Study of the velocity of the magnesium reduction process of zirconium production

An apparatus is described for studying the velocity of the magnesium reduction process of zirconium production by the floating-crucible method, in which the rise in level of the molten MgCl₂ + KCl, due to the increase in weight of the reaction crucible immersed in the melt, was measured by means of a Co⁶⁰ -level gauge. The pressure of the zirconium vapor in the reduction apparatus was measured by means of an ordinary pressure gauge, connected by a molten tin seal to the space within the reduction apparatus.Measurements showed that when the vaporizer was heated to 450 °C, the temperature of the zirconium chloride during magnesium reduction did not exceed 330 °C.The mean velocity of the magnesium process for the reduction of zirconium chloride increases with increase in the temperature of the reaction crucible from 760 to 850 °C more slowly than with increase in the temperature of the vaporizer from 460 to 490 °C. The deciding factors determining the velocity of the magnesium reduction process of zirconium production are the temperature of the vaporizer and the rate of sublimation and pressure of the zirconium chloride vapor, which are dependent on it.     

Interaction of the metallic components of melt in a reactor core with zirconium dioxide ceramic

Conclusions The results permit drawing the following conclusions: the penetration depth of the melt into the ceramic during the experiment was equal to approximately 0.12–0.2 mm in the case of a mixed melt of the steel and zirconium and approximately 0.35–0.4 mm in the case of the pure-zirconium melt; the crucible ceramic does not undergo erosion under the action of both types of melts; in the case of an interaction with the zirconium melt, a zone of softening forms in the ceramic on account of the reduction of zirconium dioxide to ZrO_0.35; zirconium and to a lesser degree iron in the melt are partially oxidized, primarily on account of diffusion transfer of oxygen from the ceramic into the melt, in the process of the interaction with the zirconium-dioxide based ceramic in an inert medium; very little iron is transferred from the melt into the ceramic; under the conditions of an inert medium the zirconium is the main corroding component of the melt; for a comparatively low content of zirconium in the melt steel + zirconium the zirconium dioxide based ceramic is quite highly resistant to the action of the melt; and, the zirconium dioxide based ceramic is highly resistant to the action of the melt under these conditions as compared with the refractory zirconium dioxide concrete [1, 2] and construction concrete [8]. Scientific-Research Center TIV Industrial Association "Institute of High Temperature," Russian Academy of Sciences. Translated from Atomnaya énergiya, Vol. 79, No. 6, pp. 454–458, December, 1995.     

Chlorination of UO2, PuO2 and rare earth oxides using ZrCl4 in LiCl-KCl eutectic melt

A new chlorination method using ZrCl₄ in a molten salt bath has been investigated for the pyrometallurgical reprocessing of nuclear fuels. ZrCl₄ has a high reactivity with oxygen but is not corrosive to refractory metals such as steel. Rare earth oxides (La₂O₃, CeO₂, Nd₂O₃ and Y₂O₃) and actinide oxides (UO₂ and PuO₂) were allowed to react with ZrCl₄ in a LiCl-KCl eutectic salt at 773 K to give a metal chloride solution and a precipitate of ZrO₂. An addition of zirconium metal as a reductant was effective in chlorinating the dioxides. When the oxides were in powder form, the reaction was observed to progress rapidly. Cyclic voltammetry provided a convenient way of establishing when the reaction was completed. It was demonstrated that the ZrCl₄ chlorination method, free from corrosive gas, was very simple and useful.     

高温熔融物与金属堆腔相互作用实验研究

为掌握船用反应堆严重事故工况下压力容器失效初期堆芯熔融物热冲击对金属堆腔的破坏效应,开展了堆芯熔融物与金属堆腔相互作用机理实验。根据相似准则设计缩比金属堆腔实验装置,利用已有高温熔融物实验平台制备2 700 ℃高温氧化锆熔融物,通过特制卸料机构将高温熔融物卸料到实验段,对热冲击下实验段温度和变形响应特性及主要影响因素进行了研究。实验结果表明,高温熔融物进入金属堆腔初期,热冲击导致的金属堆腔最高温度为601 ℃,最大塑性变形量为0.44 mm,高温熔融物未导致金属堆腔热失效及断裂失效,金属堆腔实验段能保持完整。由于船用反应堆金属堆腔材料、结构和外部冷却条件更有利于保持金属堆腔完整性,基于实验结果推断,严重事故下压力容器下封头失效初期热冲击导致金属堆腔失效的风险较低。     

Morphologies of uranium and uranium–zirconium electrodeposits

Uranium and uranium–zirconium electrodeposits produced in the Fuel Cycle Facility at Argonne National Laboratory-West were examined using standard metallurgical techniques. Substantial differences in the morphologies of the two types of deposits were observed. Samples from pure uranium deposits were comprised of chains of uranium crystals with a characteristic rhomboidal shape, while the morphologies of samples from deposits containing zirconium in excess of approximately 0.5 wt% showed more polycrystalline features. Zirconium was found to be present as a second, zirconium metal phase at or very near the uranium–zirconium dendrite surfaces. Higher collection efficiencies and total deposit weights were observed for the uranium–zirconium deposits; this performance increase is likely a result of better mechanical properties exhibited by the uranium–zirconium dendrite morphology.     

