Pellet-cladding Mechanical Interaction Research Articles

The effect of hydride morphology on the failure strain of stress-relieved Zircaloy-4 cladding with an outer surface pre-crack under biaxial stress states

ABSTRACT Hydride precipitates are considered to affect cladding integrity adversely during pellet-cladding mechanical interaction (PCMI) in a reactivity-initiated accident (RIA). This study aims to clarify the role of hydride precipitates in cladding failure under the biaxial stress condition. A displacement-controlled loading method was applied with a biaxial expansion-due-to-compression (biaxial-EDC) test apparatus to maintain a constant ratio of axial to hoop strains. The tests were conducted at room temperature on unirradiated stress-relieved Zircaloy-4 cladding tube samples with various hydrogen contents, pre-crack depths, and strain ratios. The amount and distribution of hydride precipitates (hydride morphology) were evaluated quantitatively and hydrogen content was measured to assess its effect on the decrease in outer surface hoop strain at failure (failure strain) of the samples. The decrease in failure strain of the hydrided sampleswas found to be more significant under lower strain ratios in the samples with shallower pre-crack. The failure strain of sample tended to be more sensitive to hydrogen content under the strain ratio with a higher axial component in the case of samples with hydrogen contents higher than ~150 wppm. This tendency might be explained by a scenario that considered the formation of micro-cracks in the hydride precipitates during the EDC test.

Evaluation of FRAPCON-4.0’s uncertainties predicting PCMI during power ramps

The Pellet-Cladding Mechanical Interaction (PCMI) is a fuel failure-preceding mechanism and, as a consequence, a relevant aspect when the fuel rod performance is modelled. Given the current trend towards more flexible modes of operation, PCMI studies are getting focused on operational power ramps.This study assesses FRAPCON-4.0′s uncertainties predicting PCMI under ramping conditions. A set of ramps from a Halden test performed on high burnup fuel has been analysed in terms of cladding elongation. An important error between data and predictions has been obtained (overestimation higher than 50% at the end of ramps), which cannot be fully explained by the uncertainty associated to power and in-code models. However, if the rigid pellet approximation adopted in FRAPCON-4.0 is removed and a formulation for fuel creep is introduced, the code response is notably enhanced to the extent that the uncertainty from this approximation allows explaining the error found.

Simulation on Pellet–Cladding Mechanical Interaction of Accident Tolerant Fuel With Coated Cladding

In this study, based on the code Fuel ROd Behavior Analysis (FROBA), a thermal–mechanical analysis code initially developed for traditional UO2-Zr fuel elements by our research group, a modified version was developed to perform the fuel performance simulation of accident tolerant fuels (ATFs), named FROBA-ATF. Compared with initial version, the cladding could be divided into arbitrary number control volumes with different materials in the new code, so it can be used to perform the calculation for multilayer coatings. In addition, a new nonrigid pellet–cladding mechanical interaction (PCMI) calculation model was established in the new code. The FROBA-ATF code was used to predict PCMI performance of two kind fuels with coated claddings, including the internal surface coating and external surface coating. The calculation result indicates that because the coating surface was close to the inner surface of the cladding where also was the PCMI surface, the absolute value of the combine pressure of internal surface-coated cladding was substantial larger than that of the external surface-coated cladding, which might be harmful the coating behavior. However, the internal surface-coated mode can provide a protection for alloy due to the isolation from direct contact with fuel pellets, which can result in a much lower equivalent stress of zirconium body during the PCMI.

