Zircaloy-2, corrosion

As was mentioned above, adjusting the coolant pH to the range which is required in modified or even in elevated pH chemistry by addition of a sufficiently high concentration of LiOH poses some problems in the first months of a fuel cycle. The LiOH concentrations which would be needed are on the order > 5 ppm and may adversely affect stress corrosion of Inconel 600, in particular in the U-bends, as well as Zircaloy corrosion however, according to Roesmer (1989) these effects seem to be of minor significance. 

Zircaloys are preferred because they do not "poison" the nuclear reaction as much as do the stainless steels. The Zircaloys perform well in the 285 to 315°C temperature of the primary water for the two to three-year expected life of a nuclear core. They develop a shiny, adherent black oxide film, which is protective and has excellent heat transfer properties. However, after long exposure times (two to four years), or shorter times at higher temperatures, for example, 40 to 120 days at 360 to 400°C, Zircaloys corrosion rate increases and a white, relatively nonadherent insulating film develops. 

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. 

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). 

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. 

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. 

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. 

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... 

In PWRs, the fuel is U02, enriched typically to 3.3% 235U while for BWRs, the fuel is U02, enriched to 2.6%. (Natural uranium is 0.72% 235U). The fuel elements are clad in Zircaloy, a zirconium alloy that includes tin, iron, chromium, and nickel that prevents fission product release and protects them against corrosion by the coolant. The control rod material in BWRs is B4C, while PWRs have Ag-In-Cd or Hf control materials. 

Since operation of TRU began, numerous process changes have been made to counteract problems caused by equipment corrosion, to satisfy new processing requirements, and to utilize improved processes. In initial operations at TRU, unexpected corrosion of the Zircaloy-2 equipment occurred in high specific-activity chloride solutions, and the dissolved zirconium caused operating problems that made some of the flowsheets for the processing steps either difficult to operate or totally unusable. Because of these and other problems, several new flowsheets were devised or adopted, tested, and scaled to pi ant-size equipment. 

A second solvent extraction process (Pharex) was developed to partition the transcurium actinides from the americium and curium in the Tramex product ( 3) The Pharex process utilized 2-ethylhexyl phenylphosphonic acid as the extractant for the transcurium actinides. During early operations/ the selectivity of the Pharex extractant was found to be severely reduced by the presence of zirconium ions, which were introduced into the process solutions by corrosion of Zircaloy-2 equipment in TRU. At zirconium concentrations above 10 ppm, the achievable separation began to be diminished and, at 100 ppm, a practical separation could not be made (4). Thus, a replacement for the Pharex process was needed, and the LiCl AIX process was the most immediate alternative ... 

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. 

Electrolytic dissolution in nitric acid has been used at the Savannah River [B22] and Idaho Qiemical Processing plants [AlO, All] to dissolve a wide variety of fuels and cladding materials, including uranium alloys, stainless steel, aluminum, zircaloy, and nichrome. The electrolytic dissolver developed by du Pont [B22], pictured in Fig. 10.4, uses niobium anodes and cathodes, with the former coated with 0.25 mm of platinum to prevent anodic corrosion. Metallic fuel to be dissolved is held in an alundum insulating frame supported by a niobium basket placed between anode and cathode and electrically insulated from them. Fuel surfaces facing the cathode undergo anodic dissolution in a reaction such as... 

The major use for zirconium is in the nuclear industry. Zirconium alloys (zircaloys) are used extensively as a cladding for nuclear (uranium oxide) fuel rods in water cooled reactors. Zircaloys were favoured over stainless steel cladding because they had a considerably lower neutron cross-section, appropriate thermal conductivity and both corrosion and mechanical resistance. As indicated, hafnium is an impurity in nearly all zirconium ores. Hafnium, however, has a much higher neutron cross-section than zirconium and, as such, the two elements must be separated prior to using zirconium in fuel rod cladding. For many years the separation was very difficult due to the chemical similarity of the two elements. Zirconium hydride is used as a moderator in nuclear reactors. 

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. 

During operation, a slow corrosion of the can is unavoidable. As long as the corrosion products stick to the surface, corrosion rates drop with time. For zircaloy clad fuel in water cooled reactors the corrosion rate follows a parabolic equation (in the normal operating temperature range)... 

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. 

Figure 19. Corrosion of beta-quenched Zircaloy-2 in water and steam ( A)...

Figure 20. Corrosion weight gain of Zircaloy-2 in water at SSO C (15)...

WAPD-MRP-107, "Pressurized Water Reactor (PWR) Project Technical Progress Report October 24, 1963-January 23, 1964," Westinghouse Atomic Power Div. Griggs, B., Maffei, H. P., and Shannon, D. W., HW-67818, "Multiple Rate Transitions in the Aqueous Corrosion of Zircaloy," Hanford Laboratories, General Electric Company, December 20, 1960. 

Uranium dioxide has a number of properties that make it suitable for a fuel. The crystal structure is the fluorite (CaF2) type, similar to that of calcia-stabilised zirconia, and is stable to temperatures in excess of 2000 °C. Because it is a ceramic oxide, the material is refractory, chemically inert and resistant to corrosion Enrichment does not change these features. The oxide powder is pressed into pellets and sintered to a density of about 95 % maximum by traditional ceramic processing technology but is carried out in conditions that minimise risks from radiation effects. The pellets are contained in zirconium alloy (zircaloy) containers, which are then introduced into the reactor. The moderator, which... 

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