Showing papers on "Zirconium alloy published in 1977"

Nuclear fuel element

Abstract: A nuclear fuel element for use in the core of a nuclear reactor is disclosed and has an improved composite cladding comprised of a moderate purity metal barrier of zirconium metallurgically bonded on the inside surface of a zirconium alloy tube. The metal barrier forms a shield between the alloy tube and a core of nuclear fuel material enclosed in the composite cladding. There is a gap between the cladding and the core. The metal barrier forms about 1 to about 30 percent of the thickness of the composite cladding and has low neutron absorption characteristics. The metal barrier serves as a preferential reaction site for gaseous impurities and fission products and protects the alloy tube from contact and reaction with such impurities and fission products. Methods of manufacturing the composite cladding are also disclosed.

Corrosion of zirconium-base alloys: an overview

Abstract: The corrosion and hydriding performance of zirconium-base alloys under pressurized water reactor (PWR) and boiling water reactor (BWR) conditions, as gaged by a comprehensive review of the technical literature, has been evaluated. Starting with a brief historical description of the development of zirconium for cladding and structural material in nuclear reactors and the corrosion problems associated with the use of the pure metal, it is shown that the development of zirconium-base alloys proceeded down two major paths. One development involved the zirconium-tin system and led to the development of the Zircaloys, whereas the other concentrated upon zirconium-niobium materials and produced the two major alloys of this system in use today: Zr-1Nb and Zr-2.5Nb. The corrosion data generated for each system, both in and ex-reactor, are evaluated, and the benefits and potential problems associated with each alloy are discussed for both PWR and BWR applications. Potential areas of concern for the Zircaloy alloys in both applications include exposure temperature limitations and the formation of nonuniform accelerated corrosion products in the oxygenated irradiation environment. The zirconium-niobium alloys are found to be very sensitive to oxygen in the coolant and to prior heat treatment in ex-reactor experiments but show either minimum or negativemore » acceleration due to the presence of neutron irradiation. Alloys that combine these two additives (for example, Zr-3Nb-1Sn and Ozhennite-0.5) do not appear to show promise as possible replacements for the Zircaloys under present-day conditions. The lack of a unifying theory for describing the mechanisms involved in the corrosion of zirconium-base alloys may hamper seriously possible future applications under different design conditions.« less

Elastic constants of the bcc phase in niobium-zirconium alloys between 4.2 and 300 K

Abstract: The adiabatic elastic constants have been determined for quenched single crystals of niobium‐zirconium alloys of the following nominal compositions: 10, 15, 20, 25, 30, 35, 40, 45, 50, 55, and 70 at.% Zr. The c11, c12, and c44 constants as well as the shear constant C′= (1/2) (c11−c12) and the bulk modulus are given between 4.2 and 300 K. We observe for the whole range of compositions an anomalous temperature behavior for c44 similar to that found in pure Nb, but here the amplitude of the effect disappears progressively with increasing Zr concentration. The Debye temperatures were computed from the constants at 4.2 K.

Surface segregation in alloys: dilute binary iron/zirconium solid solutions

Abstract: We have examined the equilibrium surface composition of two very dilute Fe–Zr binary alloys: Zr–0.05‐at. %‐Fe and Fe–0.06‐at. %‐Zr. We find that reversal of the segregating element occurs in the respective dilute limits. Iron segregates to the surface of the zirconium rich alloy, while zirconium segregates to the surface of the iron rich alloy. Segregation of iron to the surface of the zirconium rich alloy was reversible for T≳970 K with a heat of segregation of 71.1±8 kJ/mole. Segregation of zirconium to the surface of the iron rich alloy was large and not reversible. The observed behavior is discussed in light of two widely used models of surface segregation.

