Zircaloy fuel cladding

The work on hi level waste solidification has led to applications of the same materials to other areas of waste management. These include decontamination of defense wastes currently in tank storage at Richland, WA, selective separation of Cs for beneficial uses, and development of a process flowsheet for conversion of Zircaloy fuel cladding hulls to sodium zirconate for use in waste stabilization. Each is briefly described below. 

Generic Safety Issue (GSI) 82 in NUREG-0933 (Reference 1), addresses the potential for a beyond-design-basis accident in which the water is drained out of the spent fuel pool, allowing the Zircaloy fuel cladding to ignite and thus release fission products from the spent fuel to the atmosphere. (The spent fuel pool is usually located outside the primary containment.)... 

I0.6.8.I Cladding failure in oxide fuel pins of nuclear reactors. The long-term operational performance of nuclear fuel pins is critically governed by the reactions that occur in the gap between the fuel and its cladding. Ball et al. (1989) examined this for the cases of (1) Zircaloy-clad pellets of U02+, in a pressurised water reactor (PWR) and (2) stainless-steel-clad pellets of (U, P)02+, in a liquid-metal-cooled fast-breeder reactor (LMFBR). In particular they were interested in the influence of O potential on Cs, I, Te and Mo and the effects of irradiation on the gaseous species within the fuel-clad gaps. 

The total yearly production of neutron-activation tritium in the PWR coolant is 400 Ci, as shown in Table 8.10. Another source of tritium in the coolant is fission-product tritium that diffuses through the fuel cladding and escapes through pin-hold penetrations through the cladding. Estimates of the amount of fission-product tritium reaching the coolant in LWRs with zircaloy fuel range from 0.2 to 1 percent of the total fission-production tritium produced within the fuel. 

The principal steps in the Purex process as applied to fuel clad with stainless steel or zircaloy are shown schematically in Fig. 10.2. Each of these steps will be described in more detail later in Sec. 4. 

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. 

The reactor core of an IPWR generally uses proven technology from loop-type PWRs or BWRs to minimize cost Typically, the fuel assemblies are square arrays of Zircaloy-4-clad, enriched UOj (< 5% U) fiiel rods with some burnable poison rod (GdjO,) positions. 

A fuel rod consists of UO2 pellets and a Zircaloy 2 cladding tube. UO2 pellets are manufactured by compacting and sintering UO2 powder in cylindrical pellets and grinding to size. The immersion density of the pellets is approximately 95% of theoretical UO2 density. 

Smith, D. H., Baldwin, D. L. An investigation of thermal release of carbon-14 from PWR Zircaloy spent fuel cladding. J. Nucl. Materials 200, 128—137 (1993)... 

Number of fuel pins 17x17 per assembly, Zircaloy-4 cladding... 

Two options are being considered for FBNR fuel elements. One is the zircaloy-cladded uranium dioxide spherical fuel pellet, and the other is spherical fuel element made of TRISO type coated particles. The enrichment for the U02/zircaloy option is about 3%, and for the coated particle option it is about 8%. Light water acts as both coolant and moderator. In a coated particle option, graphite also contributes to the neutron moderation. The module size will depend on fuel type and on the enrichment allowed. For example, the core of a reactor with U02/zircaloy fuel may have a diameter of 25 cm. The core tube may need to have larger diameter when coated particle fuel is used. 

The fuel elements are concentric tubes of metallic uranium enriched to 0.947 w/o u. The fuel cladding Is Zircaloy-2. Additional lattice dimensions are given in Table C-1 and a cross-eectlonal view of the lattice is shown in Figure C-1. 

The core is made up of a total of 800 fuel assemblies, arranged to form an array of roughly circular cross section, as shown in Fig. 9.8. Each of the assemblies consists of an 8 x 8 square array of fuel pins, with zircaloy-2 cladding, surrounded by a square-shaped channel of zircaloy-4. The assembly has tie plates at both ends, the lower of which has a nosepiece which fits into the fuel support and distributes the coolant flow to the rods. The use... 

Zirconium alloys, such as the zircaloys and Zr-2.5Nb, have been developed to better meet these requirements. In water-cooled reactors, zirconium alloys have found extensive use for fuel cladding and as pressure tubes. In systems in which the first of the listed requirements is of overriding importance for reason of neutron physics, the choice is virtually restricted to zirconium or one of its alloys. 

Corrosion-Irradiation Tests on Zircaloy 2-Clad, High-Uranium Core Fuel Plates - W. S. Brown, G. A. Freund... 

In the unalloyed form, zirconium is used for the construction of chemical equipment. Of much higher importance are however the zirconium alloys, from which especially the types zircaloy-2 (1.5 % Sn, 0.1 IS Fe, 0.1 % Cr, 0.05 % Ni) and zircaloy-4 (1.5 % Sn, 0.1 % Cr, 0.2 % Fe) are of interest. They are used as fuel cladding materials in pressure and boiling water nuclear reactors and for structural elements in the reactor core. 

In reality, the cladding mechanical failure occurs at some point in the plastic strain regicai. In the case of Zircaloy claddings of BWRs or PWRs, experimental results have shown that the cladding failures can be prevented as long as the cladding plastic strain level is less than 1%. Such experiments need to be conducted for the Super LWR fuel claddings in the future. In the meantime, the elastic strain limit is conservatively determined for the conceptual development. 

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