Fuel pins/rods

Once enriched, the UFg needs to be reduced to either uranium metal or UO2 to be formed into fuel pins. A variety of methods can be used to accomphsh the conversion to the oxide however, the predominately used technique involves reduction of the UFe to U metal fully, using Ca at high temperatures, followed by burning in oxygen. Once formed, the UO2 is pressed into pellets, which are then fed into fuel rods. 

Wesley and Yovanovich [127] compared the predictions of the proposed gap conductance model and experimental measurements of gaseous gap conductance between the fuel and clad of a nuclear fuel rod. The agreement was very good and the model was recommended for fuel pin analysis codes. 

The principal component in any reactor is the core which contains the fissionable fuel material. This is usually UO2 enriched in to 2 — 4% shaped as short cylinders, pellets, which are stacked in zirconium alloy tubes (the can, canning or cladding) forming fuel rods or fuel pins. The fuel rods are mounted together in clusters formingyue/ elements or fuel assemblies. 

The important features of the PWR core are shown in Figure 19.12, which, although taken from three different reactors, represents the typical Westinghouse design. A typical core contains 40 000 fuel rods (a) in 193 assemblies, each with space for 208 fuel pins... 

Fuel pin failure rate (FPFR) for fuel rods of Ukrainian WWER-1000... 

This arrangement maximizes heat transfer from fuel to coolant and is used to minimize peak fuel temperature. A variant of these fuel pins could be used to replace the fuel pins in an AGR fuel assembly. The stainless steel grid structure that holds the AGR fuel pins in place and the tie rod could be replaced by carbon-carbon composites. A preliminary assessment of the carbon--carbon technology and the AGR design has not identified any insurmountable fabrication challenges to create an equivalent carbon-carbon composite of the AGR fuel assembly. However, only limited analysis of such fuel designs has been performed. A significant fuel development effort would be required. 

Once water has reached the core via the EPR tubes, the corrosion will spread out concentrically from each of the three ingress sites into the Pb-Bi, as shown in Figure 16. The nearest fuel pins are about 6 mm distance from the rod channels corrosion will progress at the Pb-Bi corrosion rate until... 

As indicated earlier, the EPR and CCR are made of EuB, a material with ceramic properties. Little information is available about corrosion rates of EuB in water. It is assumed here that the rate is similar to that of other ceramics such as BeO, and a rate of 0.01 mm a is assumed. Water enters the channels at D + 125 years hence, corrosion of the rods will commence at that date. The EPR and CCR have radii of 10 mm and 8.5 mm, respectively, and the CCRs are surrounded by approximately 1 mm of Pb-Bi coolant within the channel. Both types of rod channel were filled with Furfurol(F) prior to dumping, but the CCRs are also surrounded by solidified Pb-Bi filling the channel to a height of approximately 500 mm above the core, which will prevent water reaching the rods from above for at least 50 000 years. Hence, only the EPRs will be attacked by sea water within the timescales of interest, beginning at D + 125 years. The EPRs have a 10 mm radius and so will corrode away after 1000 years (D + 1125 years). TTie resulting release rates of activated Eu will be modified by the change in size (i.e., surface area) of the rod this mechanism will be identical to the modification to the fuel pin corrosion rate. 

In the case of the submarine PWRs and icebreaker, the SNF is initially surrounded with Furfurol(F), which can be assumed to degrade in 100 years, allowing water to enter the core and surround the fuel pins. This will increase the reactivity of the core however, in normal operation the control rods will have been designed to produce a safe shut down condition, and with all rods present there is no risk of criticality. Combined with control rod corrosion and compaction of SNF under its own weight at the bottom of the RPV, however, the presence of water will increase the probability of criticality. 

The use of double clad fully tested fuel pins under conditions where fuel pin overheating is technologically excluded, ensures long-term and reliable localization of fission products inside the fuel rods. 

Nonetheless, the experimental criticality data cur> rently available for LMFBR fuels do provide a secure validation point for (Pu, U)C>2 fuels to be used in the fast flux test facility (FFTF) program. Indeed, the data in Table H were obtained for this program. As a consequence, the criticality safety parameters for FFTF fuel pin handling are currently based directly on the experimental data, since the fuel rods used in the experiments are nearly the same sizes as those to be used in the FFTF reactor. These criticality parameters are as follows ... 

Additional data" also concern heterogeneous systems containing PuOa end UQ) in fuel pins that were made critical in water and in boronated water. Reported also are similar measurements with V(2.35)0> fuel i ns. Close agreement between calculated and observed gen-values, power distributions and rod worths was obtained."... 

Analysis of the data for the 22.9-mm lattice spacing shows that the critical number of fuel pins increased from about 223 in water without neutron poison to about 422 in water containing 0.21-g Gd/litre. 392 rods were required for 1.1 g Cd/litre, in water, and 433 rods for 0.6-g S/litre in water. 

The Advanced Fuel Recycle Program is concerned with the safe reprocessing of mixed plutonium and uranium oxide fuels, characteristic of fast reactors. The safe handling and storage of these fuels hin on calculations and these in turn depend on clean, well-deflned experiment data for validation. Benchmark experiment data have been. acquired for fast test reactor (I R)-type fuels for impoisoned systems and systems intermixed with soluble poisons. However, there, are no data now available, fliat explore the criticality of these fuels intermixed with solid neutron absorbers (poisons). In this paper, we will present the results of experiments performed at the Pacific Northwest Laboratory (Critical Mass Laboratory) on fast test reactor fuel elements intermixed with solid neutron absorbers. The isons used were Bbral and cadmium plates and gadoliniuth cylindrical rods. Each absorber was separately examined to see its reactivity effect on lattices of FTR fuel pins in water none was intermixed. 

The critical assemblies were reflected on all sides by at least 150 mm of deionized water. The rods, however, were immersed in water only to a height of 150 mm above the top of the fuel region. Consequently, the top reflector region above each fuel rod was perforated by fuel pin hardware, projecting through the top reflector region. 

Verification of the calculations involving 15 wt% Pu and 30 wt% Pu was obtained by calculating keff for a critical array experimentally determined by Bieiman et al. The fuel pins in this experiment were composed of 19.84 wt% Pu in (U,Pu)02, with the plutonium containing 11.5 wt% Pu. The fuel materia) had a diameter of 4.94 mm, was clad in Type 316 stainless steel, and the rods were arranged in a 9.677-mm square lattice. The csdculated value of kerf was... 

PRISM fuel rods are bundled together into a hexagonal lattice. A fuel assembly schematic is shown in Figure 6.18. Rod spacing is maintained by a spiral wire wrap around each rod. Each fuel assembly includes a hexagonal duct around the fuel pins to maintain coolant flow around the pins. After the entire fuel cycle, spent assemblies decay for one fuel cycle in in-vessel storage. 

The calculational method used in Step 6 is usually diffusion theory, often improved through the use of "transport corrections" of various sorts. Diffusion theory, as was previously mentioned, is the simplest neutron transport calculational method. It assumes a very low-order directional flux shape, which makes it inappropriate for use in problems that involve strong isolated sources or strong isolated absorbers. It is not accurate enough for Steps 3-5 because of fuel pin and control rods once these sources and absorbers have been "smeared" into the assembly cross sections that come from Step 5, diffusion theory is generally accurate enough for Step 6 full reactor calculations. 

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