Corrosion study of a highly durable electrolyzer based on cold crucible technique for pyrochemical reprocessing of spent nuclear oxide fuel

The application of the cold crucible technique to a pyrochemical electrolyzer used in the oxide-electrowinning method, which is a method for the pyrochemical reprocessing of spent nuclear oxide fuel, is proposed as a means for improving corrosion resistance. The electrolyzer suffers from a severe corrosion environment consisting of molten salt and corrosive gas. In this study, corrosion tests for several metals in molten 2CsCl–NaCl at 923 K with purging chlorine gas were conducted under controlled material temperature conditions. The results revealed that the corrosion rates of several materials were significantly decreased by the material cooling effect. In particular, Hastelloy C-22 showed excellent corrosion resistance with a corrosion rate of just under 0.01 mm/y in both molten salt and vapor phases by controlling the material surface at 473 K. Finally, an engineering-scale crucible composed of Hastelloy C-22 was manufactured to demonstrate the basic function of the cold crucible. The cold crucible induction melting system with the new concept Hastelloy crucible showed good compatibility with respect to its heating and cooling performances.     

金属铈在NaCl-KCl和NaCl-KCl-CeCl3熔盐体系中的溶解行为

通过离子浓度测定、气体生成法、熔盐耗酸量测定、X射线衍射和差重分析法,综合评判了金属铈在NaCl-KCl和NaCl-KCl-CeCl₃熔盐体系中的溶解行为。结果表明,金属铈在熔盐中的损失存在物理溶解和化学反应,可生成氧化物、氮氧化物、氯氧化物等,同时还可与坩埚成分发生反应。金属铈及其化合物在熔盐中的溶解度很低,但会生成新相,新相中含有大量的铈。采用嵌套式坩埚并增加内部小坩埚高度,可有效降低金属铈的溶解损失量。     

Dissolution Behavior of Cerium in NaCl-KCl and NaCl-KCl-CeCl3 Molten Salt System

The dissolution behavior of cerium in NaCl-KCl and NaCl-KCl-CeCl₃ molten salt systems was evaluated by ion concentration determination, gas generation method, determination of acid consumption of molten salt, X-ray diffraction and differential gravimetry analysis method. The results show that the loss of cerium in molten salt involves physical dissolution and chemical reactions, which can produce oxides, nitrogen oxides, chlorine oxides and other forms. The cerium can also react with the composition of crucible. Cerium and its compounds have a low solubility in molten salts, but new phases are formed, which contain large amounts of cerium. The dissolution loss of cerium can be effectively reduced by using nested crucible and increasing the height of small crucible.     

Creep behavior of niobium-modified zirconium alloys

Zirconium (Zr) alloys remain as the main cladding materials in most water reactors. Historically, a series of Zircaloys were developed, and two versions, Zircaloy-2 and -4, are still employed in many reactors. The recent trend is to use the Nb-modified zirconium alloys as the Nb addition improves cladding performance in various ways, most significant being superior long term corrosion resistance. Hence, new alloys with Nb additions have recently been developed, such as Zirlo^™2 and M5^™3. Although it is known that creep properties improve, there have been very few data available to precisely evaluate the creep characteristics of new commercial alloys. However, the creep behavior of many Nb-modified zirconium alloys has been studied in several occasions. In this study, we have collected the creep data of these Nb-modified alloys from the open literature as well as our own study over a wide range of stresses and temperatures. The data have been compared with those of conventional Zr and Zircaloys to determine the exact role Nb plays. It has been argued that Nb-modified zirconium alloys would behave as Class-A alloys (stress exponent of 3) with the Nb atoms forming solute atmospheres around dislocations and thus, impeding dislocation glide under suitable conditions. On the other hand, zirconium and Zircaloys behave as Class-M alloys with a stress exponent of ?4, attesting to the dislocation climb-controlled deformation mode.     