Isotropic softening model for fuel cracking in BISON

Abstract In oxide nuclear fuel, the temperature gradient from the centerline to the radial edge of the pellet induces cracking due to thermal stress. We present a model that represents the effect of cracking as an isotropic softening of the material. The model considers a reduction of the elastic constants as a function of the number of cracks present in the fuel. This pragmatic approach aims to represent the average effect of cracking on the fuel stresses without attempting to explicitly describe the crack pattern or localization. Albeit simplistic, this approach has definite advantages in terms of computational expense and numerical convergence behavior for the mechanical analysis. It is also consistent with the uncertainties inherent in modeling the fuel cracking process, and suitable for engineering calculations aimed at representing the global fuel element behavior. The applied scaling of the elastic constants conserves the principal strain values and redistributes the principal stresses isotropically. Moreover, the model allows for the effect of an evolving number of cracks. In particular, an empirical correlation for the number of cracks as a function of the rod average linear heat rate is developed. Although the basic concept of the model was known, in this work we revisit and improve the original formulation, and implement the model in the BISON fuel performance code. Application in BISON is demonstrated through simulations of an idealized single fuel pellet irradiation and two integral fuel rod experiments. Results showcase the impact on calculated fuel stresses, the coupling to fuel creep, and the effect on cladding strains during pellet-cladding mechanical interaction.

Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents

The mechanical response of a nuclear-grade silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite was investigated under mechanical loading conditions mimicking the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA). In a RIA, cladding deformation and failure can be induced by the rapid thermal expansion of the nuclear fuel. A pulse-controlled modified-burst test was used to investigate RIA-like PCMI scenarios on SiC/SiC composite samples at pulse widths from 12 to 100 ms.The strain-driven nature of the cladding sample deformation was due to the rapid internal pressurization and subsequent expansion of a secondary tube. A digital-image correlation technique was used to measure strains from the speckle-painted outer surface of the tubes. The failure strains of samples tested at slower rates, such as RIA event durations of 52 and 100 ms, showed good agreement with the literature-reported values for similar composites tested at slow strain rates. Additionally, the failure strain showed good agreement with reference expansion-due-to-compression tests at slow strain rate. However, a decrease in the failure strain was determined for the fast-rate (12 ms) tests. This indicated that the failure strain of these composites might be influenced by the strain rate during RIA-like events. The failure strains observed in the tests corresponded to local energy depositions of approximately 50 cal/g UO2 from hot zero power, with an initial condition of pellet–cladding gap closure prior to the event. In-pile transient testing of these concepts that would result in hoop strain due to PCMI in the range of 0.5–1.0% is recommended.

The impact of azimuthally asymmetric carbon deposition upon pellet-clad mechanical interaction in advanced gas reactor fuel

Cracked nuclear fuel pellets were modelled in the r-θ plane with an azimuthally varying clad surface temperature boundary condition using the PELICAN set of fuel performance models for the commercial finite element software, Abaqus. The temperature boundary condition was assumed to represent heat transfer impairment due to an azimuthally asymmetric carbon deposit on advanced gas-cooled reactor pins. The model predicts the radial and azimuthal displacement of the idealised fuel fragments, together with the resulting elastic, creep and plastic strains in the cladding. These were compared to simulations assuming a uniformly hot or cold boundary condition. Apart from a short period during the return to power from reduced power (70%) operation and outages, the hoop stress in the simulation with an azimuthally varying clad surface temperature was bounded by that of models with a uniform hot or cold surface temperature. The reduced stress was proposed to be due to the greater ability of the fuel fragments to relocate in order to accommodate changes to the power level. As a result, the creep strains in the model with an azimuthally varying clad surface temperature were lower than assuming either a uniform hot or cold boundary condition.

Fuel relocation recovery implementation in Bison

During the initial rise to power in light water reactors (LWR), thermal gradients within a pellet cause radial and axial cracks to form in the fuel. The effect of these cracks is to reduce the pellet-cladding gap and accelerate the interaction between the fuel and cladding. This process is known as fuel relocation and may also include contributions from pellet eccentricity and cladding ovality. Since the cladding experiences both elevated temperatures and high external pressure due to the coolant, the cladding typically creeps inward further reducing the pellet-cladding gap. Once the gap is closed and pellet cladding mechanical interaction (PCMI) begins, both the thermal and mechanical behavior of the fuel is affected. The compressive forces exerted on the fuel due to contact with the cladding cause the fractured fuel sections to move back toward their original position which is termed relocation recovery. A model for this phenomena is implemented in the Bison fuel performance code and applied to a set of validation test cases. The predicted fuel rod diameter is compared to experimental measurements to evaluate the influence of relocation recovery over a range of operating conditions.

