Microstructure evolution during PIA tests

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Fission gases

The set of stable gases created in PWR is of the order of 0.31 at/fission. As can be seen from Table 1.1, the gases thus formed, especially Xenon, rep-resent significant amounts. These fission gases are mostly insoluble in bulk UO2 and can lead to macroscopic phenomena like swelling and gas release. These phenomena directly affect the thermo-mechanics of the fuel rods and may shorten their lifetime. Due to their characteristics, the fission gas be-haviour needs to be understood properly as it concerns the safety of the nu-clear reactors. The fission gas behaviour will be discussed further in Section 1.4.

Defects in nuclear fuel

No material existing in nature is a perfect crystal. Atom arrangements in real materials do not completely follow perfect crystalline patterns. The imperfec-tions in the materials are called defects. These defects are best described in terms of their dimensions. The 0-dimensional defects affect isolated sites in the crystal structure, and are hence called « point defects ». The defects other than the point defects are extended in space and are referred to as « extended defects ». The 1-dimensional defects are lines along which the crystal pattern is broken and are called dislocations. The 2-dimensional defects are surfaces, such as the external surface and the grain boundaries along which the differ-ent crystallites are joined. The 3-dimensional defects are those which change the crystal pattern over a volume. These defects are discussed below.

Point Defects

Point defects are defects that occur only at or around a single lattice point. They are not extended in space in any dimension. During elastic collisions, if the energy transferred by the projectile is greater than the binding energy of the atoms in their lattice sites, the lattice atoms are displaced from their origi-nal positions. The displaced atoms, also known as Primary Knock-On atoms (PKA), can cause an avalanche of other atomic collisions if they have enough kinetic energy resulting in a succession of cascade displacement to form point defects. Point defects typically involve at most a few extra or missing atoms. Slightly larger defects in an ordered structure are usually dislocation loops (discussed later). The different point defects are shown in Fig.1.6 and are described below.

Vacancy

Vacancy defects are lattice sites which would be occupied in a perfect crystal, but are vacant. If a neighboring atom moves to occupy the va-cant site, the vacancy moves in the opposite direction to the site which used to be occupied by the moving atom. The stability of the surrounding crystal structure guarantees that the neighboring atoms will not simply collapse around the vacancy.

Interstitials

Interstitial defects are atoms that occupy a site in the crystal structure at which there is usually not an atom. They are generally high energy configurations, but small atoms in some crystals can occupy interstices without high energy. They may be the same type of atom as the others (self interstitial) or an impurity atom.
A Frenkel pair defect arises when a lattice atom is pushed to an intersti-tial site as a result of ballistic collision with an energetic particle or when the crystal is heated at sufficiently high temperature, provided that the energy absorbed by the atom is higher than its displacement energy. According to Olander [10], Frenkel pair defects are most commonly ob-served in UO2 under irradiation.
Schottky Defect and other defects including a U vacancy in UO2 x The Schottky defect is characterized by one uranium vacancy and two oxygen vacancies. For stoichiometric UO2, the Schottky defect is the most probable defect including a U vacancy [11].

Extended Defects

As the name suggests, these defects are extended in space. These can be categorized as one-dimensional line defects, two-dimensional planar defects and three-dimensional bulk defects.

Line defects

One-dimensional defects refer mainly to dislocation linear defects. The atoms of the crystal lattice around a dislocation are misaligned. Dislo-cation networks are likely to be formed in nuclear fuels due to the slow-ing down of fission fragments in high temperature conditions during in-reactor operations. There are two basic types of dislocations, the edge dislocation and the screw dislocation. « Mixed » dislocations, combining aspects of both types, are also common. Fig.1.7 depicts a schematic representation of the different types of dislocations. Edge dislocations are caused by the termination of a plane of atoms in the middle of a crystal. Screw dislocations result from shear distortion such that the atoms over the cut surface are shifted in a direction parallel to the dis-location line. For the mixed dislocations the shift is neither parallel nor perpendicular to the dislocation line.
Two-dimensional defects include surfaces, grain boundaries and stalk-ing faults. These defects act as the internal interfaces that separate neighboring regions within the same crystal structure but with different orientations. Solids are made of a number of small crystallites which are also known as « grains ». All the grains are separated by boundaries which are called « grain boundaries » and the atoms in this region are not in perfect arrangement. Grain boundaries occur where the crystal-lographic direction of the lattice changes abruptly. This usually occurs when two crystals begin growing separately and then meet. Fig.1.8(a) represents a grain boundary.
Stacking faults occur most commonly in close-packed structures. They are formed by a local deviation of the stacking sequence of layers in a crystal. An example would be the ABCABCBCABCA stacking sequence as shown in Fig.1.8(b).
Three-dimensional defects, also referred as volume or bulk defects, in-clude pores (voids), cracks and bubbles. Voids are small regions where there are no atoms, and which can be thought of as clusters of vacan-cies. In the nuclear fuels, voids are produced during the fuel fabrication process. The porosity of sintered UO2 can be classified either as open pores associated to the pellet surface or closed pores isolated from the surface. Fig.1.9(a) indicates the pores presented in as-fabricated UO2 disk sintered at 1400 C during 4 hours.
During nuclear operations, a large amount of insoluble gases (mainly Xe and Kr) precipitate as bubbles and start to agglomerate in the fuel matrix. The behaviour of these gaseous bubbles needs to be considered as they can significantly alter the physical and mechanical properties of the fuel. As an illustration in Fig.1.9(b), Michel et al. [13] presented the formation of a multimodal bubble population during the implantation of UO2 thin film with Xe ions which was then followed by an annealing at 1500 C.