Pellet-cladding interaction phenomenon

Abstract: Results of three related projects undertaken to elucidate the mechanism of Zircaloy cladding fracture caused by pellet-cladding interaction (PCI) in water reactor fuel rods are described. A detailed microscopic examination of incipient i.d. cladding defects in some Maine Yankee Core I fuel rods determined that these defects and clad penetrations in related rods were caused by a PCI mechanism that was promoted by chemical species, i.e., stress corrosion cracking (SCC). A consideration of the internal fuel rod chemistry and fission product distribution indicates that one potential agent for SCC of Zircaloy cladding is iodine released from CsI deposited on the i.d. surface and another is cadmium metal. A simple analytical model of crack propagation in Zircaloy cladding based on linear elastic fracture mechanics indicates two possible rate-controlling events, depending on the value of the stress intensification K/sub ISCC/. If K/sub ISCC/ for irradiated Zircaloy is very low, i.e., on the order of 2.2 to 3.3 MN/m/sup 3///sup 2/ (2 to 3 ksi ..sqrt..in.), crack growth is relatively easy, and hence the rate-limiting step must be the nucleation of sharp cracks in the cladding i.d. surface. However, if K/sub ISCC/ for irradiated Zircaloy is relatively large, i.e., greater than or equalmore » to 11 MN/m/sup 3///sup 2/ (10 ksi ..sqrt..in.), a high interfacial friction coefficient, for example, caused by fuel-clad bonding, would be required to propagate the i.d. defect.« less

Zirconium alloy heat treatment process and product

Abstract: Zirconium-base alloy channels and fuel cladding tubes having unique resistance to accelerated pustular corrosion in the boiling water reactor environment are produced by a heat treatment causing segregation of intermetallic particulate precipitate phase in two dimensional arrays preferably located along grain boundaries and subgrain boundaries throughout the alloy body.

Process for regenerating a pickle acid bath

Abstract: A HF-HNO 3 pickle acid bath for pickling zirconium, zirconium alloys, hafnium, and hafnium alloys is regenerated by adding to the spent pickle acid a sufficient amount of NaF to precipitate out, in the case of zirconium, ZrF 4 as Na 2 ZrF 6 . HF and HNO 3 are added to the bath to make up losses and the pickle acid is recycled for use.

Susceptibility of Zirconium Alloys to Delayed Hydrogen Cracking

Abstract: Smooth and notched cantilever beams and round-notched bars were machined from pressure tubes of cold-worked Zr-2.5 Nb and Zircaloy-2. They were loaded in the temperature range 290 to 520/sup 0/K. After two thermal cycles and at high stress, cracks were initiated in smooth beams of cold-worked Zr-2.5 Nb. Under the same test conditions, cold-worked Zircaloy-2 plastically deformed with no cracking. When notches were present, cracks propagated at the same rate in both materials by delayed hydrogen cracking. In cold-worked Zr-2.5 Nb, the crack velocity followed an Arrhenius plot with an apparent activation energy of 42 kJ/mol. Below 420/sup 0/K, the threshold stress intensity factor for delayed hydrogen cracking was about 5 MPa root m. Therefore, cracking can be prevented by keeping tensile stresses very low.

Uniaxial tensile properties of Zircaloy containing oxygen: summary report

Abstract: The uniaxial stress-strain behavior of Zircaloy-2 and -4, Zircaloy-oxygen alloys with a uniform oxygen distribution, and composite specimens with a ZrO2/alpha/beta layer structure was investigated over the range of experimental conditions: temperature 25-1400 degrees C; strain rate ; oxygen content 0.11 - 4.4 wt %; grain size 5-50 micrometers; texture longitudinal, transverse, and diagonal orientations; and microstructural state, which consists of the equiaxed alpha phase and various transformed beta acicular structures. The work-hardening and strain-rate sensitivity parameters were determined from the experimental results, and the tensile properties were correlated with oxygen concentration, oxygen distribution in the material, and microstructure. Dynamic strain-aging phenomena were observed in Zircaloy at 200, 400, and 700 degrees C, and super-plastic deformation occurred at 850 and 1000 degrees C. An increase in the oxygen concentration in homogeneous Zircaloy-oxygen alloys increased the ultimate tensile strength and decreased the total strain, particularly below approximately 900/sup 0/C. In composite specimens with the ZrO2/alpha/beta structure, the total oxygen content had little effect on the ultimate tensile strength below approximately 1000 degrees C, but the strength increased with oxygen content at higher temperatures. Information on the effects of grain size, oxygen content, texture, and strain rate on the stress-strain behavior suggests that the dominant mechanism of super-plastic deformation in Zircaloy near approximately 850/sup 0/C is grain-boundary sliding at the alpha-beta interface with accommodation by diffusional creep, dislocation slip, and grain-boundary migration. Good correlation was obtained between ductility and values of the strain-rate sensitivity parameter.

