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A Paper on Zircaloy

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Zirconium metallurgy has been developed essentially due to the nuclear industry. They are used as a structural material in nuclear industry mainly because of good corrosion resistance in water at high temperatures, resistance to radiation damage, optimum mechanical properties and low cross-sectional absorption of thermal neutrons. Four elements satisfy the last two conditions but Zirconium is the only sufficient choice for core components of nuclear reactors. The other elements such as Beryllium (Brittle and chemically toxic), Magnesium (chemically reactive and cannot be used in water cooled reactors), Aluminum (low melting temperature hence only used in research reactors) are not chosen. Important Characteristics of Zirconium-Nuclear Structural Material
1) Low thermal neutron absorption cross section (0.185 barns),
2) Allotropy, the high temperature body centered cubic (β) phase transforming into the hexagonal close packed phase (α) at 1135 K,
3) Anisotropic thermal and mechanical properties leading to unequal thermal expansions along different crystallographic directions and formation of strong crystallographic texture during mechanical working
4) Hexagonal close packed structure of the a phase with a c/a ratio of 1.593, which is less than ideal, making the prismatic slip on {I 0 1 O} planes most predominant,
5) High reactivity with oxygen, nitrogen and carbon and high solubility of these interstitial elements in the α phase, necessitating special care during melting and fabrication processes,
6) Low solubility of hydrogen in the α-phase 0.034 ppm at 293 K, excess hydrogen, if present, causing hydride precipitation which may result in serious embrittlement.
Development of Zircaloy:
Initially while trying to manufacture a material suitable for nuclear submarines, availability of metal at sufficient quantities and at reasonable costs were the primary problems. Zirconium also was not a choice in the beginning to manufacture. Once Kaufman at MIT and Pomerance at Oak-Ridge successfully separated hafnium and zirconium and proved that pure zirconium only absorbs few neutrons, focus was shifted on to it.
As there was no technical specification laid out, U.S.Navy who first used zirconium as the reference structural material adopted crystal bar process for the production of high-purity zirconium. This process is a refining rather than a metal reduction. Although they achieved a crystal zirconium it did not last when tested in high temperature water autoclaves due to corrosion.
In the next stage of development, the decision was to use a pure grade of sponge zirconium as raw material-Kroll process. They achieved an improved-yield and a corrosion resistant product. People stopped purifying it further once they observed lower performance of purer product compared to the impure one. So in later stage of the development focus shifted towards identification of those elements that improved corrosion resistance. Tin was the one that was first experimented with. Initially the level of addition was as high as 5wt% but later on it was reduced to 2.5wt% in order to achieve a good balance between corrosion resistance, strength and fabricability. They named this alloy as Zircaloy-1. Later on a melter in Bettis fabrication shops accidentally melted Zircaloy-1 ingot that is contaminated with stainless steel and the resultant material proved to have a better corrosion resistance. While coming up with alloy composition iron content was nominally set at .15% mainly keeping in mind the range of iron present in the then available Kroll sponge.
Although, Zircaloy-2 was good and performing satisfactorily, as the reactor life time kept on increasing, people were skeptical about the deleterious effect tin would have. Thus in the next development, tin was limited to .25% and Cr and Ni were totally removed. But during corrosion testing, zircaloy -3 displayed a network of fine white corrosion indications. They were later identified as stingers of Fe-Cr intermetallic compounds. This resulted in lower mechanical properties of the alloy.