Modeling of central void formation in LWR fuel pellets due to high-temperature restructuring

Abstract Analysis of the GRSW-A model coupled into the FALCON code is extended by simulation of central void formation in fuel pellets due to high-temperature fuel restructuring. The extended calculation is verified against published, well-known experimental data. Good agreement with the data for a central void diameter in pellets of the rod irradiated in an Experimental Breeder Reactor is shown. The new calculation methodology is employed in comparative analysis of modern BWR fuel behavior under assumed high-power operation. The initial fuel porosity is shown to have a major effect on the predicted central void diameter during the operation in question. Discernible effects of a central void on peak fuel temperature and Pellet-Cladding Mechanical Interaction (PCMI) during a simulated power ramp are shown. A mitigating effect on PCMI is largely attributed to the additional free volume in the pellets into which the fuel can creep due to internal compressive stresses during a power ramp.

Probabilistic view of SiC/SiC composite cladding failure based on full core thermo-mechanical response

As part of Accident Tolerant Fuel (ATF) campaign, Silicon carbide (SiC) is being considered as a promising alternative cladding material. A new numerical model for mechanical calculation of multi-layer claddings was established and implemented in FRAPCON 4.0. The upgraded FRAPCON 4.0 was used to simulate fuel performance of a Pressurized Water Reactor (PWR) using a two layer Composite/Ceramic SiC cladding design. A probabilistic assessment was carried out based on the full core simulation to determine the possibility for SiC material cracking, referred to as “failure” probability. The results indicated that the large stresses induced during the refueling period in SiC cladding can be mitigated if high conductivity fuel.e.g. UO2 with center void and BeO additive, is utilized. However, it was found that full core average leakage probability from failure of the monolithic SiC layer and clad fracture rate from failure of both layers were higher than the failure probability of current Zircaloy cladding. According to the failure risk map, Pellet Cladding Mechanical Interaction (PCMI) is the primary leading cause of failure. The result of the PCMI sensitivity study on the influence of relocation recovery on SiC cladding failure risk motivates the need to develop a more mechanistic fuel cracking model relevant to SiC based ceramic fuel/cladding system.

Multiphysics modeling of accident tolerant fuel-cladding U3Si2-FeCrAl performance in a light water reactor

The performance of the proposed advanced fuel-cladding system U3Si2-FeCrAl was studied based on the suggestion that U3Si2-FeCrAl system is a potential accident tolerant fuel candidate from a neutronics assessment. The U3Si2 fuel was reported with a number of advanced thermophysical properties and FeCrAl alloy was reported with a better oxidation resistance and a larger thermal neutron absorption cross section. This study presents the development of a model for U3Si2-FeCrAl combination system by the modification of our CAMPUS code. Then the fuel performance was compared with UO2-Zircaloy, UO2-FeCrAl and U3Si2-Zircaloy systems. And finally sensitivity analysis has been performed on the fuel input parameters of UO2 fuel and cladding thickness of FeCrAl cladding. The U3Si2-Zircaloy system was found to decrease the fuel centerline temperature dramatically but with a very early gap closure time due to the large thermal expansion coefficient of U3Si2 fuel, while the U3Si2-FeCrAl system was found to not only decrease the fuel centerline temperature a lot, but also further delay the gap closure time, which would delay the pellet-cladding mechanical interaction time. Sequentially, the fission gas release and plenum pressure were also found to be decreased a lot. Then the nuclear reactor safety is expected to be improved by applying the U3Si2-FeCrAl system.

