Zircaloy corrosion resistance

Niobium is also important in nonferrous metallurgy. Addition of niobium to tirconium reduces the corrosion resistance somewhat but increases the mechanical strength. Because niobium has a low thermal-neutron cross section, it can be alloyed with tirconium for use in the cladding of nuclear fuel rods. A Zr—l%Nb [11107-78-1] alloy has been used as primary cladding in the countries of the former USSR and in Canada. A Zr—2.5 wt % Nb alloy has been used to replace Zircaloy-2 as the cladding in Candu-PHW (pressurized hot water) reactors and has resulted in a 20% reduction in wall thickness of cladding (63) (see Nuclear reactors). 

Another characteristic that makes zirconium useful is the production of zircaloy, which does not absorb neutrons as does stainless steel in nuclear reactors. Thus, it is ideal to make nuclear fuel tubes and reactor containers. Zircaloy is the blend (alloy) of zirconium and any of several corrosion resistant metals. 

Except for the Zircaloy in the reactor core, the reactor internals are stainless steel or other common corrosion-resistant alloys. The reactor vessel is a pressure vessel with a single full-diameter removable head. The base material of the vessel is low alloy steel, which is clad on the... 

The can should be as thin as possible, consistent with satisfactory mechanical strength and corrosion resistance (Fig. 19.12(a)). To reduce the danger of hydride formation a protective oxide layer can be produced by autoclaving the tube before filling it with pellets. In case of UO2 pellets in zircaloy, the bonding material, e.g. graphite, is put onto the inner surface of the zircaloy tubes before the pellets are introduced. In case of stainless steel clad fast reactor fuel the production and assembly is similar, but the bonding is usually by sodium metal. 

Zircaloy clad oxide fuel elements can be stored for decades in storage pools with very little risk of leakage. Metal fuels, especially those canned in magnesium or aluminum alloys, are less resistant and should not be stored as such in this manner for a prolonged time. The corrosion resistance of aluminum or magnesium clad fuel can be improved by electrolytic treatment yielding a protective oxide layer. 

Cr-Fe-Zr is one of the ternary systems relevant to Zircaloys, which arc widely used as fuel cladding material in nuclear industry due to their low neutron-capture cross-section, high mechanical strength, high thermal conductivity and good corrosion resistance. In fliese alloys Fe and Cr, which are essentially insoluble in Zr at temperatures lower than about 600°C, are present in flie form of Zr(Fe,Cr)2 precipitates. This is one of the reasons why several studies have been performed on properties and characteristics of the Zr(Fe, Cr)2 phase. 

To help improve the corrosion resistance of Zircaloy, several new zirconium alloys have been developed, such as Zirlo (Zr-1.0% Nb-1.0% Sn-0.1% Fe). Notwithstanding the progress so far, materials reliability does have a significant effect on the economics of nuclear power plants, and there is considerable incentive to develop a full understanding of the mechanisms of corrosion of zirconium alloys in reactors and to develop alloys that are resistant to both irradiation and corrosion in reactors [18]. 

In practice, the strength and corrosion resistance of zirconium can be greatly improved by minor alloying additions. The principal alloys are zircaloy-2 and zircaloy-4. The former has 1.2-1.7wt % tin, with minor additions of iron, chromium, and nickel. Zircaloy-4 is similar, but without the nickel. Another important alloy is Zr-2 %Nb, which offers a further improvement in strength and thus permits the use of thinner tubes to give better neutron economy. 

Zr-2.5Nb is considered to be somewhat less resistant to corrosion than the zircaloys with exception. Nevertheless, Zr-2.5Nb is suitable for many applications, such as pressure tubes in the primary loops of some reactors. Furthermore, the corrosion resistance of Zr-2.5Nb can be substantially improved by heat treatments. " Also, Zr-2.5Nb is superior to zircaloys in steam at temperatures above 400°C. ... 

The nuclear fuel must be clad with a corrosion-resistant material to prevent the release of radioactive gases and fission products to the primary water. The fuel itself is usually uranium oxide, which is quite resistant to the primary water. The fuel is clad with austenitic stainless steel or Zircaloy-2 (R60802) or the extra-low nickel Zircaloy-4 (R60804). 

The corrosion -irradiation program for Zircaloy 2-clad, uranium-niobium core alloy (3 and 6% niobium) plates described in ANLi-5260 was extended to include the I-1/2 % nipbium-5%zirconium alloy. The plates are to be exposed in the Argonne Water Loop at MTR both in and out of neutron flux in contact with 500F water. Deliberate, (drilled hole) defects will then be put into the clad and the plates will be further corrosion tested out of flux to determine the effect of irradiation on corrosion resistance., ... 

Zirconium alloys, especially Zircaloy-2, have shown extremely good corrosion resistance relative to other materials. Fairly severe attack in a high-velocity wake region by a very abrasive thoria slurry has been noted. 

A wide variety of binary zirconium alloys was exposed in autoclaves and in in-pile loops to uranyl-sulfate solutions (see Section 5-5) however, only alloys of zirconium with niobium, palladium, or platinum showed greater corrosion resistance than Zircaloy-2. With the exception of the phase diagrams and a few mechanical property tests on low-niobium alloys [138], no information on any of these systems was available in the open literature. The few results for the mechanical property tests indicated very brittle alloys. 

The increase in fuel temperature from 250°C in the HRE-1 to 300°C was based upon the more favorable corrosion resistance of both stainless steel and Zircaloy-2 to dilute uranyl sulfate at the higher temperature and the possibility of improved thermal efficiencies. Also the temperature at which the two-liquid phase region appears is higher for the more dilute fuel, permitting this increase. 

Reactor-grade zirconium is essentially free of hafnium. Zircaloy(R) is an important alloy developed specifically for nuclear applications. Zirconium is exceptionally resistant to corrosion by many common acids and alkalis, by sea water, and by other agents. Alloyed with zinc, zirconium becomes magnetic at temperatures below 35oK. 

The fifth component is the stmcture, a material selected for weak absorption for neutrons, and having adequate strength and resistance to corrosion. In thermal reactors, uranium oxide pellets are held and supported by metal tubes, called the cladding. The cladding is composed of zirconium, in the form of an alloy called Zircaloy. Some early reactors used aluminum fast reactors use stainless steel. Additional hardware is required to hold the bundles of fuel rods within a fuel assembly and to support the assembhes that are inserted and removed from the reactor core. Stainless steel is commonly used for such hardware. If the reactor is operated at high temperature and pressure, a thick-walled steel reactor vessel is needed. 

The most important applications of zirconium involve its alloys, Zircaloy. The aUoy offers excellent mechanical and heat-transfer properties and great resistance to corrosion and chemical attack. This, in conjunction with the fact that zirconium has a low neutron absorption cross section, makes this ahoy a suitable choice as a construction material for thermal nuclear reactors and nuclear power plants. Other uses are as an ingredient of explosive mixtures, as getter in vacuum tubes, and in making flash bulb, flash powder (historical), and lamp filaments, in rayon spinnerets, and in surgical appliances. 

The zircaloy series of alloys was developed by the U.S. Navy Nuclear Propulsion Program for service in the core of water-cooled nuclear reactors [R3]. Compared with pure zirconium, these alloys have greater strength and better resistance to corrosion by water or steam. Zircaloy-4 was developed later than zircaloy-2 and became the preferred material, because the nickel in zircaloy-2 promoted the absorption of hydrogen, leading to reduction in ductility. 

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