Determination des conditions optimales de polissage electrolytique du zirconium α par mesure d'impedance d'anode

Optimal conditions for electrolytic polishing of metallic materials may be determined with accuracy from anodic impedance measurements, such as: composition of the bath, temperature, stirring and electric parameters. The method was applied successfully to α zirconium, usually prepared by chemical polishing due to the problems met with other types of polishing.     

Luminescence properties of zirconium oxide films

Various luminescence techniques have been applied in the characterisation of the zirconium oxide film formed on the metal matrix. Our investigation shows that some alloying elements in the zirconium alloys show characteristic emissions which could be used to study their distribution in the zirconium oxide. While the origin and detailed mechanism of luminescence in the zirconium oxide system is not well established, cathodoluminescence (CL) and thermoluminescence results from this study support the theory that oxygen vacancies in complexes with impurities are responsible for the intrinsic luminescence in the oxide system. The specific oxygen vacancy and impurity complex is reported in literature as the T-defect, having an energy of 2.2 eV. To aid in interpretation of the CL data, optical transmission properties of the oxides were also measured. The latter investigation showed the thickness of the oxide illuminated by the CL technique is greatly dependent on the optical transmission properties of the oxide, and it ranges from 20 μm for pure zirconium or low-alloyed zirconium oxides to less than 3 μm for oxides of alloys with more absorbing, or higher alloying element concentrations. The 20 μm depth of illumination is significantly deeper than the electron penetration depth, which suggests a secondary source of excitation, possibly characteristic X-rays emitted by the specimen. These combined properties of the oxide and technique can result in CL images showing structural contrast not easily seen during SEM or optical microscopy of oxidised surfaces.     

Importance of zirconium cross sections in calculating reactivity effects for inert matrix fuels

Zirconium is employed as the “inert” component in several different inert matrix fuel concepts currently under development (cermet, oxide solution, etc.). Relative to the situation in standard light water reactors, this implies a very significant increase in the zirconium inventory, with a correspondingly enhanced importance of the nuclear data for zirconium in calculating safety related parameters. The present paper discusses new numerical results for various reactivity effects. On the basis of the current findings, it is recommended that a high priority be assigned to both the reduction of uncertainties in the epithermal data for zirconium and to the improvement of methods for resonance self-shielding of these data.     

Behavior of plasma-sprayed zirconium dioxide ceramic under the thermochemical effect of components of the reactor core melt

Conclusion The investigations have shown that, even though the zirconium dioxide layer sprayed on the metal interacts considerably with molten oxides of other metals (Fe, Ti, Ni), forming a porous loose structure, the saturation zone even with air cooling (not to mention water cooling) of the cold surface does not reach the metal, i.e., a considerable part of the zirconium dioxide remains integral and capable of transmitting large heat fluxes with no substantial damage. That attests to a promising outlook for the construction of a protection refractory layer of both an external and in-reactor core catcher by plasma-spraying ceramic onto a metal substrate. NITs TIV OIVT RAN. Translated from Atomnaya énergiya, Vol. 85, No. 2, pp. 115–119, August, 1998.     

Mitigation of hydride embrittlement of zirconium by yttrium

Brittle hydrides of plate-shaped morphology are known to embrittle the host zirconium matrix. The embrittlement effect is a strong function of the aspect ratio of hydride plates and their major dimension, as thinner plates behave akin to cracks resulting in stress-concentration around their edges, especially when tensile load is acting normal to the broad face of the plates. The embrittlement of the host matrix is due to loss in load bearing area as a result of cracking of the hydride plates under load and severe localized deformation of the ligaments joining the hydride plates. In this work, mitigation of hydride embrittlement was attempted by exploiting the synergistic effect of yttrium addition to zirconium and microstructural modification of the Zr-Y alloy by quenching. This was expected to enable creation of a very high density of nucleation sites much stronger than those available otherwise and, thus, facilitate precipitation of much smaller hydrides that do not embrittle the host matrix. The results obtained in the present work on a dilute Zr-Y alloy do support this idea.     

Corrosion behavior of Hastelloy-N alloys in molten salt fluoride in Ar gas or in air

The effects of air on the corrosion of Hastelloy-N alloys in molten salt coolant containing fission product elements were investigated to determine the safety of structural materials in high-temperature reactors cooled with fluoride salt. Corrosion tests of Hastelloy-N in the molten fluoride salt FLiNaK in an alumina crucible and a graphite crucible under argon gas or air were performed at 773–923 K for 100 h. The depth of corrosive attack, as well as the extent of chromium and molybdenum depletion, increased with increasing temperature. The extent of Hastelloy-N corrosion in molten salt under air was significantly greater than under argon gas. The effect of adding the impurity cesium iodide to molten salt containing nuclear waste fuel on the corrosion behavior was negligible.     