Also, at that point of time hydrogen effect on mechanical properties of Zircaloy became evident. This was partly due to the notched bar impact testing initiation. Low impact values came on samples that showed the presence of crystallographically-oriented zirconium hydride platelets. When Zircaloy-2 was coated with nickel hydrogen absorption increased drastically. Thus a nickel free Zircaloy 2 was developed with an increase in iron content to improve corrosion resistance. This came to be known as Zircaloy-4. Zircaloy-4 was developed to reduce the hydrogen embrittlement. Also tin content has been reduced, this nominal level helped counteract the deleterious effects of nitrogen. As you can see the table below, the development kept on happening to meet the demands of new reactor cores and the increase in burn-up rates. The table below gives a brief idea of the development of Zircaloy. | | Kroll Zr | Sn alloying addition to counteract detrimental effect of ‘N’ impurity on corrosion | Zircaloy –1 (Sn-2.5) | Corrosion Resistance not satisfactory, Accelerated corrosion after breakaway or transition time. | Zircaloy–2 (Sn-1.5,Fe,Cr&Ni-0.1) | Sn reduced, Accidental addition of SS (Fe, Cr, Ni),Hydrogen pick-up increases due to Ni | Zircaloy–3, (Sn&Fe.25) | Less Sn –Stringer Corrosion | Zircaloy-4,(Sn-1.5,Fe-0.22,Cr-0.1) | Hydrogen pick-up rate decreases (~1/3) | Optimized Zircaloy-4, (Sn-1.5, Fe-0.2, 0.1Cr,Si-(50-120)ppm, C< 100 ppm) | Suitable for High Burn-up PWR Clad |
Composition of Zircaloy is mentioned as a mean composition in weight%
Manufacturing process: Fuel failure in the reactor core most of the times is because of cladding that has failed, thus leading to a tightening of cladding tube specifications. Generally, cladding defect is the last step in a series of events which was initiated, mostly because of inadequate fuel design or fabrication practice, or even plant operation or conditions. Some of the potential defect sources are: international corrosion or hydriding; mechanical interaction between fuel and cladding; cladding collapse after fuel densification; stress corrosion; fretting; strain fatigue; brittle end cap welds; crud deposits; and external corrosion.
In Zircaloy the alpha phase is stable below 7900 C, the beta phase is stable above 9500 C whereas a two phase region the alpha + beta occurs between 7900 C and 9500 C. In alpha phase Zirconium atoms are arranged in HCP (hexagonal close packed) lattice and in beta phase in BCC (body centered cubic) lattice. Generally, beta quenching is done during the manufacturing of zircaloy to obtain improved corrosion resistance.
It has been common practice to manufacture cladding and construction tubes of Zircaloy-2 and Zircaloy-4 by a process which includes: hot working of an ingot into a solid billet, heating said billet to the beta phase range followed by quenching, so-called beta-quenching. Then, it is machined to obtain a hollow extrusion tube billet, Later on the billet is extruded as a hollow tube at high temperature within the alpha phase range followed by reduction thereof in several steps by cold rolling to substantially final dimension. The said tube is subjected to a recrystallization annealing in vacuum before each cold rolling step. The cold worked tube which substantially has a final dimension is then subjected to vacuum annealing in the alpha phase range or in case of a construction tube is subjected to an additional cold deformation such as rolling or drawing before the final vacuum annealing. The final annealing in vacuum is performed consisting of a stress-relief, a partial recrystallization or a complete recrystallization. The kind of final vacuum annealing is selected based upon the specification by the fuel manufacturer regarding the mechanical properties for the cladding tubes and the construction tubes. There are many other manufacturing processes available for manufacture of Zircaloy. It depends upon the specification asked and the capabilities of supplier. Also these days as the burn-up rates are increasing there is a heavy research going on to make it more compatible with those temperatures and pressures. The material is being developed to reduce nodular corrosion and hydride cracking and is being fatigue tested and creep tested to make sure that the material will meet the requirements.

