Fuel evolution in normal operating conditions .1 Thermal evolution .1 Thermal evolution

Due to the poor thermal conductivity of UO2 and MOX fuels [20], [21], a radial temperature gradient appears in the fuel pellet. Indeed, its periphery, closer to the coolant, is cooled more efficiently than the center of the pellet. In the typical range of linear power in a PWR (180 to 270 W.cm^-1), the temperature in the center of the fuel thus ranges from 900 to 1300°C whereas it is 500-600°C at the periphery of the pellets, as shown in Figure I-9 [5]. The temperature difference observed between the inner face of the cladding and the periphery of the pellet is mainly due to the different nature of the materials (oxide / metal) and the gap filled in He.

During the irradiation cycles, the fuel undergoes chemical and physical evolutions, which can modify the temperature profile within the pellets. Moreover, in contact with water, a ZrO2 layer is formed at the surface of the cladding, reducing its thermal conductivity. The gap between the fuel and cladding also tends to close during irradiation (swelling of the materials, pressure of the coolant on the fuel rods…).

14/300 The strong thermal gradients imposed to the fuel coupled to the defects induced by irradiation lead to the fracturing of the pellets in the first cycles of irradiation. They are also a driving force for the diffusion of elements within the fuel pellets, which can lead to gaseous FP release in PWR normal operating conditions [5].

Figure I-9: Radial temperature profiles observed in a PWR fuel rod as a function of the linear power, extracted from [5].

I.2.2.2 Mechanical evolution

The thermal gradient applied on the fuel pellets leads to differential dilatation between the center and the periphery. This phenomenon induces tangential traction stresses on the pellet and given the fragile behavior of the fuel at the temperatures considered, axial cracking occurs in the first cycle of irradiation [22].

A densification occurs in the pellet which adopts a diabolo shape. Simultaneously, fission induces swelling of the fuel. Below 15 GWd.tHM-1, these effects combine and cause a contraction of the fuel.

Above, the swelling of the fuel coupled to the irradiation creep of the cladding and the pressure of the coolant lead to the progressive closure of the fuel-cladding gap, which is almost complete for a burn-up of 30 GWd.tHM-1. These different phenomena are represented in Figure I-10.

Figure I-10: Evolution of the fuel rod during irradiation extracted from [4]: cracking of the fuel pellets, appearance of the diabolo shape and closure of the fuel-cladding gap. The space observed between two pellets is due to the dishing applied

at the top and bottom surfaces of each pellets during their fabrication.

15/300 I.2.2.3 Microstructural evolution

I.2.2.3.1 High Burn-up Structure

Non fissile elements such as ²³⁸U composing the main part of the fuel pellets are fertile elements.

Given the neutron energy spectrum of a PWR, they are able to capture an epithermal neutron and turn into a fissile nucleus (i.e. ²³⁹Pu) [23], [24]. This reaction has a higher probability to occur at the pellets periphery because of spatial self-shielding effect of the rods. As, ²³⁹Pu has a higher probability to undergo fission compared to ²³⁵U [25], the local burn-up in the periphery of the pellet or in the Pu agglomerates can be two to three times higher than in the rest of the pellet.

As soon as a threshold local burn-up of 50 GWd.tHM-1 has been reached at temperatures below 1100°C, the fuel undergoes restructuring [26]. The structure appearing is called High Burn-up Structure (HBS) and was characterized experimentally in LWR fuels in several studies [27]–[35]. The main characteristics of this HBS region are:

 High porosity (up to 20 %) linked to a high density of fission gas bubbles.

 High FP content and notably a high density of metallic precipitates composed of Mo, Ru, Rh, Pd and Tc.

 Grain subdivision resulting in submicronic grains of two types [32]: the polyhedral subgrains observed in the bulk of the HBS region have a size up to 0.8 µm and round subgrains with a size around 0.1 µm are surrounding by gas bubbles and open surfaces.

In LWR UO2 fuels, the HBS extends in the very periphery of the pellets on several µm at 40 GWd.tHM-1, up to more than 1 mm beyond 100 GWd.tHM-1 [35]. A picture of the periphery of a PWR UO2 fuel pellet irradiated up to a burn up of 70 GWd.tHM-1 is shown in Figure I-11.

Figure I-11: Fractography of the periphery of a PWR UO2 fuel irradiated up to a burn-up of 70 GWd.tHM-1 [36].

The behavior of MOX fuel is more complex than UO2. Indeed, typical HBS develops only in the Pu agglomerates where the local burn-up overcomes the HBS formation threshold (up to mid pellet radius). These agglomerates appear as dark phases in Optical Microscopy (OM), due to their high porosity. The restructuration limit in LWR irradiated MOX pellets is usually set around 0.5 R (where R is the pellet radius). Pu agglomerates in the center of the pellet are hardly distinguishable as few porosity is present [33], [34]. Moreover, the size of fission gas bubbles and subgrains increase in the higher temperature region of the fuel (towards the center of the pellet) [27].

This restructuration of the pellet induces a decrease of the thermal conductivity as well as softening of the fuel in the HBS region [26].