Fission gas behaviour

Under irradiation, the crystallographic network of UO2 undergoes changes that have an effect on its physio-chemical properties. Fission gases play a significant part in this evolution and their behaviour needs to be understood. About 15% of the generated fission products consist of the noble gases, xenon and krypton. The solubility of these noble gases in the UO2 matrix is extremely
low [14] and they either tend to be released from the fuel or precipitate in the form of small nanometer size clusters leading to macroscopic phenomena like Fission Gas Release (FGR) and swelling in the fuel. Both swelling and FGR are detrimental to fuel safety.
If the gas is released from the fuel, the pressure within the fuel rod is correspondingly increased, subjecting the cladding to additional stress. The thermal conductivity of the gap is lowered, since xenon has a lower conductiv-ity than the cover gas, helium. This further increases the temperature which accelerates gas release and, eventually, the cladding is subjected to further stresses that can ultimately result in failure.
On the other hand, if the fission gases are retained in the fuel, they sig-nificantly precipitate as bubbles. As the density of the gas in such bubbles may be considerably lower than that of the gas in the solid fuel (especially for inter-granular bubbles, or at high temperature), gas atoms residing in bubbles occupy more volume than either the fissile atoms they replaced or fission-product atoms that segregate as solid phases. The precipitation of fission gases thus leads to swelling of the fuel to a larger degree than the volume ex-pansion that would occur if the xenon and krypton had remained dispersed on an atomic scale in the fuel matrix. Swelling contributes to the pellet-cladding interaction, again exposing the cladding to higher stress and ultimately limit-ing the life time of the fuel rod. Pellet-cladding interaction due to swelling of fuel in the fuel rod is depicted in Fig.1.10.
To better understand the evolution and release of gaseous fission prod-ucts, numerous studies have been conducted for several decades on the dif-fusion of Xe and Kr atoms in UO2, as well as nucleation and growth of bub-bles/cavities. The mechanism of FGR out of the fuel can be understood as a two step process. The first and basic step is the transfer of gas atoms from within the grain to the grain boundaries (intra-granular fission gas release). The second step is the coalescence and inter-connection of grain-boundary bubbles to give pathways to gas atoms at the grain boundaries to go outside the fuel (inter-granular fission gas release). The latter is clearly observed in high temperature conditions (ramp tests or annealing tests). We will be discussing the different mechanisms associated with these two steps in the following sections.

Mechanisms of FGR

Recoil and Knock-out

At temperatures below 1000 C, where the thermally activated processes do not dominate, the fission products formed at the external surface of the UO2 pellets can escape by the direct recoil or knock-out mechanisms. The fission fragments having large kinetic energies (60 to 100 MeV) can dissipate their energy to the fuel lattice, primarily by interaction with the electrons of the ma-terial. However, a fission fragment close enough to a free surface (< 6 to 7 mm), can escape from the fuel due to its high kinetic energy. This mechanism of release is known as (direct) recoil release.
Similarly, when the fission fragments make elastic collisions with the nu-clei of atoms of the lattice, these atoms can also become energetic particles, called primary knock-ons with mean energy of 100 keV. These can then be released, or transfer part of their energy to neighbouring atoms, thereby, creating higher order knock-ons with mean energy of 200 eV [16]. The in-teraction of a fission fragment, a collision cascade or a fission spike with a stationary gas atom near the surface can cause the latter to leave the fuel. This mechanism of release is known as knock-out release.
These release mechanisms in fuel operating at low powers were studied by Olander [10], Wise [17], and Lewis [18]. Recoil release can contribute significantly in the release of short-lived radioactive nuclides in failed rods and in experiments [19] on low temperature fission product release in which high density gas surrounds the fuel [17, 18]. The release by knock-out is negligible compared with that of recoil for short-lived isotopes. These release mechanisms, active at low burn-ups, contribute less than 1% to the release of generated gas [17, 18].

