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Zirconium (Zr) is a greyish white metal with unique physical and chemical properties that make it highly suitable for a wide range of industrial and scientific applications. Zirconium is the 20th most abundant element in the earth’s crust and commonly occurs as the mineral Zircon in silicate form and, less frequently, as the oxide mineral, Baddeleyite.
Zirconium metal and its alloys have many uses; among them in surgical instruments and medical implants, television screens, in the removal of residual gases from electronic vacuum tubes, catalytic converters and as a vital addition in aluminium alloys and steels. In the paper and packaging industries, zirconium compounds also make good surface coatings as they have excellent water resistance and strength.
Most of the zirconium chemicals and zirconium metal produced worldwide is obtained from zirconium silicate (ZrSiO4), or zircon, a natural mineral that is recovered from ancient mineral sands deposits.
After mining and production of a heavy mineral concentrate, zircon is separated, beneficiated and commercialised in the form of zircon sand and is either used directly in certain applications (foundry sands including investment casting) or processed for use in refractories, ceramic opacification, or to numerous zirconium compounds from which the metal can be produced.
The early development of zirconium metallurgy was essentially due to the nuclear power generating industry, and where zirconium alloys are now regarded as the proven structural material for nuclear fuel cladding in light water reactors. This is primarily because of the alloy’s unique combination of good corrosion resistance in the water chemistry at 300oC and low capture cross-section for thermal neutrons.
During the early 1950’s, the decision to use a zirconium alloy in nuclear fuel cladding was essentially due to Admiral Hyman Rickover of the US Navy. Charged with developing nuclear-propelled ships and submarines, Rickover realised that the reactor for use in vessels had to be compact and be able to operate when the ship was rolling or pitching, or at an angle when the submarine was diving or surfacing.
A pressurised water reactor (PWR) was envisaged, with the need for a nuclear fuel cladding material that would withstand corrosion at high temperatures and over long periods of time, maintain its integrity in an environment of intense radiation and would not absorb neutrons required for the nuclear reaction.
Low neutron absorption is vital to any structural material used in a nuclear reactor because large numbers of neutrons produced by the reaction must be free to interact simultaneously with all the nuclear fuel confined inside hundreds of fuel rods. This interaction sustains the necessary chain reaction throughout the reactor's core.
Zirconium was the metal of choice for this application as it absorbs relatively few of the neutrons produced in a fission reaction and because the metal is highly resistant to both heat and chemical corrosion.
Stainless steel, beryllium and aluminium were compared and considered unsuitable. Initial tests with zirconium showed that it absorbed neutrons needed for the fission process, due to the fact that the zirconium contained about 2% (by weight) of hafnium (Hf), which affected the high level of absorption cross-section for neutrons.
Zirconium and hafnium have to be separated for nuclear power purposes and tests at the Oak Ridge National Laboratory in Tennessee successfully achieved this with the result that zirconium in its purest form absorbed very few neutrons.
In this liquid-liquid extraction process, the mixed ZrHf-chloride is dissolved in hydrochloric acid. The Zr and Hf ions are complexed with ammonium-thio-cyanate and Hf is then extracted with methyl isobutyl ketone (MIBK) in a counter current liquid-liquid extraction system. The aqueous phase, containing Zr, is mixed with sulfuric acid to precipitate the Zr as hydroxide with the addition of ammonium hydroxide.
After filtering, the Zr-hydroxide is calcined to ZrO2 and Hf is removed from the MIBK with hydrochloric acid. For this separation process the carbo-chlorination has to be repeated to produce a Hf-free ZrCl4, which is finally reduced to pure zirconium.
As nuclear power generation became commercialised there was a clear need to decrease the production cost of zirconium, but the high-purity zirconium that proved to be such a good choice for the Shippingport Atomic Power Station (1958) in the USA was extremely expensive at the time – around $300 per pound – however, as the price of high-purity zirconium gradually decreased to around US$10 per pound by the 1970s, subsequent fuel loads for all light water reactors used variants of the zirconium alloy (zircaloy) cladding.
In the very early days it was found that, contrary to expectations, the corrosion resistance of the purer Zr material was lower than that of the impure. The addition of small amounts of iron and tin were found to improve corrosion resistance, and over the years the various PWR fuel suppliers began developing proprietary zirconium alloy families by varying the amounts of added tin, iron, niobium and chromium.
These changes, coupled with corresponding improvements to the alloy annealing treatments, enhanced the in-reactor dimensional stability of the materials and minimised corrosion while operating at higher fuel duties, thus improving the fuel cycle economics for the reactor’s owner.
As a strategic material, the zirconium market is commercially confidential and fluctuates with the supply and demand for various forms of nuclear-grade zirconium.
The fuel fabricators can take different approaches to the procurement of nuclear grade Zr metal:
However, given the nuclear power sector consumes only a small percentage of the market for zircon, it has a minimal impact on overall demand for the mineral.
There are three key safety functions that a nuclear reactor has to achieve, commonly known as the 3Cs:
The zirconium alloy cladding and grids that provide the structural materials for the nuclear fuel support all three key safety functions.