Pellet-cladding Interaction Research Articles

Fuel pin failure modes in the UK's advanced gas cooled reactors

The current paper presents a review of key failure modes in AGR fuel, a description of failure mechanisms and the post failure effects that follow.The main failure modes of AGR fuel are found in the following proportions: ratchetting (axial clad straining and fuel-stacking with resulting fuel-stack gap formation and clad collapse) 3%; over pressure 4%; pellet clad mechanical interaction 7%; pin-brace interaction/rotating pins (fuel pin and support structure interaction, also known as fretting) 37%; clad restructuring/microstructural effects (including clad grain growth reducing ability of clad to withstand pellet clad interaction) 41%; other (including chipped pellets, endcap welds, and non-fuel component damage) 8%.A description of each failure mode is provided detailing the characteristics of the failure mode and a summary of the factors contributing to the failure mode.An overview of post-failure effects in AGR fuel is provided. Observations of failed fuel in hot-cells reveal a range of post-failure effects, in addition to the characteristics which contributed to the failure. Understanding post failure effects is important in order to isolate features which may have occurred prior to failure, and to inform on the behaviour of fuel between failure and discharge from reactor.

Thermomechanical modeling of pellet-cladding interaction using state-based peridynamics

Pellet-Cladding Interaction (PCI) is a significant concern for the safe and reliable design of nuclear fuel rods. This study presents a modeling approach for investigating PCI employing the Ordinary State-Based PeriDynamics (OSB-PD) while considering irregular discretization to improve the modeling of curved boundaries of fuel rods with geometric precision. Unlike the existing PD models for fuel pellet, the present study considers contact and heat transfer between fuel pellet and cladding. Also, it presents a new frictional contact model with stick and slide friction. The heat conduction is modeled through gas, contact and radiation between pellet and cladding. In order to capture complex fragmentation and failure in pellet, material variability is considered by applying randomized critical stretch values with normal distribution. Specifically, it presents the effects of friction coefficient, gap size, and power level on crack patterns in fuel pellets while considering PCI. The PD predictions show that the number of major radial cracks remains constant as the friction coefficient increases; however, the number of circumferential cracks increases significantly. As the gap size increases, both the temperature in fuel pellet and the temperature difference across the gap increase. Also, the number of cracks in fuel pellet increases with the increasing value of gap size. The power level has a strong influence on the temperature gradient of the fuel rod, and the number of cracks in fuel pellet directly correlates with the power level.

Hydrogen diffusion in zirconium cladding alloys with an inner liner as quantified by neutron radiography and nanoindentation

Embrittlement caused by hydride precipitates in zirconium claddings constitutes a significant risk factor for the integrity of spent nuclear fuel rods, especially during the handling and transport operations necessary before and after long-term intermediate dry storage. Zircaloy-2 is the most commonly used alloy for cladding tubes in boiling water reactors. It is typically manufactured with an inner liner to better cope with issues related to pellet-cladding interaction. Previous work showed that in clad samples possessing a liner, a significant portion of hydrogen tends to migrate towards the liner region during cool-down. In this work, the hydrogen precipitation and distribution after thermal exposures comparable to service was investigated in different commercial claddings for boiling water reactors. The resulting hydrogen distribution was quantified by means of high-resolution neutron imaging at the SINQ spallation neutron source at Paul Scherrer Institut in Switzerland, and the presence of hydrides was subsequently confirmed by metallography and nanoindentation mapping of selected samples. The results show that when a cooling rate of 30 °C/h is applied, the vast majority of the hydrogen tends to accumulate at the liner/substrate interface when a liner is present. At lower cooling rates, the hydrogen tends to distribute more homogenously into the liner material, and the cladding bulk material is left completely hydrogen-free. Neutron radiography of samples after water quenching from homogenization temperatures between 370 °C and 460 °C revealed a significant amount of hydrogen trapped in the hydrides form at the liner/substrate interface at temperatures up to 400 °C. Undissolved hydrides may act as nucleation seeds for hydrides upon cooling, contributing to the observed hydrogen segregation. A model that takes into account the interfacial properties of the region between the liner and the bulk material is proposed.

Diffusion in undoped and Cr-doped amorphous UO2

UO2 fuel pellets are often doped with chromium oxide to obtain favourable properties such as higher density, improved thermal stability, large grain sizes, improved pellet-clad interaction margins, and increased fission gas retention during transients. Chromium has a low solubility limit in UO2, with past experimental work reporting solubility limits ranging between 0.004 to 0.06 wt.% Cr. Due to its low solubility, segregation of Cr ions to the grain boundary may occur. Further, the complexity of these boundaries may be high as observed in other ceramics resulting in disordered or amorphous regions along the boundary, affecting a range of material and operational properties of the fuel pellet. To assess these disordered regions, in this work we study amorphous undoped and Cr doped UO2 systems (containing 10–50 at.% Cr3+) that have been modelled using classical molecular dynamics methods incorporating Cr3+ into the well-used CRG potential library. Diffusion coefficients, pre-exponential factors, and activation energies for diffusion were computed for oxygen ions, assessing the impact of structure and extrinsic species on migration. Oxygen diffusion was observed to be much faster in the undoped amorphous system compared to its crystalline counterpart. Oxygen diffusion in doped systems decreased with increasing Cr concentration, highlighting the importance of additives to retain fission products and other migratory species.

