Fuel cladding material

IDENTIFY the four materials suitable for use as fuel cladding material and their applications. 

The nuclear properties of fuel cladding material must also be satisfactory. For thermal reactors, it is important that the material have a reasonably small absorption cross section for neutrons. Only four elements and their alloys have low thermal-neutron absorption cross sections and reasonably high melting points aluminum, beryllium, magnesium, and zirconium. Of these, aluminum, magnesium, and zirconium are or have been utilized in fuel-element cladding. 

The choice of cladding material for fast reactors is less dependent upon the neutron absorption cross section than for thermal reactors. The essential requirements for these materials are high melting point, retention of satisfactory physical and mechanical properties, a low swelling rate when irradiated by large fluences of fast neutrons, and good corrosion resistance, especially to molten sodium. At present, stainless steel is the preferred fuel cladding material for sodium-cooled fast breeder reactors (LMFBRs). For such reactors, the capture cross section is not as important as for thermal neutron reactors. 

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 creep test of fuel cladding materials under irradiation with MARICOfMaterial Jesting Rig with Temperature Control . 

Many numbers of irradiation tests, such as for the nitride fuel and the carbide fuel, are now in progress. The creep test of fuel cladding material under irradiation with MARICOf Material Testing Rig with Temperature Control) was performed at the 29th duty cycle. 

Fuel cladding material Elio orE635 alloy... 

Transuranium elements (TRU) and FPs that cannot be recovered from the spent fuel cladding material. 

Fuel clad material (multiple choice magnesium alloy, stainless steel, zirconium alloy, zirconium-magnesium alloy, carbide compound) 2 ... 

Fuel clad material - the material of the tube containing the fuel pellets or rods. Basically, there are three types of cladding - magnesium alloys (e.g. Magnox), stainless steel, or various zirconium alloys. HTGR reactors use fuel particles coated with pyrolitic carbon and silicon carbide in two layers. Data providers should choose the appropriate option from the multiple-choice menu magnesium alloy, stainless steel, zirconium alloy, zirconium-magnesium alloy, carbide compound. 

Fuel clad material specification - Data providers should specify the fuel cladding material... 

Deleted (unclear meaning, not found at any type offuel) Renamed Fuel clad material moved to Reactor core group Renamed Fuel clad thickness moved to Reactor core group Moved to Reactivity control group Combined with Number of control rod assemblies ... 

The fuel element design is similar to that used in the WER reactors. The fuel element cladding has a 9.1 mm outer diameter and a 7.73 mm inner diameter. The fuel cladding material is E-110 or E-635 zirconium alloy. 

It is assumed that the corrosion resistance of the fuel cladding material is provided at chosen operational temperatures by a system for maintenance of the oxygen concentration in the coolant. 

The first possible system which emerged was one using the pressure tube construction, heavy water as the moderator and steel as the fuel cladding material. 

Table 8.1 Scorecard for fuel cladding material candidates...

Furthermore, ODS alloys are envisaged as promising fuel cladding material for both SFR and LFR to enhance fuel bum-up. In principle one can assume that ODS alloys based on ferritic or ferritic-martensitic steels show corrosion behavior similar to the steel matrix without dispersed oxide particles. However, this assumption would need on one hand an experimental confirmation and on the other hand it would be important to assess the corrosion performance of ODS tubes in the long term (corresponding to higher fuel bum-up) under relevant reactor conditions. 

Pellet—cladding interaction has not been a major focus of the on-going SCWR research as, for the most part, data under relevant conditions already exist for similar fuel and fuel cladding materials. Oka et al. [3] note that while the thermal expansion coefficient of the Zr alloy cladding used in LWRs is smaller than that of the UO2 fuel pellets, the thermal expansion coefficients of the stainless steels or nickel-base alloys being considered for the SCWR are expected to be close to that of UO2. Pellet—cladding interaction will need to be considered during future SCWR fuel qualifications tests. 

Currently, only austenitic stainless steels and some nickel-based alloys can meet these requirements. Table 4.1 lists candidate fuel cladding materials that have been considered for the Canadian, EU, and Japanese SCWR concepts. 

F. Bevilacqua, G.M. Brown, Chloride deposition from steam onto superheater fuel clad materials, Gen. Nucl. Eng. Corp. Rep. GNEC 295 (1963). 

The materials have to retain their properties in corrosive in-reactor environments (reactor coolant, MOX fuel), and for fuel cladding materials in reprocessing chemical environment. 

Fig. 8.24(a) compares the average deformation of the BACCHUS 1 bundle to the pro-filometry of a 15/15Ti (CF2 lot shown in Fig. 8.18(b)) irradiated at the same maximum dose of 109 dpa and Fig. 8.24(b) places this nuance among all the other fuel cladding materials of Phdnix in terms of maximum deformation [32]. 

V. Levy, et al.. Structural and fuel cladding materials, in H. BaiUy, D. Menessier, C. Prunier (Eds.), The Nuclear Fuel of Pressurized Water Reactors and Fast Neutrons Reactors Design and Behavior, Lavoisier Edition, 1999 (Chapter 4 of the hook). 

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. 

The LFR and GFR are in the very early stages of development as commercial reactors. While LFRs have been built, the operating temperatures are much lower than required for H2 production. New fuel clad materials would be required for high-temperature operation. No fuel has yet been chosen for the GFR—an advanced concept with many uncertainties. 

It was the development of the nuclear reactor for navy propulsion that led to the use of zirconium and zirconium alloys as a fuel cladding material. This program, headed by Captain, and eventually Admiral, Rickover demonstrated flie viability of using fission energy to propel navy ships and submarines. 

The criteria for abnormal transients to ensure the fuel integrity are very important. They limit the maximum allowable coolant temperature and the choice of the fuel cladding material to be used at high temperature. So, to maximize the economical potential of the Super LWR, and minimize the research and development efforts, the criteria need to be rationalized based on detailed fuel rod analyses. 

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