interface as mentioned in this paper.
Abstract: The oxide-metal interface formed during the aqueous corrosion of a Zr-2.5 wt% Nb alloy has been studied by SEM, using a specimen preparation method which also has potential application to other materials containing an oxide-metal interface. The oxidation of cold-worked Zr-2.5 wt% Nb pressure tubing in pressurized lithiated water proceeds first along grain boundaries at which there is β-zirconium or its decomposition products, and then continues on the α-zirconium grains. Oxides at the oxide-metal interface formed in an aqueous environment were mainly extended along grain boundaries, and were characterized by long filaments which consisted of many fine zirconia grains. This oxidation behaviour is attributed to short-circuit diffusion at the grain boundaries which is caused by the nature of the crystallite boundaries of the oxide, the flaws arising from the oxidation of the grain boundary phases (β-Zr, its decomposition products as well as impurities), and cracking of the oxide due to phase transformations in ZrO2. When different plane sections of the tubing are corroded, there are different corrosion rates, with the sections containing a higher area fraction of the grain boundaries with β-zirconium exhibiting higher corrosion rates. The formation of long ‘fingers’ of oxide made up of longer filaments of oxide, together with the higher corrosion rates in the post-transition specimens, result in the destruction of the barrier oxide layer and an increase in the oxide rate.
Patent•
30 Jun 1992
TL;DR: A burnable absorber can be used to control axial power peaking or moderator temperature coefficient while additional elements are added to improve strength and/or corrosion resistance in a zirconium alloy containing erbium in a range of from about 0.05 to 2 wt..
Abstract: A burnable absorber controls axial power peaking or moderator temperature coefficient while additional elements are added to improve strength and/or corrosion resistance in a zirconium alloy containing erbium in a range of from about 0.05 to 2 wt. % selected from the group consisting of a naturally occurring distribution of isotopically enriched erbium-167 and a combination thereof; in a range of from a measurable amount up to 1.4% tin; from 0.2 to 0.5 wt. % iron; from 0.07 to 0.25 wt. % chromium; in a range of from a measurable amount up to 0.6 wt. % niobium; in a range of from a measurable amount up to 0.5 wt. % vanadium; 50-120 ppm silicon; 1000-2200 ppm oxygen and a balance of zirconium. Alternatively, the erbium can be replaced by gadolinium in a range of from about 0.05 to 6 wt. % selected from the group consisting of a naturally occurring distribution of gadolinium isotopes, isotopically enriched gadolinium-157 and a combination thereof.
01 Jan 1992
TL;DR: In this paper, a model based on image analysis and hydride embrittlement micro-mechanism observations is used to calculate the upper limit hydrogen content which makes Zircaloy-4 totally brittle.
Abstract: In order to better understand the embrittlement of Zircaloy-4 by hydrides and the ductile-brittle transition on this alloy, Zircaloy-4 sheet tensile specimens in the stress-relieved, recrystallized and β treated states were hydrided (10 to 1500 ppm wt H) and then tested at two temperatures (20°C, 350°C). Metallographic and fractographic analyses were carried out to determine the fracture micro-mechanisms. The results showed that, at 20°C, Zircaloy-4 undergoes a significant ductile to brittle transition for high hydrogen contents. Heat treatment shifts this transition (to zero elongation) considerably, from 1050 ppm wt H for the stress-relieved state to less than 250 ppm wt H for the β treated state. However, at 350°C, Zircaloy-4 remains ductile up to hydrogen content higher than 1100 ppm wt. At 20°C, the fracture surfaces are characterized by voids and secondary cracks for low and medium hydrogen contents, and by intergranular crack and decohesion through the continuous hydride network for high hydrogen content. A model based on image analysis and hydride embrittlement micro-mechanism observations is used to calculate the upper-limit hydrogen content which makes Zircaloy-4 totally brittle. The difference between the mechanical behaviors of stress-relieved and recrystallized states is also explained.
TL;DR: In this paper, a new approach has been made for a general analysis of oxide specimens from scales grown on the zirconium-based cladding alloys of PWR rods to analyse the morphology of these scales, the topography of the oxide/metal interface and the crystal structures close to this interface.
TL;DR: In this article, the dimensional changes of Zirconium alloys are predicted using texture analysis and crystal plasticity to predict the lifetime of Zircaloy fuel cladding, both out ofpile and in-reactor.
Abstract: Zirconium alloys are commonly used in light-water reactors as thin-walled tubing to clad highly radioactive fuel. The tubes experience varied stresses at high temperatures while being exposed to high-neutron radiation, resulting in thermal creep and radiation growth and creep. However, the dimensional stability of these materials is important to preventing leakage of fission gases and contamination of the coolant water. Predicting the dimensional changes of the thin-walled tubes is further complicated by the anisotropic nature of the hexagonal close-packed metals. This article summarizes the procedures used in the texture analyses and crystal plasticity in developing model equations to predict the dimensional changes of Zircaloy fuel cladding, both out-of-pile and in-reactor. These methodologies can be extended to the life prediction of these important structures in nuclear reactors.
TL;DR: In this paper, the thermoelectric power (TEP) and electrical resistivity measurements of various zirconium alloys (Zircaloys-2 and -4, Zr-Fe, Zhr-O-Fe) were studied by TEP measurements and a good correlation was found between the variations of these two properties due to precipitation.
TL;DR: In this article, Zirconium cobalt (ZrCo) is evaluated as an alternative to uranium for tritium storage, and the pumping speed and pressure-composition-temperature isotherm profiles for the ZrCo hydrogen and triti...