16/300 I.2.2.4 Chemical evolution

I.2.2.4.1 Oxygen potential

Due to the fission reaction, the fuel chemistry is also strongly impacted. Indeed, several new elements are created in the fuel and can react with it. The majority of these elements have an oxidation state inferior to U or Pu in the fuel which globally implies an increase of the oxygen on metal ratio (O/M ²) in the pellet.

This O/M value enables to calculate the oxygen potential³ in the fuel, which plays a major role on the reactivity of the different FP, the fuel and the cladding. Indeed, as shown in Figure I-12, an element M having a Gibbs energy below the fuel oxygen potential is able to react with O and form an oxide (MOx). Conversely, if the Gibbs energy of the reaction M + x/2 O2  MOx is over the fuel oxygen potential, it stays as an element in the fuel matrix as a metal for example.

As the equilibrium oxygen potential of Zr/ZrO2 is below the ones of U/UO2 and Pu/PuO2, the internal part of the cladding will thus oxidize as soon as a fuel-cladding contact is set, enabling O to diffuse.

This reaction would consume the excess of oxygen available at the periphery of the fuel leading to a decrease of the O/M ratio at the periphery of the fuel.

2 The O/M ratio is defined as the total amount of O on the sum of the amount of elements found as oxides (U, Pu and all the oxide FP dissolved or not).

3 The oxygen potential is defined as ∆𝐺 = 𝑅𝑇 ln 𝑝 where R is the ideal gas constant, T is the temperature and pO2 is the oxygen partial pressure.

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Figure I-12: Ellingham diagram describing the relative partial molar Gibbs free energies of oxygen of the fission product oxides, UO2±x and U0.8Pu0.2O2±x extracted from [37].

Finally, due to the radial thermal gradient and O concentration gradient present in the fuel pellet, the oxygen can diffuse from the center (hottest area) to the periphery (coolest area). This oxygen redistribution thus contributes to an increase of the oxygen potential in the periphery of the pellet, which will still stay inferior to the oxygen potential in the center of the pellet.

According to the work of [38]–[40], the oxygen potential in UO2 LWR fuels with burn-ups from 30 to 75 GWd.tHM-1 would range from -440 kJ.molO2-1 in the periphery of the pellets to -340 kJ.molO2-1 in the center. As shown in Figure I-13, this range of oxygen potential at the temperature of the fuel is close to the oxidation of Mo in MoO2. Oxygen buffering of the fuel by the Mo/MoO2 couple thus maintains the oxygen potential of the fuel quasi-constant (around -450 kJ.molO2-1 at 750°C in the periphery) in normal operating conditions [38], [39].

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Figure I-13: Temperature profile and calculated oxygen potential as a function of the radial position in the high burn-up region, extracted from [40].

As shown in Figure I-12, the equilibrium oxygen potential of stoichiometric UO2 is slightly lower than the one of stoichiometric U0.8Pu0.2O2.00 in the nominal temperature range of a PWR. Thus, the oxygen potential in stoichiometric MOX fuel is slightly higher than in stoichiometric UO2.00. Thus, at 1000°C for a LWR MOX fuel with a burn-up of 100 GWd.tHM-1, the stoichiometry is reached for an oxygen potential of -400 kJ.mol^-1 (about 80 kJ.mol^-1 above UO2.00) [41]. The oxygen potential in MOX fuels decreases with O/M and Pu content [42], [43].

I.2.2.4.2 Fuel-Cladding Interaction

As stated in the previous paragraph, the oxygen potential equilibrium of Zr/ZrO2 is under the ones of U/UO2 and Pu/PuO2 (Figure I-12) inducing the formation of an oxide layer in the inner surface of the cladding during in-pile irradiation.

Post-irradiation observation performed on LWR fuels highlighted the presence of a bonding layer resulting of the interpenetration of the HBS region of the fuel pellet and the ZrO2 layer of the cladding. The thickness of this layer depends on the burn-up of the fuel and is typically in the range of 10 to 15 µm in the burn-up range 40-70 GWd.tHM-1 [27], [30].

LWR MOX fuels with a burn-up of 42 GWd.tHM-1 and a composition UO2 + 5% PuO2 obtained by a co-milling process were analyzed in the work of [33]. The size of the Pu agglomerates were 100 µm in average. The bonding layer of the Zircaloy-4 cladding near these agglomerates was two times thicker than near the UO2 matrix. This was inferred to the higher oxygen potential in the Pu agglomerates due to the higher yield of Pu fission compared to U.

RAMAN spectroscopy and Electron Probe Micro-Analyses (EPMA) characterizations performed in [44], [45] on PWR UO2 fuel irradiated up to 58.7 GWd.tHM-1 in its Zircaloy-4 cladding revealed three different crystallographic layers in the inner oxide layer of the cladding. Near the fuel pellet, a tetragonal ZrO2 structure, probably stabilized by irradiation and FP implantation and corresponding

19/300 to the bonding layer was observed. Then a monoclinic ZrO2 structure followed by a “damaged tetragonal” ZrO2 structure were observed. The latter may have been stabilized by compressive stresses or hypostoichiometry close to the metal-oxide interface. Tests realized on model materials revealed no effect of the type of cladding on the fuel-cladding interactions. Nevertheless, hyperstoichiometric fuel induced more rapid Zr-U interdiffusion leading to the quicker growth of the bonding layer.