High temperature properties of zircaloy-oxygen alloys

Abstract: The effect of oxygen on thermal expansion, elastic moduli, and thermal diffusivity of Zircaloy-4 nuclear reactor fuel cladding was determined for the main purpose of providing baseline data for LOCA evaluation codes. Measurements were made over the temperature range 298 to 1473 K and 0.7 to 28 at% oxygen. Expansion and moduli were measured in the two relevant directions, while diffusivity was measured in the principal heat transfer direction (through the clad). Thermal expansion and elastic moduli both increased with oxygen, while diffusivity decreased. Similar measurements were made on Zircaloy-2, but only to 5 at% oxygen. Differences between the two alloys were within experimental error of the measurements made.

Zirconium alloys containing transition metal elements

Abstract: Zirconium alloys containing at least two of the transition metal elements of iron, cobalt and nickel are disclosed. The alloys consist essentially of at least two elements selected from the group consisting of about 1 to 27 atom percent iron, about 1 to 43 atom percent cobalt and about 1 to 42 atom percent nickel, balance essentially zirconium plus incidental impurities. The alloys in polycrystalline form are capable of being melted and rapidly quenched to the glassy state. Substantially totally glassy alloys of the invention evidence unusually high resistivities of over 200 μΩ-cm.

Fabrication of Zirconium Alloys into Components for Nuclear Reactors

Abstract: Over 80 percent of the world's annual output of zirconium is fabricated into components for nuclear reactors. In 1976 the CANadian Deuterium Uranium (CANDU) power reactor program will require nearly 500 Mg of zirconium and by 1986 the requirement will have risen to nearly 800 Mg. Most of this is fabricated into tubes which can cost as much as $80/kg. Small thin walled tubes for fuel cladding are fabricated by heavily cold working extruded tubes. The most important fabrication parameter in producing the desired crystallographic texture, hydride orientation, and mechanical properties is the cold reduction route. The deformation also must be controlled carefully to ensure that these properties are uniform, both through the wall thickness and around the circumference of the tube. Large-diameter thin-walled tubes are used to guide coolant flow, guide control rods, or for calandria tubes. These tubes can be made by seam welding annealed sheet and they can be made to very close dimensional tolerances. The weld and heat-affected zones must be recrystallized by cold work and heat treatment so that they have adequate ductility after irradiation. Pressure containment tubes usually are fabricated by hot extrusion and cold worked a limited amount to size. The most important fabrication parameters are the forging and extrusion processes because they determine the crystallographic texture and the microstructure of the tubes. The advantages of different fabrication techniques and the properties of the tubes are discussed.

Cadmium metal embrittlement of Zircaloy-2

Abstract: EFFECTS of chemical environments upon the mechanical properties of zirconium alloys are of great interest because such alloys are used to contain fissionable fuel in water-cooled nuclear reactors. Particular interest attaches to stress corrosion cracking or liquid metal embrittlement because both of these phenomena are known to produce non-ductile fractures in otherwise ductile materials. Rosenbaum et al. reported iodine stress corrosion cracking of Zircaloy-2 some years ago1,2, and this phenomenon has been extensively investigated by others3–5. Zircaloy-2 is an alloy of zirconium containing by weight about 1.5% tin, 0.15% iron, 0.1% chromium, 0.05% nickel and 0.12% oxygen. Cox has reported embrittilement of Zircaloy by mercury or liquid caesium at room temperature3. However, the phenomenon of liquid metal embrittlement is known to be favoured by low temperatures6, and Zircaloy nuclear fuel cladding operates in the vicinity of 300 °C.

Oxidation of Zirconium During a High-Temperature Transient

Abstract: A physical model describing oxidation and oxygen diffusion under non-isothermal conditions is presented. It assumes (a) equilibrium at phase interfaces, as is usual in the treatment of diffusion, and (b) oxygen supersaturation of the ..beta..--zirconium to precipitate ..cap alpha..--zirconium on cooling. An important consideration, not required in isothermal diffusion, is that the equilibrium oxygen concentrations in the ..cap alpha..- and ..beta..-phases decrease with temperature. During isothermal oxidation at temperatures above 850/sup 0/C, the phase distribution is zirconium dixoide (ZrO/sub 2/)/..cap alpha../..beta...