Single gas atom transport

The transport of single gas atoms in the bulk UO2 is determined by the diffu-sion rate of fission gas atoms and by the interaction with the fuel microstruc-ture, in particular, the trapping and re-solution of gas atoms by and from intra-granular bubbles, respectively [10, 20, 21]. In this section, the transport refers to the unperturbed diffusion of gas atoms (mainly Xe) in the bulk lattice with-out the interaction with the fuel microstructure. The interactions with the fuel microstructure like trapping and re-solution are discussed in the next sections.
Several studies based on density functional theory (DFT) [22, 23, 24, 25, 26, 27] and molecular dynamics (MD) with emperical potentials [28, 29, 30, 31, 32] show that Xe gas atoms prefer uranium vacancy trap sites. There is a general agreement between DFT and empirical potential calculations for the most favourable sites for Xe atoms in the UO2 lattice, with regard to thermo-dynamics. These are the divacancies (one uranium vacancy and one oxy-gen vacancy) in stoichiometric (UO2) matrices; a bound Schottky defect (one uranium vacancy and two oxygen vacancies) for hypostoichiometric (UO2 x) matrices; and single Uranium vacancies for hyperstoichiometric (UO2+x) ma-trices [33]. The computational study of Nicoll et al. [34] showed a shift of the most probable site for Xe atoms from di- to tri-vacancies when the fuel be-comes hypostoichiometric. As a concequence, the lattice diffusion coefficient for inert gas atoms in UO2 is influenced not only by the temperature, but also by the stoichiometry.
Above 1000 K, the single gas atom diffusion coefficient in UO2 can be expressed by the Arrhenius law suggested by Matzke [35] as:
where D0 is the pre-exponential factor (= 5*10 4 to 2.90*10 12 m2/s [27]), DH represents the activation enthalpy (= 2.87 to 3.95 eV/atom [27]), and k is the Boltzmann constant (= 8.6173 eV/K).
Depending on the stoichiometry of uranium oxide fuel, although there is general agreement that Xe-release from UO2+ x is faster than release from UO2, and, conversely that the Xe-release is inhibited in UO2 x, there is es-sential disagreement on the details [36, 35, 37, 38].
Several kinds of doping are added to the UO2 fuel in order to improve different aspects of fuel performance [39, 40, 41, 42]. Doping of UO2 with foreign cations can simulate the non-stoichiometry, hence it affects the defect structure as well.
During irradiation, Xe atom mobility is increased due to the fission that creates more vacancies (irradiation enhanced diffusion) or by directly moving the Xe atoms in the displacement cascades (athermal diffusion).

Trapping of gas atoms

The migration of gas atoms in nuclear fuels involves more than simple lat-tice diffusion. The solubility of noble gases in the UO2 matrix is extremely low (See Section 1.5.5). A direct consequence of the low solubility is the tendency of the moving gas atoms to be easily trapped in more stable config-urations in various clusters or sinks, such as closed pores, vacancy clusters, precipitation of gases into bubbles, as well as trapping on dislocations or at grain boundaries [43, 44]. Experiments suggest that for burn-ups reached in power reactors, gas atom trapping by fission produced defects should be more important than trapping at natural defects in the as-manufactured fuel [10]. TEM observations have shown that the most possible trap is the popula-tion of fission gas bubbles [45]. Experiments on single crystals and sinters of stoichiometric UO2 indicate that the apparent diffusion coefficient of Xe is dra-matically decreased for fission doses exceeding 1015 fissions/cm3 [36, 46]. The strong tendency of the gas atoms to precipitate into bubbles has been demonstrated for samples implanted with Kr and Xe ions, even at tempera-tures as low as 300-350 C [47]. Volatile fission products such as Cs, I, and Rb can also form bubble-like structures in the bulk depending on the temper-ature and the chemical nature of the impurity [48]. The factors affecting the trapping of migrating fission products by bubbles are the diffusion coefficient of the fission product and the bubble number density and size.
Other significant trapping sites for fission gas atoms are the dislocations. Nerikar et al. [49] investigated Xe segregation to edge and screw dislocations in UO2 and the segregation trends were found to be significantly dependent on the dislocation characteristics. Xe prefers to segregate to screw disloca-tion rather than to edge. The Xe segregation to dislocations is found to be thermodynamically favorable. Furthermore, MD simulations using empirical potentials of Xe diffusion around edge dislocations in UO2 were performed by Murphy et al. [50] to analyze the scenario for FGR via this pathway. The results suggest that the activation energy for Xe diffusion is dramatically re-duced for free gas atoms in the vicinity of a dislocation as compared to the bulk. However, Xe atoms diffusing along the dislocation aggregate to form small bubbles, which incorporate all of the isolated mobile Xe atoms and in-hibits fast diffusion of Xe along the dislocation core.
During irradiation, an irradiation-induced re-solution mechanism is also operational (see section 1.5.4), hence the trapping and the re-solution rates can be balanced. Accordingly, the migrating atoms are at equilibrium be-tween the lattice and the traps. Speight [51] has shown that saturation of the intra-granular bubbles occurs relatively fast and many FGR models define an effective diffusion coefficient, considering the trapping and re-solution of gas atoms. Practically, the free, untrapped gas atoms still diffuse with their in-trinsic diffusion coefficient, but the fraction of them reduces. The re-solution of gas atoms from the bubble to the bulk UO2 is discussed in the upcoming sections.

