Fuel rod cladding

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. 

Zirconium is used for fuel-rod cladding in water-cooled power reactors. 

The problem of hyperthermal corrosion resistance of structural materials was got over by development of preliminary protective coatings for the working steel surfaces. In particular, the most important structural units of circuit, e.g. fuel rod claddings and steam generator tubes, are covered by these coatings at the final stage of their manufacture. Additional barriers are also formed directly on the inner surfaces of liquid metal circuit under effect of the coolant in the early stage of the reactor operation. 

Double zirconium fuel rod cladding and silumin material Blling the clearance space between them. 

As a tesult of the recent incident at the Three Mile Island-2 (TMI) reactor there was concern that sufficient fuel rod cladding damage had occurred so that many of the fuel pellets were loose in the vessel. These pellets might arrange Aemselves in a configuration that would lead to recriticality unless sufficient soluble boron was in the system. The present calculations were undertaken to determine what boron concentration would be necessary to preclude recriticality under a number of assumed conditions. 

Fuel rod cladding is stainless steel alloy lTT-9, as well as much of the internal core structure [22]. HT-9 is favored in fast neutron applications for its neutron transparency, resistance to radiation damage, and thermal performance. 

Most LWR fuel rod cladding is made of Zircaloy (and its derivatives), which is an alloy of primarily zirconium and tin. Other alloying elements include niobium, iron, chromium, and nickel. Zircaloy was chosen because it has a very low cross section for thermal neutrons. Naturally occurring zirconium contains about l%-5% hafnium. The hafnium must be removed because it has a very high thermal neutron cross section and is often used in making control rods for reactors. The separation process used in the United States is a liquid-liquid extraction process. It is based on the difference in solubility of the metal thiocyanates in methyl isobutyl ketone. In Europe, a process known as extractive distillation is used to purify zirconium. This method employs a separation solvent that interacts differently with the zirconium and hafnium, causing their relative volatilities to change. This enables them to be separated by a normal distillation process. The separated zirconium is then alloyed with the required constituents. 

In addition to the radionuclides produced in the nuclear fuel, others are generated in the core structural materials such as the fuel rod cladding, spacers, fuel assembly end pieces, control rods, and core support materials. Here, the main production mechanism is neutron activation. A fraction of the radionuclides generated in this manner is also released to the primary coolant during steady-state operation of the plant or to the primary circuit in the course of a severe reactor accident. Therefore, the following discussion will also cover these radionuclides. 

In light water reactors, Zircaloy is commonly used as the fuel rod cladding material, a zirconium alloy with various metallic constituents. Pressurized water reactors use Zircaloy-4 (Zry-4), while in boiling water reactors Zircaloy-2 (Zry-2) is the preferred cladding material the compositions of both alloys are shown in Table 1.2. In German PWRs the mass of Zircaloy amounts to about 290kg/Mg HM (heavy metal), in BWRs to about 320 kg/Mg HM (including the fuel assembly channels). 

BWR fuel assemblies are surrounded by a fuel channel made of Zircaloy-2. The radionuclides generated here and their activity concentrations are on the same order of magnitude as in the fuel rod claddings. 

Some of the metallic fission products are plated out to a large extent onto the outer surfaces of the fuel rod claddings in the immediate vicinity of the defect this means that only small fractions of them are transported into the coolant. This is true in particular for tellurium, probably due to an electrochemical reaction on the Zircaloy surface resulting in the formation of the compound SnTe. The rather long-lived Te (halflife 76.3 h) decays there under the production and continuous release of its daughter to the coolant this mechanism is assumed to be the reason for the specific release behavior of this radionuclide, which is markedly different from that of the other iodine isotopes, both during constant load operation and during transients. 

In both PWRs and BWRs, corrosion of the primary circuit materials is an essential factor in the buildup of contamination layers on the surfaces of the pipes and the components. The materials used in BWRs which are in contact with the reactor water and, therefore, are potential sources of radionuclides are mainly stainless steels wear-resistant hardfacing alloys such as Stellite are also present in most of the plants. Zircaloy as the material of fuel rod claddings, spacers and fuel assembly casks need not be considered in this context, because of the extremely small release of activated constituents from this material. Due to differences in temperature and environment, the mechanisms of the corrosion process and the resulting metal release rates, which contribute to the input of corrosion products into the region of the reactor core, may show differences in different regions of the plant. Thus, corrosion of materials in the water-steam cycle exhibiting H2O phase transformations and considerable temperature differences will proceed differently than in the recirculation lines and the reactor water cleanup system, which are in contact with liquid water exclusively and show comparatively small variations in operating temperature. 

In addition to zirconium oxidation, the chromium content of the stainless steel internals of the reactor pressure vessel can also be oxidized by steam in this phase of the accident, resulting in the production of additional hydrogen. Since the rate of chromium oxidation is lower by a factor of 2 to 3 than that of zirconium under comparable conditions and since the temperatures of the core internals are distinctly lower than that of the fuel rod claddings, the contribution of this reaction to total hydrogen production during this phase of the accident can be neglected. 

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