Zirconium-Niobium Alloys for Core Elements of Pressurized Water Reactors

  title={Zirconium-Niobium Alloys for Core Elements of Pressurized Water Reactors},
  author={A. V. Nikulina},
  journal={Metal Science and Heat Treatment},
  • A. Nikulina
  • Published 1 July 2003
  • Materials Science
  • Metal Science and Heat Treatment
The main characteristics of niobium-bearing zirconium alloys used for fabricating fuel element claddings of pressurized water reactors are considered. It is shown that the high corrosion and radiation resistance of zirconium parts is provided by the chemical composition, structure, and phase composition of the alloys. The Zr – Nb alloys developed in Russia provide reliable operation of fuel elements and fuel rod arrays in active reactors and serve as a basis for new modifications. 

Crystal Structure of the ZrO Phase at Zirconium/Zirconium Oxide Interfaces**

The combination of first-principles materials modeling and high-resolution electron microscopy is used to identify the structure of this sub-oxide phase of zirconium-based alloys, bringing us a step closer to developing strategies to mitigate aqueous oxidation in Zr alloys and prolong the operational lifetime of commercial fuel cladding alloys.

A Review of Development of Zirconium Alloys as a Fuel Cladding Material and its Oxidation Behavior at High-Temperature Steam

Currently, Zr-alloys are widely used in nuclear power reactors for fuel cladding and structural components. Many types of zr-based alloys were developed to overcome the challenges encountered in the

Fracture behavior of thin-walled Zircaloy fuel clad tubes of Indian pressurized heavy water reactor

For structural integrity analysis of thin walled tubular components such as fuel claddings used in nuclear reactors, knowledge of valid fracture mechanics parameters are required. As axially-cracked

Effect of aluminum, chromium, and iron doping on the heat resistance of zirconium

The influence of aluminum, chromium, and iron doping on the heat resistance of zirconium is studied. It is shown that a dense oxide film starts forming in the oxidation of the alloy containing 8 wt.%

Corrosion Cracking of Zirconium Cladding Tubes (A Review). I. Methods of Study and Mechanisms of Fracture

Methods of study and criteria of evaluation of stress corrosion cracking (SCC) of zirconium alloys are generalized as applied to cladding tubes of nuclear reactors. Mechanisms of SCC in zirconium

The Role of Β-Zr in Zr-2.5nb Alloys During Aqueous Corrosion: A Multi-Technique Study

Zr–2.5Nb alloy are used as pressure tubes in Canadian Deuterium Uranium (CANDU) nuclear reactors, and the typical starting microstructure consists of α-Zr grains elongated in both transverse and

Study on hydride precipitation and the induced embrittlement behaviours in Zircaloy-4

The zirconium alloy, Zircaloy-4 (Zr-4), has extensive applications in the nuclear industry as core structural components and fuel claddings in the nuclear reactors. The mechanical performance of

The Pressure Compaction of Zr-Nb Powder Mixtures and Selected Properties of Sintered and KOBO-Extruded Zr-xNb Materials

A very favorable effect of niobium on the increase in corrosion resistance was observed, compared to the material obtained from the powder without the addition of niOBium, for materials without the additional annealing process.



Zirconium Alloy E635 as a Material for Fuel Rod Cladding and Other Components of VVER and RBMK Cores

Data are given on Zr alloy E635 (Zr-1.2Sn-1Nb-0.4Fe), developed in Russia as a fuel rod cladding and other component material for use in cores of VVER and RBMK types. The alloy is much superior to

Evolution of Microstructure in Zirconium Alloys During Irradiation

X-ray diffraction (XRD) and transmission electron microscopy (TEM) have been used to characterize microstructural and microchemical changes produced by neutron irradiation in zirconium and zirconium

Influence of Neutron Irradiation on Dislocation Structure and Phase Composition of Zr-Base Alloys

Studied were evolution of dislocation structure, phase, and element composition of binary alloys Zr-1Nb and Zr-2.5Nb and multicomponent alloys Zr-1Nb-1.2Sn-0.4Fe and Zr-1.25n-0.4Fe under neutron

Irradiation induced growth and microstructure evolution of Zr-1.2Sn-1Nb-0.4Fe under neutron irradiation to high doses

Zirconium alloy components subjected to long-term operation and high doses in thermal reactors need to be highly irradiation resistant to provide integrity of components, primarily, their geometrical