Irradiation induced re-solution of gas atoms

The low solubility of rare gas atoms in UO2 provides a strong driving force for the precipitation of gas atoms in bubbles. The trapped gas atoms can, however, simultaneously be freed from the traps during irradiation. Early ex-periments [52, 53, 54, 55, 56, 57] have shown that the population of gas in the matrix significantly increases due to irradiation induced re-solution of the gas, which thus, remains capable of diffusing out of the grains. The trapping effect can saturate either if there is no more gas in the solid or if there is re-solution of gas into the bulk. This saturation of the trapping effect has been demon-strated in experiments [36], which is in accordance with the conclusions of Lawrence [58].
Among g-rays, neutrons and fission fragments, which can all cause the re-solution of fission gases from bubbles, only the fission fragments are efficient as observed experimentally. The reason for this being that the fission frag-ments have a higher kinetic energy (50-100 MeV) than fast neutrons (2 MeV), and they are also charged, thus, having a higher cross-section for transferring energy to the lattice, or to the gas atoms. The plausible mechanisms of irra-diation induced re-solution have been reviewed by Turnbull [57] and Olander [10]. Olander and Wongsawaeng [59] have reviewed the re-solution of fission gas bubbles by two radiation induced mechanisms: heterogeneous mech-anism, based on the work of Turnbull [55] and homogeneous mechanism, based on the work of Nelson [54].
Homogeneous mechanism involves the re-solution of individual fission gas atoms, one by one, from the bubble by direct elastic collisions with fission frag-ments or energetic primary knock-on atoms (PKA). Heterogeneous mecha-nism, on the other hand, involves instantaneous complete or partial re-solution of the fission gas contained in a bubble by a fission fragment passing through the bubble due to the energy lost by it through electronic excitation. These en-ergetic ions first heat the electrons, then part of the energy is transferred from the hot electrons to the lattice through electron–phonon coupling. This pro-duces a cylindrical thermal spike which can easily reach temperatures above the melting point of the material, and if the spike intersects a gas bubble, a purely thermally-driven re-solution may occur [60].

Table of contents :