Abstract: Zirconium cobalt (ZrCo) is being evaluated as an alternative to uranium for tritium storage. The pumping speed and pressure-composition-temperature isotherm profiles for the ZrCo hydrogen and triti...
TL;DR: In this article, the effect of alloying additions to zirconium through the results obtained on the low stress creep behavior of zircaloy-2 has been studied employing helical spring specimens.
Abstract: Zircaloy-2 is a popular zirconium base nuclear fuel cladding material having tin as a major alloying additions along with Fe, Ni and Cr, each around 0.1 wt %. Low stress creep behavior of a-zirconium and zircaloy-2 has been studied employing helical spring specimens. Detailed experimental results are reported and analyzed elsewhere. Here we present data on a-Zr and zircaloy-2 obtained under identical test conditions of grain size and test temperature with the aim of bringing out the effect of alloying additions to zirconium through the results obtained on the low stress creep behavior of zircaloy-2. The high purity zirconium contained, by ppm, 110 Sn, 350 Fe, 37 Ni, 150 Cr and 880 O while zircaloy-2 contained, by wt %, 1.3 Sn, 0.16 Fe, 0.06 Cr, 0.04 Ni and 0.11 O. Helical spring specimens of wire radius 500{mu}m and spring radius 0.012m were employed in both the cases. Spring specimens were annealed under vacuum (Zr at 923K/1.8ks and zircaloy-2 at 1113K/2.7ks) while on mandrel to develop a matching grain size in both the materials. The mean linear intercept grain size was found to be 16.8{mu}m in alpha zirconium and 16.6{mu}m in zircaloy-2. Close coiled helices were creep tested at 873K using more » the set-up and procedure described elsewhere. The maximum applied stress was 5.3MN/m{sup 2} in the case of zirconium while the same in the case of zircaloy-2 was 3.9MN/m{sup 2}. Deflection versus time curves for tests conducted at 873K for both the materials are shown in this paper. « less
TL;DR: In this article, thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal, and a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions.
Abstract: This paper reports that, in pressurized heavy water nuclear reactors of the type standardly used in Canada (Canada deuterium uranium-pressurized heavy water reactors), the zirconium alloy pressure tubes of the core absorb deuterium produced by corrosion reactions. This deuterium weakens the tubes through hydrogen embrittlement. Thin palladium coatings on the outside of the zirconium are analyzed as a method for deuterium removal. This coating is expected to catalyze the reaction D{sub 2} + 1/2O{sub 2} {r reversible} D{sub 2}O when O{sub 2} is added to the annular (insulating) gas in the tubes. Major reductions in the deuterium concentration and, hence, hydrogen embrittlement are predicted. Potential problems such as plating the tube geometry, neutron absorption, catalyst deactivation, radioactive waste production, and oxygen corrosion are shown to be manageable. Also, a simple set of equations are derived to calculate the effect on diffusion caused by neutron interactions. Based on calculations of ordinary and neutron flux induced diffusion, a palladium coating of 1 {times} 10{sup {minus}6} m is recommended. This would cost approximately $60,000 per reactor unit and should more than double reactor lifetime. Similar coatings and similar interdiffusion calculations might have broad applications.
Patent•
14 Feb 1992
TL;DR: A zirconium alloy which imparts good creep strength, while also providing favorable neutron cross section, improved corrosion resistance, low hydrogen uptake and good fabricability is described in this paper.
Abstract: A zirconium alloy which imparts good creep strength, while also providing favorable neutron cross section, improved corrosion resistance, low hydrogen uptake and good fabricability is described which contains vanadium in a range of from an amount effective to indicate its greater-than-trace presence up to 1.0 wt %, wherein either limit is typical; niobium in a range of from an amount effective to indicate its greater-than-trace presence up to 1.0 wt %, wherein either limit is typical; antimony in a range of from an amount effective to indicate its greater-than-trace presence up to 0.2 wt %, wherein either limit is typical; tellurium in a range of from an amount effective to indicate its greater-than-trace presence up to 0.2 wt %, wherein either limit is typical; tin in a range of from an amount effective to indicate its greater-than-trace presence up to 0.5 wt %, wherein either limit is typical; iron in a range of 0.2 to 0.5 wt %, typically 0.35 wt %; chromium in a range of from 0.1 to 0.4 wt %, typically 0.25 wt %; silicon in a range of 50 to 200 ppm, wherein either limit is typical; and oxygen in a range of from an amount effective to indicate its greater-than-trace presence up to 2200 ppm, wherein either limit is typical and the balance zirconium.
TL;DR: In this article, the potentiodynamic oxidation of zirconium, zircaloy-2 (Zry-2) and Zry-4 was studied in the 0 V ⩽ V⩽ 8 V potential range.
TL;DR: In this article, a combination of XPS and SNMS depth profiling is applied to examine changes in the oxidation state, as well as in the relative concentrations of the metal and of the zirconium components.
Patent•
30 Jun 1992
TL;DR: In this paper, the authors describe a coating for zirconium alloy components of nuclear reactor fuel assemblies, which consists of a metal silicate binder, particles of burnable-poison particles, such as boron carbide, optional graphite particles and an optional rheology-enhancing component.
Abstract: Coatings for zirconium alloy components of nuclear reactor fuel assemblies are described. The coating consists of a metal silicate binder, particles of burnable-poison particles, such as boron carbide, optional graphite particles and an optional rheology-enhancing component. The coating is deposited from a liquid suspension which also includes a polar solvent.