Thermally-Conductive and\xa0Mechanically-Robust Graphene Nanoplatelet Reinforced UO2 Composite Nuclear Fuels

Low thermal transport behavior along the radial direction of nuclear fuel pellets and pellet-cladding mechanical interaction significantly impact fuel performance and the safety of current nuclear energy systems. Here we report a new strategy of advanced fuel design in which highly thermally-conductive and mechanically-robust graphene nanoplatelets are incorporated into UO2 fuel matrix to improve fuel thermal-mechanical properties. The 2D geometry of the graphene nanoplatelets enables a unique lamellar structure upon fuel consolidation by spark plasma sintering. The thermal conductivity along the radial direction of the sintered fuel pellets at room temperature reaches 12.7 and 19.1 wm−1K−1 at 1 wt.% and 5 wt.% loadings of the graphene nanoplatelets, respectively, representing at least 74% and 162% enhancements as compared to pure UO2 fuel pellets. Indentation testing suggests great capability of the 2D graphene nanoplatelets to deflect and pin crack propagation, drastically improving the crack propagation resistance of fuel matrix. The estimated indentation fracture toughness reaches 3.5 MPa·m1/2 by 1 wt.% loading of graphene nano-platelets, representing a 150% improvement over 1.4 MPa·m1/2 for pure UO2 fuel pellets. Isothermal annealing of the composite fuel indicates that the graphene nano-platelet is able to retain its structure and properties against reaction with UO2 matrix up to 1150 °C.

Effects of Zr-hydride distribution of irradiated Zircaloy-2 cladding in RIA-simulating pellet-clad mechanical interaction testing

Abstract A series of simulated reactivity-initiated accident (RIA) tests on irradiated fully recrystallized boiling water reactor Zircaloy-2 cladding has been performed by means of the expansion-due-to-compression (EDC) test method. The EDC method reproduces fuel pellet–clad mechanical interaction (PCMI) conditions for the cladding during RIA transients with respect to temperature and loading rates by out-of-pile mechanical testing. The tested materials had a large variation in burnup and hydrogen content (up to 907 wppm). The results of the EDC tests showed variation in the PCMI resistance of claddings with similar burnup and hydrogen content, making it difficult to clearly identify ductile-to-brittle transition temperatures. The EDC-tested samples of the present and previous work were investigated by light optical and scanning electron microscopy to study the influence of factors such as azimuthal variation of the Zr-hydrides and the presence of hydride rims and radially oriented hydrides. Two main characteristics were identified in samples with low ductility with respect to hydrogen content and test temperature: hydride rims and radial hydrides at the cladding outer surface. Crack propagation and failure modes were also studied, showing two general modes of crack propagation depending on distribution and amount of radially oriented hydrides. It was concluded that the PCMI resistance of irradiated cladding under normal conditions with homogenously distributed circumferential hydrides is high, with good margin to the RIA failure limits. To further improve safety, focus should be on conditions causing nonfavorable hydride distribution, such as hydride reorientation and formation of hydride blisters at the cladding outer surface.

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Abstract The ALCYONE multidimensional fuel performance code codeveloped by the CEA, EDF, and AREVA NP within the PLEIADES software environment models the behavior of fuel rods during irradiation in commercial pressurized water reactors (PWRs), power ramps in experimental reactors, or accidental conditions such as loss of coolant accidents or reactivity-initiated accidents (RIAs). As regards the latter case of transient in particular, ALCYONE is intended to predictively simulate the response of a fuel rod by taking account of mechanisms in a way that models the physics as closely as possible, encompassing all possible stages of the transient as well as various fuel/cladding material types and irradiation conditions of interest. On the way to complying with these objectives, ALCYONE development and validation shall include tests on PWR-UO2 fuel rods with advanced claddings such as M5® under “low pressure–low temperature” or “high pressure–high temperature” water coolant conditions. This article first presents ALCYONE V1.4 RIA-related features and modeling. It especially focuses on recent developments dedicated on the one hand to nonsteady water heat and mass transport and on the other hand to the modeling of grain boundary cracking-induced fission gas release and swelling. This article then compares some simulations of RIA transients performed on UO2-M5® fuel rods in flowing sodium or stagnant water coolant conditions to the relevant experimental results gained from tests performed in either the French CABRI or the Japanese NSRR nuclear transient reactor facilities. It shows in particular to what extent ALCYONE—starting from base irradiation conditions it itself computes—is currently able to handle both the first stage of the transient, namely the pellet-cladding mechanical interaction phase, and the second stage of the transient, should a boiling crisis occur. Areas of improvement are finally discussed with a view to simulating and analyzing further tests to be performed under prototypical PWR conditions within the CABRI International Program. M5® is a trademark or a registered trademark of AREVA NP in the USA or other countries.

Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimating clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations.While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL.Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. Based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.

Deformation behavior of recrystallized and stress-relieved Zircaloy-4 fuel cladding under biaxial stress conditions

ABSTRACT Pellet–cladding mechanical interaction (PCMI) under reactivity-initiated accident conditions may lead to the failure of high-burnup fuel rods. Biaxial stress states generated by PCMI in Zircaloy cladding may make the cladding more susceptible to failure. In this study, we investigated the deformation behavior of Zircaloy cladding under biaxial stress conditions based on the concept of contours of equal plastic work. The major axis angles of the initial work contours of recrystallized (RX) and stress-relieved (SR) specimens were investigated and it was found that the shapes of the initial work contours of these kinds of specimens were almost symmetric across the direction where the ratio of axial stress to circumferential stress is 1. The shapes of subsequent work contours tended to change for the RX specimen while being the same as the initial for the SR specimen, as deformation proceeded. It was suggested that the textures and slip systems in the RX and SR specimens affect their initial work contours while the slip system in the RX specimens and the residual strain in the SR specimens influence the subsequent work contours.

Extending the reactivity initiated accident (RIA) fuel performance code SCANAIR for boiling water reactor (BWR) applications

In this paper, capabilities of the SCANAIR transient fuel performance code are evaluated and extended for boiling water reactor (BWR) fuel low temperature cladding failure predictions and high temperature thermal hydraulics modelling. The SCANAIR code, developed by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN), is designed for modelling the behaviour of a single fuel rod during reactivity initiated accident (RIA) in a pressurized water reactor (PWR). In a previous study (Arffman et al., 2012), new BWR cladding material property correlations were developed and implemented into SCANAIR. Here, SCANAIR’s ability to predict BWR cladding failures due to pellet-cladding-mechanical interaction (PCMI) is evaluated by modelling the NSRR FK test series. SCANAIR is found to give correct predictions with reasonably good accuracy when applied to a larger dataset of several tests. As the standard thermal hydraulics model in SCANAIR is one-dimensional and able to model single phase coolant only, the simulation of a BWR RIA, the control rod drop accident, is not possible when the bulk boiling regime is reached. In this study, the code’s application field is broadened to consider bulk boiling in BWRs. In the chosen approach, SCANAIR is coupled with an external thermal hydraulics code. For that, VTT’s in-house general thermal hydraulics code GENFLO has been used. The first demonstration simulations show promising results.

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified-burst test instrument that simulates transient events with a pulse width from 10 to 300ms. This paper includes validation tests of unirradiated and pre-hydrided ZIRLO™ cladding tubes. A ZIRLO™ cladding sample with 168wt.ppm of hydrogen showed ductile behavior and failed at the maximum limits of the mechanical test setup; hoop strain to failure was greater than 9.2%. ZIRLO™ samples showed high resistance to failure even at very high hydrogen contents (1466wt.ppm). When the hydrogen content was increased to 1554wt.ppm, “brittle-like” behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature of tubes fabricated from FeCrAl, a candidate material for accident-tolerant cladding were conducted to reproduce the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of RIA integral tests for current cladding materials and for candidate accident-tolerant cladding materials.