General Introduction 
1 Context and Literature review 
1.1 Introduction and Context
1.1.1 Pressurized Water reactor (PWR)
1.2 Description of the UO2 nuclear fuel
1.2.1 Chemical composition of Uranium dioxide
1.2.2 Fission and fission products
1.2.3 Fission gases
1.3 Defects in nuclear fuel
1.3.1 Point Defects
1.3.2 Extended Defects
1.4 Fission gas behaviour
1.5 Mechanisms of FGR
1.5.1 Recoil and Knock-out
1.5.2 Single gas atom transport
1.5.3 Trapping of gas atoms
1.5.4 Irradiation induced re-solution of gas atoms
1.5.5 Thermal re-solution of gas atoms
1.5.6 Diffusion in the temperature gradient
1.5.7 Grain boundary diffusion
1.5.8 Intra-granular bubble migration
1.5.9 Grain boundary sweeping
1.5.10 Inter-granular bubble interconnection
1.5.11 Burst release
1.6 Mechanisms associated with intra-granular gas release during PIA tests
1.7 Experimental observations during post-irradiation annealing (PIA)
1.7.1 Microstructure evolution during PIA tests
1.7.2 FGR during PIA tests
1.8 Modelling approaches for FGR
1.8.1 Booth Model
1.8.2 Improving the Booth Model
1.8.3 Speight’s Model for effective diffusion
1.8.4 Beyond Speight’s Model
1.8.5 FGR modules in Fuel performance codes
1.9 Detailed modelling of intra-granular bubbles
1.9.1 Different scales of modelling
1.9.2 Meso-scale modelling
1.9.3 Prominent meso-scale methods
Cluster Dynamics
Kinetic Monte Carlo (kMC)
Phase Field Model (PF)
1.9.4 Application of meso-scale modelling for intra-granular bubbles
Conclusions from the application of meso-scale models for intra-granular bubbles
1.10 In Closing
2 A new spatialized meso-scale model: BEEP Model 
2.1 BEEP Model: Introduction
2.1.1 Assumptions used in the model
2.2 Modelling methodology
2.2.1 Representation
2.2.2 Physical properties used in the model
2.3 Methodology adopted in the model
2.3.1 Initialization
2.3.2 Diffusion of point defects and crystal atoms
2.3.3 Space and Time Discretization
2.3.4 Bubble volume calculation
2.3.5 Determining the next center (« Target center ») of bubble
2.3.6 Re-drawing the bubble keeping the atom and volume balance
2.3.7 Random movement of the bubbles
2.4 Parallelizing the BEEP Model
2.4.1 Using OpenMP
2.4.2 Implementing sub-domains
2.4.3 Message Passing Interface (MPI)
2.5 Testing the Model
2.5.1 Crystal atom balance
2.5.2 Verification for calculation of solid ratio « RS »
2.5.3 Verification for diffusion calculation
2.5.4 Verification of particular functionalities of BEEP Model
Coalescence of two bubbles
Vanishing of a bubble in the presence of a large bubble
2.5.5 Verification for bubble movement
2.6 Technical note on the exterior region (« Flat bubble »)
2.7 In Closing
3 FGR due to bubble migration in a vacancy gradient 
3.1 Directed movement of intra-granular bubbles in a vacancy gradient
3.1.1 Evans’ model for directed movement of bubbles
3.1.2 Testing the movement and growth of bubbles in the BEEP Model
3.1.3 Grid size and Coalescence
3.2 Analysis of FGR via Directed Movement Mechanism
3.2.1 FGR values from experiments for comparison
3.2.2 Conditions for numerical analyses
3.2.3 FGR analysis
3.3 FGR analysis until vacancy gradient lasts
3.4 Uncertainty analysis of parameters
3.4.1 Uncertainty analysis of Dv
3.4.2 Uncertainty analysis of Ceq
3.5 In Closing
4 Brownian motion of bubbles and impact on FGR 
4.1 Random (Brownian) movement of bubbles
4.1.1 Bubble diffusion by volume diffusion mechanism
4.1.2 Bubble diffusion by surface diffusion mechanism
4.2 Influence of random bubble movement without vacancy diffusion
4.2.1 Random movement of bubbles due to volume diffusion mechanism
4.2.2 Random movement of bubbles due to surface diffusion mechanism
4.3 FGR due to combined random and directed bubble movement in a vacancy concentration gradient
4.4 Dependence of bubble diffusion coefficient via surface diffusion mechanism on the bubble radius
4.5 FGR due to combined movement until vacancy gradient lasts
4.6 Parametric investigation of Mikhlin’s suppression term
4.7 In Closing
5 Bubble movement via Dislocations 
5.1 Introduction to Dislocations
5.1.1 Mechanisms of Dislocation movement
Dislocation glide
Dislocation climb
5.2 Scenario of bubble movement with dislocation climb
5.3 Implementation of Dislocations in BEEP Model
5.3.1 Assumptions
5.3.2 Formulation
5.3.3 Methodology
Initialization
Visualization
Calculating the volume of pinned bubbles
Equilibrating bubbles on a dislocation
Center targets for bubbles pinned on dislocations
Updating dislocations
Calculating the new dislocation velocities
Choosing an acceptable Dt for dislocation evolution
5.4 First 3D Test Case for Dislocation and bubble movement
5.5 Scenario of FGR via bubble movement with Dislocations
5.5.1 Conditions for analysis
5.5.2 FGR analysis without any diffusion of vacancies from free surface
5.5.3 Factors influencing the velocity of dislocations
5.6 In Closing
General Conclusion

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