Quantitative characterization of crystallographic textures in zirconium alloys

The techniques for determining inverse pole figures and direct pole figures of zirconium alloys by X-ray diffraction are summarized, and their advantages, disadvantages, and limitations are discussed. A critical review is made of the various parameters that have been used to quantify the texture in zirconium alloys. A new series of four quantitative texture numbers F_T, (SD)_T, F_A, and S, which are obtained from the direct pole figure, are proposed. Pole figures are determined for Zircaloy-2 tubing produced by three tubing manufacturers. The four texture numbers are calculated and are used to compare the textures of the three manufacturers and the through wall texture gradient of one manufacturer.     

Treatment of molten salt wastes by phosphate precipitation: removal of fission product elements after pyrochemical reprocessing of spent nuclear fuels in chloride melts

The removal of fission product elements from molten salt wastes arising from pyrochemical reprocessing of spent nuclear fuels has been investigated. The experiments were conducted in LiCl-KCl eutectic at 550 °C and NaCl-KCl equimolar mixture at 750 °C. The behavior of the following individual elements was investigated: Cs, Mg, Sr, Ba, lanthanides (La to Dy), Zr, Cr, Mo, Mn, Re (to simulate Tc), Fe, Ru, Ni, Cd, Bi and Te. Lithium and sodium phosphates were used as precipitants. The efficiency of the process and the composition of the solid phases formed depend on the melt composition. The distribution coefficients of these elements between chloride melts and precipitates were determined. Some volatile chlorides were produced and rhenium metal was formed by disproportionation. Lithium-free melts favor formation of double phosphates. Some experiments in melts containing several added fission product elements were also conducted to study possible co-precipitation reactions. Rare earth elements and zirconium can be removed from both the systems studied, but alkaline earth metal fission product elements (Sr and Ba) form precipitates only in NaCl-KCl based melts. Essentially the reverse behavior was found with magnesium. Some metals form oxide rather than phosphate precipitates and the behavior of certain elements is solvent dependent. Caesium cannot be removed completely from chloride melts by a phosphate precipitation technique.     

Coating Microspheres with Zirconium Carbide-Carbon Alloy by Iodide Process

Simultaneous deposition of carbon and zirconium from vapor produced by the reaction between methyl iodide vapor and zirconium sponge was studied with the application of a spouted bed constituted by a funnel carrying a charge of alumina microspheres, which were blown upward and held in dynamic suspension by a jet of the vapor and gases spouting from the funnel. The purpose of the experiment was to determine the conditions favorable for obtaining a coat of zirconium carbide-carbon alloy on the microspheres. Deposition of the vapor on the microspheres, leading to the formation of the carbon alloy coating, was found to take place at temperatures exceeding 1,100°C. The C/Zr ratio of the deposited coat was found to increase with deposition temperature. The hydrogen concentration in the spouting gas affected both the deposition yield and the chemical composition of the deposit. Repeated use of the sponge was found to impair its performance due to deactivation by premature deposition of carbon.     

Cluster Dynamics modelling of irradiation growth of zirconium single crystals

This paper aims at modelling irradiation growth of zirconium single crystals as a function of neutron fluence. The Cluster Dynamics approach is used, which makes it possible to describe the variation of irradiation microstructure (dislocation loops) with neutron fluence. From the irradiation microstructure, the strain can be calculated along the axes of the lattice structure. The model is applied to the growth of annealed zirconium single crystals at 553 K measured by Carpenter and Rogerson in 1981 and 1987. The model is found to fit the experimentally measured growth of Zr single crystals very nicely, even at large neutron fluence where the ‘breakaway growth’ occurs. This was made possible by considering in the model the growth of vacancy loops in the basal planes. This growth of vacancy loops in the basal planes could be modelled by taking into account that diffusion of self-interstitial atoms (SIA) is anisotropic and that there exist in the basal planes some nucleation sites for vacancy loops (iron clusters), the density of which is considered constant over time.     

Phosphate-hydroxyquinoline method for separation and volumetric determination of zirconium

The proposed method consists of a combination of the well-known phosphate method for separation of zirconium, and determination of zirconium as the hydroxyquinolate. The separation of zirconium hydroxyquinolate from an oxalate medium after solution of the phosphate preccipitate in oxalic acid has been used for the first time. The conditions of separation of zirconium from titanium and thorium in the phosphate precipitation, and from niobium and tantalum in precipitation of the hydroxyquinolate, have been studied. By this method it is possible to separate zirconium in practice from all accompanying elements (except hafnium), and to determine small amounts of zirconium (2–5 mg) by a volumetric method to an accuracy of ± 2–4%.