Patent Review Published Year | Details | Applicant | 1978 | Method of producing zircaloy tubes, US 4090386 A | Sandvik Special Metals Corporation | 1983 | Zirconium alloy fabrication processes, EP 0085553 A2 | Westinghouse Electric Corporation | 1987 | Process for producing zirconium-based alloy and the product thereof, US 4678521 A | Hitachi, Ltd. | 1993 | Zircaloy-4 processing for uniform and nodular corrosion resistance, US 5194101 A | Samuel A Worcester, James P Dougherty, John P Foster, Westinghouse Electric Corporation | 1997 | Method of fabricating zircaloy tubing having high resistance to crack propagation | General Electric Company | 1998 | Protective coating to reduce stress corrosion cracking of zircaloy fuel sheathing, WO 1998045851 A1 | CA Atomic Energy Ltd | Below listed are the major patents published in the development of Zircaloy. It is not an exhaustive list, it is created to show some of the important patents published during the development of Zircaloy.

Current applications:
In reactor systems such as BWR (Boiling water reactor), PWR (Pressurized Water reactor), and CANDU-PHW(Pressurized heavy water cooled) fuel cladding tubes, pressure tubes and sometimes calandria tubes are all made of zirconium alloy. Also Gartner springs are made with Zircaloy. Below listed are the types of alloys used at various places in nuclear reactor core.
Fuel Cladding: Zr-2/Zr-4
Pressure Tube: Zr-2/Zr-2.5% Nb
Calandria tube: Zr-4
Garter Spring: Zr-2.5 Nb, 0.5 Cu
Fuel cladding tubes: They are thin walled (.4 to .8 mm) and undergo a sequence of heterogeneous tensile, creep and recovery strains during their life The application of cladding in the above reactors is to keep the fuel separate from coolant. It has to withstand high temperature, neutron flux and stress due to internal gases and fission gases. Generally Zr-2/Zr-4 are used in BWRs, PWRs, PHWRs but Zr-1Nb is used as a clad material in VVERs (PWR in Russia).
Limitations of Zircaloy:
Pellet-filled rods have a precise physical arrangement in terms of their lattice pitch (spacing). They also interact with water (moderator) channels and control-rod channels. Hence physical structures that holds these rods are manufactured with tight tolerances. They should be resistant to high temperature, fluid and mechanical impacts, chemical corrosion and large static loads. Major problem in use of zirconium alloys is hydrogen absorption (from the metal/water reaction) and consequent embrittlement. This imposed the main design limit on zirconium alloys. Oxidized layer generally forms on Zircaloy in air and water that does not impair function.

Technical data of Zircaloy-4 Density | 6.55 g/cc | Coefficient of thermal expansion | 6 µm/ m0 C | Heat Capacity | 0.285 J/ g- 0C | Thermal Conductivity | 21.5 Watts/m-K | Melting Point | 18500C | Hardness | 89 Rb average | Modulus of Elasticity | 99.3 Gpa(14,402 Ksi) | Poisson's Ratio | 0.37 | Shear Modulus | 36.2 Gpa (5,249 Ksi) | Ultimate Tensile Strength (Longitudinal) at 2880 C | 271 Mpa (39.3 Ksi) | Ultimate Tensile Strength (Transverse) at 288 0 C | 241 Mpa(34.9 Ksi) | Yield strength at 2880C | 152Mpa | Elongation, % | 43 |

Alternatives for Zircaloy:
In loss of coolant, Zircaloy reacts with steam to produce hydrogen, thereby creating a nuclear accident. Work is being done to replace Zr with alternate materials such a SiC. But SiC has not been tested and simulated as widely as Zircaloy. To be adopted, a lot of development has to take place one of which is ceramic-ceramic bonding techniques that last high temperatures and pressures. Also, as fracture behavior of ceramics is more statistical it can only be predicted as a statistical likelihood of certain failure modes.

1)"FACTBOX-Nuclear Industry and Zirconium." Reuters. Thomson Reuters, 23 Apr. 2009.
2), 'Nuclear Fuel Fabrication', 2015. Fabrication/.
3) TES - Today's Energy Solutions, 'Nuclear Fuel-Rod Cladding Could Make Power Plants Safer', 2015. 4) R. Rose, K. Lunde and S. Aas, 'Zircaloy cladding tubes: Manufacturing techniques and achievable quality — A tube manufacturer's view', Nuclear Engineering and Design, vol. 33, no. 2, pp. 219-229, 1975.
5) Google Books, 'Patent US4450016 - Method of manufacturing cladding tubes of a zirconium-based alloy for fuel rods for nuclear reactors',
6) Google Books, 'Patent US4450020 - Method of manufacturing cladding tubes of a zirconium-based alloy for fuel rods for nuclear reactors',
7), 'Method for the manufacture of tubes of a zirconium based alloy for nuclear reactors and their usage Patent (Patent # 5,876,524 issued March 2, 1999) - Justia Patents Database', 1997.
8) Nuclear Metallurgy: Effect of irradiation on reactor structure materials, Priti Kotak Shah, PIED/BARC
9) [8]'Zirconium alloys in nuclear technology', Proc.Indian Acad.Sci.(Engg Sci), vol. 4, no. 1, pp. 41-56, 1981.

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