Zirconium Alloys: Part One

Zirconium is a commercially available refractory metal with excellent corrosion resistance, good mechanical properties, very low thermal neutron cross section, and can be manufactured using standard fabrication techniques. The unique properties of zirconium made ideal cladding material for the U.S. Navy nuclear propulsion program in the 1950's.

The initial commercial nuclear power reactors used stainless steel to clad the uranium dioxide fuel due to cost, but by mid-1960 zirconium alloys were the principle cladding material due to the superior neutron economy and corrosion resistance.

These same zirconium alloys are available to designers of high level nuclear waste disposal containers as internal components or external cladding. Additional advantages of zirconium alloys for long term nuclear waste disposal include excellent radiation stability and 100% compatibility with existing Zircaloy fuel cladding to alleviate any concerns of galvanic corrosion.

No metal or alloy is resistant to corrosive attack in all chemical environments. Zirconium is no exception, but it does have excellent resistance to a wide variety of chemicals. Zirconium has outstanding resistance to hydrochloric acid, sulfuric acid, organic acids, and alkaline media such as sodium hydroxide. Its resistance to nitric acid is equaled only by the noble metals such as tantalum.

A tightly adherent and protective oxide film protects the metal-oxide interface to provide corrosion resistance. An additional benefit for zirconium alloys in long-term geological disposal options is the inert nature of zirconium oxide. Application of zirconium alloys alleviates the concern of nickel and chromium contamination in the ground water in severely corroded spent fuel containers.

The most common application areas for cast zirconium equipment are in hydrochloric acid, sulfuric acid, and hot organic acids. Zirconium shows excellent corrosion resistance to all concentrations of hydrochloric acid even at temperatures exceeding the normal boiling point. However, zirconium is not resistant to hydrochloric acid containing oxidizing species such as cupric chloride, ferric chloride, or wet chlorine. Zr 702C is resistant to sulfuric acid concentrations up to 70 percent and Zr 705C is resistant to concentrations up to 55 percent to the normal boiling point of sulfuric acid. Poor resistance is obtained with higher concentrations, even at room temperature.

Although Zr 702C possesses good tensile properties, it does have relatively low impact strength compared to most corrosion resistant alloys. However, with proper care zirconium equipment can provide excellent service. Zr 705C offers the user a higher impact strength and, more importantly, a higher pressure temperature limit which could eliminate the need for higher pressure class products.

Zirconium is superior to stainless steels, nickel alloys, and titanium in organic acids. This alloy is considered for these applications at high temperatures where its marked superiority results in a distinct economic advantage. Zirconium has poor resistance to concentrated sulfuric acid, hydrofluoric acid, concentrated phosphoric acid, ferric chloride, cupric chloride, wet chlorine, and other oxidizing chloride environments.

Zirconium is one of the higher priced alloys applied in the chemical process industry. It is therefore used only where service conditions necessitate its selection.

Initial cost of zirconium equipment should be compared to less expensive alternatives only after considering many factors such as the following:

• Zirconium often has far superior corrosion resistance relative to less expensive alternates resulting in greater expected service life
• Mechanical reliability is often far greater for an alloy such as zirconium as compared to some nonmetallic equipment designs
• The high cost of production downtime for routine maintenance and equipment failure may require the use of a more reliable, corrosion resistant alloy such as zirconium.

Zirconium alloys have been used in a number of water-cooled fission reactor types since the late 1950’s due to their excellent aqueous corrosion resistance, low thermal neutron absorption cross section and good mechanical properties.

By contrast, the fusion community has apparently judged Zr-alloys to be unsuitable for use as either structural support or for cooling channel material in future fusion power plants. Part of this disregard stems from the notion that Zr-alloys may suffer excessive hydrogen embrittlement arising from the several hydrogenic sources present in the fusion environment. Of concern also is the possible high level degradation of mechanical properties, as a result of the combined helium production and displacement damage in the hard neutron spectrum typical of a fusion first wall.

As mentioned above, zirconium alloys are used in water reactors because of their low capture cross-section for thermal neutrons and good mechanical and corrosion properties. Early in their application, hydrogen was identified as an embrittling agent. The source of the embrittlement was hydride precipitates that formed as platelets.

Usually, low ductility was found below about 150°C with impact testing or in tensile testing when the normal to the hydride plates was parallel to the tensile stress. Much effort was expended in keeping the hydrogen concentration low and ensuring that any hydrides were in a benign orientation.

Results from a few laboratory experiments hinted that zirconium alloys may also fracture by a time-dependent mechanism involving hydrogen, but the first practical confirmation of such a mechanism was the cracking of experimental fuel cladding made from Zr-2.5 Nb. Cracks were found in the heat affected zone in the weld between the cladding and its end-cap after several months of storage at room temperature. Hydrides were associated with the cracks and the process was called Delayed Hydride Cracking – DHC.

Cracking in the pressure tubes in RBMK reactors had a similar cause. These tubes have a length of about 8 m, an inside diameter of 80 mm and a wall thickness of 4 mm, and the Zr-2.5 Nb is partly recrystallized. The final stage of fabrication involved straightening that induced residual tensile stresses of about 350 MPa. Twenty tubes out of a population of 20,000 leaked because of cracking by DHC initiated on the outside surface. Most of the failures were in the first two years of operation of the Kursk and Chernobyl reactors.

These failures prompted much research on the phenomenology and mechanisms of DHC in zirconium alloys. Although no new cracks have been observed in CANDU or RBMK reactors during the past couple decades, as the reactors age the specter of DHC looms because of increasing hydrogen concentration from corrosion and potential mechanical damage to the surfaces of the pressure tubes. Thus constant vigilance is required.