Some properties of alloys of zirconium with niobium

A diagram of state is given for the system zirconium-nioblum, as constructed from experimental data obtained by the writers. Results of tension tests conducted at room temperature on the alloys of zirconium with niobium are described, and an estimate of their high-temperature strength is given on the basis of hardness measurements at temperatures up to 750 °C, and also an estimate of the change of hardness of the alloys resulting from low-temperature anneals (aging). The data obtained are interpreted on the basis of the diagram of state. The high-temperature durability is determined for alloys oxidized in air at 570 and 650 °C.     

The normal elastic modulus of alloys of zirconium with niobium

Results are described of measurements of the normal elastic modulus of alloys of zirconium with niobium in vacuum at temperatures up to 950 °C, and also at room temperature after various heat treatments.     

Zirconium alloys matrix as innovative material for composite fuel

A novel class of zirconium alloys is suggested as fuel matrix. They are “deep” ternary or quaternary eutectics having relatively low melting point i.e. from 963 to 1133 K in comparison with pure zirconium and intended for use as a matrix of dispersion high uranium content fuel (CERMET and METMET) particularly for thermal reactors. For fast reactors and MA burning Zr–Ti based alloys are proposed that have resistant metallurgical bonds between fuel and steel cladding. Investigations have been carried out on the structure and properties of the alloys as well as the specific technologies of their fabrication, in particular via induction furnace melting. The alloys may be also produced in the amorphous state as granules and strips. It is shown that thanks to their capillary properties they might be applied in brazing dissimilar materials. Based on novel zirconium matrix alloys high uranium content fuel compositions were produced. They have high thermal conductivity and compatibility as well as 25–50% higher uranium content than for VVER and PWR fuels.Fuel pins are fabricated by capillary impregnation method. The use of dispersion fuel with novel zirconium matrix alloys improves neutronics characteristics of reactor cores and might lead to extension of burn-up, low operating temperatures and serviceability under transient conditions.     

CAP1400燃料组件用新锆合金研究

在Zr-Sn-Nb系合金的基础上添加微量合金元素Ge和Si等,采用真空电弧熔炼,制备了多种新锆合金。使用透射电子显微镜(Transmission electron microscope,TEM)对合金基体进行显微组织分析,分别通过堆外高压釜腐蚀试验、定氢分析仪和万能材料试验机对合金的腐蚀、吸氢和拉伸性能进行评估。结果表明,常规工艺处理后,SZA-4和SZA-6合金均发生了完全再结晶,第二相细小、均匀弥散分布在晶粒内和晶界上;SZA-4和SZA-6合金在三种水化学条件下均具有优良的耐腐蚀性能,SZA-6合金的耐腐蚀性能优于参考合金,SZA-4合金的耐腐蚀性能略优于SZA-6合金;SZA-6合金的吸氢性能略优于SZA-4合金;两种合金的拉伸性能满足设计要求。基于SZA-4和SZA-6合金优良的耐腐蚀、吸氢和力学性能,未来将有望用于CAP1400自主化燃料组件。     

Aging effects on microstructural and mechanical properties of select refractory metal alloys for space-reactor applications

Refractory alloys based on niobium, tantalum and molybdenum are potential candidate materials for structural applications in proposed space nuclear reactors. Long-term microstructural stability is a requirement of these materials for their use in this type of creep dominated application. Early work on refractory metal alloys has shown aging embrittlement occurring for some niobium and tantalum-base alloys at temperatures near 40% of their melting temperatures in either the base metal or in weldments. Other work has suggested microstructural instabilities during long-term creep testing, leading to decreased creep performance. This paper examines the effect of aging 1100 h at 1098, 1248 and 1398 K on the microstructural and mechanical properties of two niobium (Nb-1Zr and FS-85), tantalum (T-111 and ASTAR-811C) and molybdenum (Mo-41Re and Mo-47.5Re) base alloys. Changes in material properties are examined through mechanical tensile testing coupled with electrical resistivity changes and microstructural examination through optical and electron microscopy analysis.     

The physical metallurgy of high strength zirconium alloys

The requirements of materials for use in the core section of unenriched uranium type nuclear reactors are, to a large extent, fulfilled by zirconium alloys. However, the operating temperature of such reactors is circumscribed by the inability of the presently available alloys to maintain their properties at temperatures exceeding 350–400° C. This paper is concerned with the selection of alloying elements for the improvement of the high temperature strength of zirconium alloys and reports the results obtained from a study of the phase transformations in dilute Zr-Cr alloys.     

Aspects of the metallurgy of uranium and constructional metals

On the basis of foreign material presented at the Second International Conference on the Peaceful Uses of Atomic Energy (Geneva, 1958), we examine aspects of uranium production (reductive melting, casting, pressure treatment, and powder metallurgy); zirconium metallurgy (new data on the technological scheme, the nature of sponge and the dimensions of ingots); the production of the new constructional metals, niobium and vanadium (comparison of various methods for their preparation; metallothermal, vacuum carbo-thermal, electrolytic refining and the quality of production); the production of thin-walled beryllium tubes for fuel element sheaths; the use of zirconium hydride as a moderator and of zirconium-uranium-hydrogen alloys for fuel element cores.     