Development of NUFORM3D module with FRAPCON3.4 for simulation of pellet-cladding mechanical interaction

In order to evaluate the PCMI (pellet-to-cladding mechanical interaction) behavior in fuel rods, a three-dimensional finite element (FE) module, termed NUFORM3D, was developed by KAERI. In the NUFORM3D module, a thermo-mechanical model considering elasto-plastic deformation and a contact model were formulated to simulate the PCMI behavior. The NUFORM3D module was implemented to compensate for the weaknesses of the analytical methods used in traditional fuel performance codes for PCMI simulations. In this work, NUFORM3D was coupled with FRAPCON-3.4, which is the NRC’s fuel performance code for a steady state, to evaluate the NUFORM3D module with an irradiated database. The REGATE experiment, involving the use of a power ramp database in a test reactor, was employed to evaluate the coupled code system. The coupled NUFORM3D was able to simulate bamboo-shaped deformation of cladding as well as stress concentrations in cladding due to pellet hourglass deformation in the power ramp. Secondary ridges of the cladding on the mid-pellet plane were found in the experimental results. To simulate the secondary ridge, a transversally fragmented pellet model was adopted. As a result, the coupled NUFORM3D module can improve the simulation capability of fuel performance codes for PCMI.

Development of a new Pellet-Clad Mechanical Interaction (PCMI) model and its application in ATFs

In this study, a non-rigid Pellet-Clad Mechanical Interaction (PCMI) calculation model was independently developed with the consideration of plenum spring. Based on the new PCMI model, the FROBA-ROD code initially developed for conventional fuel rods was modified to perform the behavior simulation for accident tolerant fuels (ATFs). The applicability and reliability of the modification was validated by the result comparison with a public reference. The influence of rigid/non-rigid PCMI model was investigated by the comparison of results obtained by the initial and modified FROBA-ROD. It was found that for conventional fuel element, the rigid hypothesis was reasonable to some extent while was not applicable any more for PCMI simulation of ATFs with ceramic claddings. The modified FROBA-ROD was used to perform the calculation for both the traditional fuel rod and the ATF with SiC cladding. The result indicated that the low creep rate of SiC clad can delay the gap closure and PCMI effectively, but also lead to a much higher contact pressure and a possible brittle rupture. In addition, the delayed gap closure resulted in a higher fuel temperature, a higher fission gas release fraction and a higher inner gas pressure.

Multi-Dimensional Simulation of LWR Fuel Behavior in the BISON Fuel Performance Code

Nuclear fuel operates in an extreme environment that induces complex multiphysics phenomena occurring over distances ranging from inter-atomic spacing to meters, and times scales ranging from microseconds to years. To simulate this behavior requires a wide variety of material models that are often complex and nonlinear. The recently developed BISON code represents a powerful fuel performance simulation tool based on its material and physical behavior capabilities, finite-element versatility of spatial representation, and use of parallel computing. The code can operate in full three dimensional (3D) mode, as well as in reduced two dimensional (2D) modes, e.g., axisymmetric radial-axial (R-Z) or plane radial-circumferential (R-θ), to suit the application and to allow treatment of global and local effects. A BISON case study was used to illustrate analysis of Pellet Clad Mechanical Interaction failures from manufacturing defects using combined 2D and 3D analyses. The analysis involved commercial fuel rods and demonstrated successful computation of metrics of interest to fuel failures, including cladding peak hoop stress and strain energy density. In comparison with a failure threshold derived from power ramp tests, results corroborate industry analyses of the root cause of the pellet-clad interaction failures and illustrate the importance of modeling 3D local effects around fuel pellet defects, which can produce complex effects including cold spots in the cladding, stress concentrations, and hot spots in the fuel that can lead to enhanced cladding degradation such as hydriding, oxidation, CRUD formation, and stress corrosion cracking.