Showing papers on "Zirconium alloy published in 1979"

MATPRO-Version 11: a handbook of materials properties for use in the analysis of light water reactor fuel rod behavior

Abstract: This handbook describes the materials properties correlations and computer subcodes (MATPRO-Version 11) developed for use with various LWR fuel rod behavior analytical programs at the Idaho National Engineering Laboratory. Formulations of fuel rod material properties, which are generally semiempirical in nature, are presented for uranium dioxide and mixed uranium--plutonium dioxide fuel, zircaloy cladding, and fill gas mixtures.

A proposed criterion for the oxygen embrittlement of Zircaloy-4 fuel cladding

Abstract: Under the condition postulated for a loss-of-coolant accident (LOCA), Zircaloy-4 fuel cladding may experience a temperature transient during which it absorbs an appreciable amount of oxygen from the coolant. Theoretical models for predicting cladding behavior during loss-of-coolant (LOC) are being developed, but until the failure mechanisms can be clearly established, an empirical criterion must be employed. Since, under LOCA conditions, the fuel cladding would be predominantly exposed to tensile hoop stresses, the effects of oxygen on the tensile properties of Zircaloy-4 were investigated. It was found that, from room temperature to 800°C, the tensile properties were essentially independent of maximum cladding temperature and cooling rate, but were dependent on the oxygen distribution through the cladding wall. On the basis of the present work, but also consistent with other published work, an interim oxygen embrittlement criterion is proposed, which states: "the oxygen concentration must be less than 0.7 weight percent over at least half the cladding thickness." Since the proposed criterion requires a prediction of the oxygen distribution for a given temperature transient, the available computer models are briefly described and compared with experiment. Preliminary results suggest that the prediction of OXWEX, SIMTRAN 1, and PRECIP 1 are in reasonable agreement with experiment.

Structure of PWR Primary Corrosion Products

Abstract: The composition and structure of “erud” taken from the fuel elements of various pressurized water reactors (PWR's) were determined by SEM, EDAX, X-ray analysis, infrared spectroscopy, and thermomagnetic analysis. The samples were taken from reactors varying in materials of construction from all stainless steel to mainly Inconel 600 and Zircaloy fuel cladding. It is shown that crud is essentially a nickel ferrite (NixFe3–xO4) with 0.45 < x < 0.75. Calibration of the lattice parameter versus x allows convenient determination of the main constituents of crud samples. Chromium also enters the inverse spinel lattice substitutionally to give a composition CryNixFe3–x–yO4. Aluminum, when present, generally formed amorphous compounds with silicon in the loose crud.

The Effect of Dispersion and Spheroidization Treatment of δ Zirconium Hydrides on the Mechanical Properties of Zircaloy

Abstract: This study is basic research on some mechanical properties of Zircaloy-4 and Zircaloy-2 addressed particularly to the influence of hydrogen attack and of the hydride-orientation and -shape effect. At room temperature, Zircaloy-4 has almost the same tensile properties as does Zircaloy-2, both before and after hydriding. Zircaloy-4 may serve well if its hydrogen content is lower than 300 ppm, although hydrogen embrittlement can be alleviated by elevated temperature. If we performed a spheroidization treatment on the platelet hydrogen in the matrix, it may serve satisfactorily when the hydrogen content is 650 ppm or more. Tensile tests of annealed Zircaloy-2 specimens, of hydrided specimens, and of spheroidized specimens containing two different hydrogen concentrations were carried out at temperatures up to 700/sup 0/C. The strain-rate effect on the mechanical properties was also studied for Zircaloy-2 specimens. The results show that a spheroidization treatment of the hydrided Zircaloy-2 can improve its mechanical properties - i.e., its ductility, toughness, and strength - as well as its hardenability.

Out-of-Pile Testing of Iodine Stress Corrosion Cracking in Zircaloy Tubing in Relation to the Pellet-Cladding Interaction Phenomenon

Abstract: To investigate iodine stress corrosion cracking on the behavior of Zircaloy tubing, a standard test procedure with internally pressurized creep specimens (I-SCC standard test) and a split-ring test (I-SCC laboratory test)were developed. The threshold value for brittle cracking in iodine is 10 - 6 g iodine per square centimetre. The uniform elongation shows a clear minimum in its dependence on strain rate. Basal pole orientations in the range of ′50 to ′70 deg relative to the radial direction are the most I-SCC sensitive textures. The I-SCC process occurs in several stages; incubation, crack nucleation, and propagation. A thermodynamic evaluation indicates that I-SCC only occurs when zirconium iodides condense on the Zircaloy surface. Results show that comparisons of the I-SCC susceptibility of tubing manufactured in different manners should be made at the same point on the strain rate versus uniform elongation curves; for example, at the strain rate with the minimum of uniform elongation.