Transient Species Participating in the SCC of Zirconium Alloys

Abstract: A study of the chemistry of the intergranular dissolution of zirconium has been made in a range of iodine/organic solutions. This process is the second of the 3 stages through which the stress corrosion cracking (SCC) of the Zircaloys proceeds. The first being cracking or breakdown of the surface oxide film and the last a relatively fast transgranular propagation process. From a study of the effects of various additives on the time to failure of smooth split-ring specimens, together with fractography of the resulting cracks, it is concluded that competition reactions between the various components of the solution and transient chemical species in solution control the extent of the intergranular dissolution step. A comparison with published chemical reaction rates suggests a hypothesis whereby this dissolution process proceeds via solvated electron and low valence zirconium intermediates. Additives which react rapidly with solvated electrons, or which complex with the lower valence states of zirco...

Oxidation of Zirconium Alloys in Steam at 1000 to 1850??C

Abstract: Oxygen embrittlement of fuel sheathing resulting from high temperature oxidation in steam is a potential fuel failure mechanism during a loss-of-coolant accident (LOCA) In high temperature steam, zirconium alloys form anouter layer of zirconium oxide (ZrO 2 ) and an inner layer of oxygen-stabilized α-Zr immediately below The metal/steam reaction for Zr-25Nb in the temperature range 1000 to 1600°C was studied and compared with earlier results for Zircaloy-2 and Zircaloy-4 Reaction kinetics and the rate of growth of the combined (ZrO 2 + α-zirconium) layers were measured The reaction rates for all three alloys were similar and obeyed parabolic rate laws The reaction rates were consistently less than those given by the Baker and Just correlation for zirconium oxidation in steam A discontinuity was observed in the temperature dependence of the rate of growth of the combined (ZrO 2 + α-zirconium) layer, and this discontinuity was attributed to a change in the oxide crystal structure The thickness ratio ZrO 2 /α-zirconium also varies with temperature A method for estimating oxidation temperature and time from metallography, based on these observations, is discussed

The Effect of Niobium Additions on the Corrosion Behavior of Zircaloy-4

Abstract: Alloying additions of 0.5 and 1.0 wt% niobium, respectively, have been added to Zircaloy-4 in an attempt to improve its high-temperature corrosion resistance. Ingots of these modified alloys were f...

Alloy useful as hydrogen storage material

Abstract: OF THE DISCLOSURE An alloy represented by the general formula Tix-yZryCr2-2Mnz, where 1 ? x ? 1.3, 0 < y ? 1, and 0 < z < 2. This alloy readily forms complex metal hydrides at low hydrogen pressures and can absorb a large quantity of hydrogen. The absorbed hydrogen can readily be liberated at room temperature.

Nodular Corrosion of Zircaloy-2 and Some Other Zirconium Alloys in Steam Generating Heavy Water Reactors and Related Environments

Abstract: In steam generating heavy water reactors (SGHWR), nodular or "patch-type" corrosion is the dominant form of attack on Zircaloy, as it is in other boiling or oxygenated reactor environments. The most severe oxidation occurs close to the stainless steel spacer grids, although nodules are seen elsewhere on cladding and on non-fueled components, a maximum thickness of 160 μm at 20 MWd/kg uranium local burn-up having been recorded to date. The extensive data from post-irradiation examination of SGHWR fuel elements are reviewed and discussed from the points of view of determination of rate laws for oxide growth and the establishment of corrosion mechanisms. A rate law for oxide growth on Zircaloy-2 at grid positions is determined empirically by seeking correlations between oxidation and irradiation exposure. It is concluded that a linear relationship between oxide thickness and local fast fluence or burn-up represents the data reasonably well, the dependence upon channel flow and exit quality being less significant. Zr-2.5Nb alloy, in the recrystallized condition, was more resistant to nodular oxidation than was Zircaloy-2. The mechanism of nodular corrosion cannot be expressed in quantitative terms yet, but a qualitative approach is valuable in suggesting wider ranging correlations and possible ameliorative measures. The suggested mechanism requires: (a) an enhancement of the general corrosion rate, controlled in-reactor by coolant chemistry and radiolysis but dependent also upon galvanic coupling or flow effects, or both; and (b) factors specifically responsible for nodule initiation and growth, for example, local cathodic depolarization at intermetallic inclusions and stress-induced recrystallization of the oxide. The effect of metallurgical factors in the metal substrate and of surface treatment of the alloy also are discussed and current experimental work, intended to check some of the postulated features of the oxidation mechanism, is described.