Microprobe analysis of neutron irradiated and autoclaved zirconium niobium claddings using synchrotron-based hard X-ray imaging and spectroscopy

Two zirconium alloys (Zr-2.5%Nb) - one oxidized in a pressurized water reactor, the other oxidized in autoclave and used as reference - are analyzed by combining synchrotron-based scanning transmission and fluorescence X-ray microscopy and micro-X-ray absorption spectroscopy (micro-XAS). Two-dimensional zirconium distribution maps recorded on the neutron irradiated and the non-irradiated autoclaved Zr-2.5%Nb alloys clearly allow the localization of the oxide and the metal parts of the interface with a micrometer spatial resolution. Micro-XAS investigations make possible the determination of the speciation of zirconium and niobium both in the oxide and the metal parts of the interface for the irradiated and non-irradiated samples. The coordination environment and/or the valency of zirconium and niobium in the metal and the oxide parts of the interface have been determined for both materials, and interpreted on the basis of comparison with metal and oxide reference compounds.     

Thermal stability and brazing characteristics of Zr0.7−xMxBe0.3 (M=Ti or Nb) ternary amorphous filler metals

The effects of Ti or Nb substitution on the thermal stability and brazing characteristics of Zr_0.7−xM_xBe_0.3 (M=Ti or Nb) ternary amorphous alloys were investigated in order to improve properties of Zr–Be binary amorphous alloy as a new filler metal for joining zirconium alloy. The Zr_0.7−xM_xBe_0.3 (M=Ti or Nb; 0x0.1) ternary amorphous alloys were produced by melt-spinning method. In the selected compositional range, the thermal stability of Zr_0.7−xTi_xBe_0.3 and Zr_0.7−xNb_xBe_0.3 amorphous alloys are improved by the substitution of titanium or niobium for zirconium. As the Ti and Nb content increases, the crystallization temperatures increase from 610°C to 717°C and 610°C to 678°C, respectively. These amorphous alloys were put into practical use in joining bearing pads on zircaloy cladding sheath. Using Zr–Ti–Be amorphous alloys as filler metals, smooth interface and spherical primary particles (proeutectic phase) appear in the brazed layer, which is the similar microstructure of using Zr_0.7Be_0.3 binary amorphous alloys. In the case of Zr–Nb–Be amorphous alloys, Ni-precipitated Zr phase that may cause some degradation in ductility and corrosion-resistance is formed at both sides of the brazed layer.     

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.     

Anisotropy of the elastic modulus and thermal-expansion coefficient of textured zirconium alloys N-1 and N-2.5

Conclusions The above calculations and measurements permit the conclusion that the calculation method employed is suitable for the prediction of anisotropy of the elastic and thermal properties of textured parts of alloys N-1 and N-2.5. All that is necessary here is to know the distribution of the basis-plane orientations and the properties of the -zirconium single crystal at the corresponding temperature. Alloying zirconium with niobium within limits of 1–2.5% has no significant influence on the elastic constants and thermal-expansion coefficients of the alloys. The elastic constants and thermal-expansion coefficients of alloys N-1 and N-2.5 may be recommended for the calculation of the stress-strain state of parts.Translated from Atomnaya Énergiya, Vol. 68, No. 2, pp. 98–101, February, 1990.     

The continuous cooling transformation behaviour of zirconium-niobium-oxygen alloys

The continuous cooling transformation behaviour of zirconium alloys containing up to 5 at% niobium and 3 at% (5300 ppm) oxygen has been examined using thermal analysis and metallographic techniques. Two types of reactions were found; nucleation and growth reactions which showed the familiar C-curve kinetics, and the athermal martensite reaction. The nucleation and growth reactions were identified as the formation of α-zirconium at the β-grain boundaries and the transformation βzr → αZr+βNb throughout the grains. Increasing niobium content lowered the temperature and increased the time required to obtain the thermal arrests. Increasing oxygen raised the temperature of both reactions and caused grain boundary nucleation to occur sooner and nucleation within the grains to occur later. The martensite start temperature was not affected by oxygen but decreased linearly with niobium content.     

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

Mechanical and thermo-physical properties of refractory metal alloys and mechanically alloyed (MA)-oxide dispersion strengthened (ODS) steels are reviewed and their potential for use in space nuclear reactors is examined. Preferable refractory alloys for use in liquid metal and gas-cooled space reactors include Nb-1%Zr, PWC-11, Mo-TZM, Mo-xRe where x varies from 7% to 44.5%, T-111 and ASTAR-811C. These alloys are heavy, difficult to fabricate, and are not readily available. The advantages of the MA-ODS alloys are: (a) their strength at high temperatures (>1000 K), which decreases slower with temperature than those of niobium and molybdenum alloys; (b) relatively lightweight and less expensive; (c) low swelling and no embrittlement with exposure to high-energy neutrons (>0.1 MeV) up to 10²⁷ n/m²; and (d) high resistance to oxidation and nitration. The few data available on compatibility of MA-ODS alloys with alkali liquid metals up to 1100 K are encouraging, however, additional tests at typical temperatures (1000-1400 K) in space nuclear reactors are needed. The anisotropy of MA-ODS alloys when cold worked, and particularly rolled into tubes, should not hinder their use in space nuclear power systems, in which operation pressure is either near atmospheric or as high as 2 MPa, but joints weldability is an issue.     