Zirconium Cladding Deformation in a Steam Environment with Transient Heating

Abstract: Zircaloy-4 cladding is being tested under loss-of-coolant accident (LOCA) conditions to determine its deformation behavior and to provide data for verification of analytical models. Data obtained thus far from 34 single-rod tests imply less ballooning, and consequently less flow restriction, than would be predicted from earlier data. The instrumented simulators, consisting of unirradiated Zircaloy-4 cladding with an internal heater to simulate fuel pellet heating, are tested to failure in superheated steam over a wide range of internal pressures at a temperature increase rate of about 28°C/s. Cladding surface temperature and internal pressure data are recorded during the transient and deformation measurements are obtained posttest. An analytical expression is given for the burst temperature as a function of burst pressure. The experimental results show excellent correlation between cladding deformation and surface temperature distribution. Deformation is extremely sensitive to small temperature variations.

Method for forming nuclear fuel containers of a composite construction and the product thereof

Abstract: An improved method for producing nuclear fuel containers of a composite construction having components providing therein a barrier system for resisting destructive action by volatile fission products or impurities and also interdiffusion of metal constituents, and the product thereof. The composite nuclear fuel containers of the method comprise a casing of zirconium or alloy thereof with a layer of copper overlying an oxidized surface portion of the zirconium or alloy thereof.

Relationship between contractile strain ratioR and texture in zirconium alloy tubing

Abstract: The crystallographic textures of zirconium alloy tubing used as cladding in nuclear reactor fuel are commonly characterized by the quantitative texture numbers F (Kaellstroem) and f/sub r/ (Kearns) which are derived from the direct and inverse pole figures. The texture numbers of zircaloy 2 and 4 tubes have been correlated experimentally with the value of the contractile strain ratio R which is a measure of the plastic anisotropy of the tube. The correlations were based on the results of 20 different tubing lots. The f/sub r/-R correlation shows much less data scatter than the F-R correlation. By assuming a simple plastic deformation model for zirconium alloys the following relations between texture and anisotropy are obtained: F = R-1/R+1 and f/sub r/ = R/R+1. The theoretically derived relations are in good agreement with the experimental data. The procedure of correlating texture with plastic anisotropy is not limited to zirconium alloy tubing, but should be equally applicable to textured sheet and plate materials and other alloys with a limited number of slip systems.

The Influence of Environment Purity on Gaseous Iodine Embrittlement of Zircaloy-2 and High Purity Zirconium

Abstract: The influences of environment purity on gaseous iodine embrittlement of high purity zirconium were investigated and compared to data obtained for Zircaloy-2. Iodide zirconium was embrittled at 350 C by constant extension rate tensile loading in highly purified and low constant partial pressure (40 Pa) iodine gas. Cracking was preceded by significant plastic deformation. Crack initiation probably involved sample pitting and subsequent cracking at the bottoms of pits. Brittle crack propagation proceeded by mixed intergranular separation and transgranular cleavage processes, compared to microvoid nucleation, growth, and coalescence typical of ductile fracture in argon. Purified and oxygenated (6 kPa) iodine gas also embrittled zirconium; however, the mechanism for crack initiation was influenced by oxygen additions. Sample pitting was eliminated, and crack initiation required local plastic strain concentrated at microstructural discontinuities such as grain boundaries. Similar findings were obtained...

Development of an Oxygen Embrittlement Criterion for Zircaloy Cladding Applicable to Loss-of-Coolant Accident Conditions in Light-Water Reactors

Abstract: To establish the mechanical response of Zircaloy cladding under thermal shock conditions typical of hypothetical loss-of-coolant accident (LOCA) situations in light-water reactors (LWRs), cladding specimens were ruptured in steam during transient heating (10 K/s), oxidized at maximum temperatures between 1140 and 1770 K for various times, and cooled from the isothermal oxidation temperature to ∼1100 K at a rate of 5 K/s, and rapidly quenched by bottom flooding with water at a rate of ∼0.05 m/s. Failure "maps" for fracture of the cladding by thermal shock were developed relative to the maximum oxidation temperature and various time-dependent oxidation parameters. In situ pendulum-load impact tests were conducted at room temperature on tubes that survived the thermal quench. Information on the total absorbed energy from these tests was correlated with more extensive results from instrumented drop-weight impact tests. The thermal shock results indicate that the present Zircaloy embrittlement criterion (that is, a total oxidation limit of 17 percent of the wall thickness and a maximum cladding temperature of 1477 K) is conservative and that a more quantitative criterion, based upon the mechanical behavior of the oxidized material, can be formulated with a specified degree of conservatism consistent with the mechanical loads imposed on the cladding during reflood and the maximum amount of oxidation set by the margin of performance of emergency core-cooling systems in LWRs.

Experimental investigation of various pellet geometries to reduce strains in zirconium alloy cladding

Abstract: In the quest to reduce cladding strain from power ramps and thereby reduce the incidence of defects, various aspects of pellet geometry have been examined. The influence of pellet end squareness wa...

High temperature deformation of oxidized β-Zr-Mo alloys

Abstract: The flow properties of β-phase Zr-Mo alloys were investigated by means of compression testing in a nominally pure (10 ppm O2) argon atmosphere. Experiments were carried out in the strain rate range 10-1 to 10-5 s-1 and from 900 to 1000°C. The stress-strain curves were unusual in that they exhibited a continuous decrease in flow stress with strain, after little or no work hardening. A further unusual feature of the data was that the flow stress in interrupted tests increased with delay time in all the alloys. By contrast, crystal bar Zr, tested under the same atmosphere, exhibited neither flow softening nor significant interruption hardening, but deformed in a conventional manner. The results obtained from X-ray investigations, as well as from interrupted tests and from tests carried out in a more purified atmosphere, indicated that the occurrence of both interruption hardening and flow softening was associated with the formation of an oxygen stabilized a-layer on the outer surface of the β-sample. Growth of the hard α-layer during annealing produces strengthening while its decrease in volume during deformation produces softening. A model, based on the assumption that the hard α-phase shares the load applied to the sample, was developed, and its predictions agree satisfactorily with the experimental observations. The extreme sensitivity of Zr-Mo alloys to trace amounts of oxygen is attributed to the presence of liquid molybdenum oxides in the surface scale, which leads to rapid oxygen transport. The stress sensitivity of the strain rate in these alloys decreases from 4.0 to 3.4 as the molybdenum concentration is increased from 0 to 6 pct, for both the yield and the steady-state regimes of flow. The alloy flow stress increases with molybdenum concentration approximately as C0.4, and it is apparent that the molybdenum atoms do not act as individual obstacles to flow, but are likely to lead to strengthening by indirect means.

Zirconium Technology—Twenty Years of Evolution

Abstract: Progress since 1955 in the technology of zirconium and its alloys is reviewed primarily from an applications viewpoint. A number of favorable features have been uncovered as their use broadened as well as problem areas which required accommodation in design or lifetime. Zirconium technology has reached its present stage of development as a result of world-wide contributions.

Review of Corrosion and Dimensional Behavior of Zircaloy under Water Reactor Conditions

Abstract: The corrosion of Zircaloy in water reactors is dependent on the water chemistry of the primary coolant and the fast neutron flux. Under standard pressurized water reactor (PWR) coolant conditions with hydrogen overpressure, resulting in a low oxygen content in the primary coolant, the fast neutron flux has no measurable influence on the corrosion rate. Therefore, the in-pile corrosion almost follows the out-of-pile behavior. However, in some cases deviations from this general behavior have been observed. Under oxygenated coolant conditions, typically for boiling water reactors (BWRs), the corrosion is enhanced by the fast neutron flux from the early beginning. Three different types of oxide have been observed. The dimensional behavior of Zircaloy under irradiation is governed by two processes: creep and growth. Creep is dependent on fast neutron flux and saturates with time. Total growth results from irradiation growth, anisotropic creep, and mechanical interference. Both in-pile creep and growth depend on the amount in recrystallization of the material.