Effect of alloying elements on proclivity of zirconium alloys for nodular corrosion

An approach to assessing the effect of alloying elements on the proclivity of zirconium alloys for nodular corrosion is described. The approach is based on an analysis of the stability of the corrosion front. The results of a simplified analysis for iron and nickel additives to Zircaloy-2 and -4 are in agreement with the experimental data. The approach described can also be used for zirconium-niobium alloys, but this requires taking account of the mutual effect of the atoms of niobium and other alloying substances. For detailed quantitative assessments of the effect of alloying elements on the proclivity of zirconium alloys for nodular corrosion, the stability analysis must be expanded and including the use of self-consistent numerical models which take account of radiation effects and oxygen transfer through the oxide film. Translated from Atomnaya énergiya, Vol. 106, No. 2, pp. 94–99, February, 2009.     

Determination of the solubility of metals in lithium

The solubility of uranium, zirconium, iron, nickel, titanium, molybdenum, niobium and beryllium in lithium at temperatures of 700–1000 ° C was determined to assess the stability of metals in lithium and establish the mechanism of corrosion. It was found that nickel and beryllium have a high solubility (of the order of 1%), iron, zirconium, titanium and uranium are slightly soluble (from hundredths to thousands of one percent) and niobium and molybdenum have a very low solubility (less than l^–4%). Crucibles of the lithium to be tested were filled in a special still with distilled lithium and hermetically sealed in a container in a medium of argon. The solubility of the metal to be tested was determined by chemical analysis of rapidly cooled lithium fusions after they had been kept for 50–100 hours in the container at a predetermined temperature. The presence of isothermal transfer of aluminum, beryllium, zirconium and silicon via lithium to steel and iron was discovered. Under these conditions maximum solubility of the metal in lithium was reached far more slowly than in the absence of transfer. Lithium can be purified by getters — uranium and zirconium — slightly soluble in lithium.     

The Mechanism of Localized In-pile Corrosion of Zirconium Alloys

The role of β-particles for the localized in-pile corrosion of components made of zirconium alloys, such as fuel channels closely facing stainless steel components, has been discussed and the distribution of β-particles near fuel channels has been evaluated using a multi-layer calculation method.

It is shown that electric charges which come from β-particles produced mainly by the Compton effect, but not by β-emitting radionuclides, make electric fields in and on the zirconium oxide surfaces in the reactor by a model analogous to an electric circuit, and that these electric fields induced by β-particles could be essential for the understanding of localized in-pile corrosion of zirconium alloys. The mechanism of the localized in-pile corrosion of zirconium alloys adjacent to other welded alloys is also discussed as an enhanced corrosion by electric potentials using the proposed electric circuit model in detail.     

Mechanical properties and corrosion resistance of hafnium-zirconium alloys in a steam-water medium

The article describes the results of studies of the mechanical properties and corrosion resistance of hafnium containing 3-50% zirconium. Hafnium containing up to 50% zirconium is highly resistant to corrosion in water at 350° C. The mechanical properties of hafnium are redlmed slightly'when the zirconium content is increased from 3 to 25%. No appreciable deterioration in the mechanical properties of hafnium containing 3% zirconium is produced by prolonged exposure to a steam-water medium at 350-400° C.Translated from Atomnaya Énergiya, Vol. 14, No. 3, pp. 290–295, March, 1963     

Impurities and their effect on the structure and properties of zirconium parts in nuclear reactors

The results of investigations of the effect of impurities, which are due to the method used to produce zirconium, on the quality, structure, and properties of parts made of zirconium alloys are presented. Most impurities have a negative effect on the characteristics of the zirconium parts. It is shown that trace impurities play a large role in the formation of the microstructure and properties of the parts. Ways to decrease the concentration of impurities and to achieve uniformity of their distribution in parts are examined. Translated from Atomnaya énergiya, Vol. 105, No. 5, pp. 258–266, November, 2008.     

Generating material strength standards of aluminum alloys for research reactors I. Yield strength values Sy and tensile strength values Su

Aluminum alloys are frequently used as structural materials for research reactors. The material strength standards, however, such as the yield strength values (S_y), the tensile strength values (S_u) and the design fatigue curve—which are needed to use aluminum alloys as structural materials in “design by analysis”—for those materials have not been determined yet. Hence, a series of material tests was performed and the results were statistically analyzed with the aim of generating these material strength standards. This paper, the first in a series on material strength standards of aluminum alloys, describes the aspects of the tensile properties of the standards. The draft standards were compared with MITI no. 501 as well as with the ASME codes, and the trend of the available data also was examined. It was revealed that the draft proposal could be adopted as the material strength standards, and that the values of the draft standards at and above 150°C for A6061-T6 and A6063-T6 could be applied only to the reactor operating conditions III and IV. Also the draft standards have already been adopted in the Science and Technology Agency regulatory guide (standards for structural design of nuclear research plants).     

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.