Exploring a surrogate of Pellet–Cladding interaction: Characterization and oxidation behavior

The present work analyzes the effect of different Zr contents on the microstructure and thermal behavior of non-irradiated UO2. A set of Zr-doped UO2 (0, 20, 40, 80, and 100 wt%) pellets was prepared via solid-state synthesis by mimicking the chemical bonding between ZrO2 and UO2. After sintering, the Zr-doped UO2 monoliths were characterized by scanning electron microscopy, Raman spectroscopy, and X-ray diffraction. Oxidation of the Zr-doped UO2 samples has been monitored by thermogravimetric analysis. Results of thermal behavior under air and low oxygen partial pressure show a profound effect of delayed oxidation with the addition of Zr to UO2 and then an increased oxidation resistance of UO2 to U3O8 when compared to pure UO2 pellet.Graphical abstract

A first principles study of zirconium grain boundaries

We present the results of first-principles calculations of selected structural and thermodynamic properties of a set of grain boundaries (GBs) in zirconium, spanning a range of misorientation angles and boundary planes. We performed plane-wave density functional theory calculations on low-Σ grain boundaries — five symmetric tilt GBs (STGBs) and three twist GBs; all with misorientation axes about [0 0 0 1] and in optimised microscopic configurations — to gain insight into the associated atomistic structures. We include in our analysis properties such as GB excess volume, interplanar spacing, volume per atom, electron density distribution, and a measure of GB width. We found the twist GBs to exhibit similar energetic and structural properties, whereas the STGBs showed substantially more variation. Our comprehensive analysis demonstrates how all five dimensions of GB space are crucial in determining properties such as the work of ideal separation and the length scale over which atoms are perturbed by the presence of the GB. Additionally, we compared our first-principles results to predictions from a widely used embedded-atom method (EAM) potential; we found the potential to perform generally well, particularly in predicting GB energy. We discuss the results in terms of representing the initial steps towards the development of a mechanistic understanding of the pellet-cladding interaction in nuclear fuel, which must involve encoding accurate materials behaviour at larger length and time scales. To this end, and so that our results can be useful for further investigations, we have published our data (including optimised structures, computed interface energetics and structural properties) to a public repository.

Multi-physics framework for whole-core analysis of transient fuel performance after load following in a pressurised water reactor

The increasing deployment of renewable energy sources and greater electrification of demand is requiring more frequent use of load following in pressurised water reactors (PWRs) . A limiting aspect with respect to load following is the thermo-mechanical behaviour of the fuel. Hence a greater use of best-estimate multi-physics tools that can more accurately capture such behaviour during load following manoeuvres – in particular, extended reduced power operation (ERPO) – has become important in recent years. To this end, we have developed a coupled, whole-core analysis framework and demonstrated it with a new generic PWR model. The demonstration is for uncontrolled RCCA bank withdrawal at power (URWAP) faults following a period of ERPO, since such scenarios are often limiting with respect to fuel integrity. The coupled, whole-core analysis framework that was developed consists of the PARCS neutronics code, the RELAP and CTF thermal–hydraulics codes, the ENIGMA fuel performance code, and the NEXUS platform for whole-core fuel performance. The generic PWR model constructed is for a 1240 MWe plant with a modern core design (maximum assembly burnup of 51.0 GWd/tHM, cycle length of 16 months, and an average fuel residence time of 2.3 cycles per assembly). Previous work in the literature on URWAP faults has focused on coupled neutronics and thermal–hydraulics analysis, with fuel performance assessment and the effects of load following neglected. These shortcomings are therefore addressed here. From a fuel performance perspective no fuel was predicted to fail due to pellet-clad interaction (PCI). Of the remaining failure indicators, margin to clad yield stress was most limiting, although no failure was predicted in all cases under the scenarios considered.

Fuel Performance Modeling at High Burn-Up by FEMAXI-6 Code

Abstract The scope of the current research in the field of fuel performance is primarily aimed at an improvement of the operating reliability, safety, and cost-effectiveness of the reactors in operation. The current requirement of nuclear industry is to have fuel suitable for load follow operation. Fission gas release, pellet-cladding mechanical interaction, and stress corrosion cracking are the main phenomena that limit the variability of reactor operation from a safety perspective. To reasonably predict the fuel performance limits, it is necessary to benchmark the computational tools against high-quality experimental data. This work is devoted to the calculation of fuel performance using the code FEMAXI-6 based on the longest irradiation experiment in the Halden reactor. The fuel burn-up was approaching 90 MWd/kgUO2 in three selected rods, which were equipped with the pressure sensors and were subjected to extensive postirradiation examination. During the experiment, the rods were exposed to several periods of power cycling. The rods were manufactured with different fuel grain size and fuel-to-clad gap size.

Post-transient examination of performance of uranium silicide fuel and silicon-carbide composite cladding under reactivity-initiated accident conditions

• First transient testing of accident-tolerant fuel designs under transient nuclear heating reactivity-initiated accident conditions. • Claddings experienced dimensional changes but maintained primary rod-like geometries. • Elemental species diffusion from the U 3 Si 2 fuel to the zircaloy cladding was observed. • No elemental species diffusion from the U 3 Si 2 fuel to the silicon-carbide cladding was observed. • Through wall thickness cracking observed in the silicon-carbide cladding resulting from pellet cladding interaction from thermal expansion of the pellet. New fuels are being developed for light water nuclear reactors with the initial goal of improving accident tolerance. More recently, these new fuel systems are also being recognized for their potential to support higher fuel burnups and thus improved economics for reactor plants through improved fuel utilization. Fuel safety testing of these new fuels systems is being conducted under transient irradiation conditions in the Transient Reactor Test facility at Idaho National Laboratory. A test campaign including two rodlets of U 3 Si 2 fuel in Zircaloy-4 cladding and two rodlets of U 3 Si 2 fuel in Silicon-Carbide Composite cladding has been performed to begin the development of fuel-safety criteria in design-basis reactivity-initiated accident conditions. Post-Transient-Irradiation Examinations have been performed and found that the accident tolerant materials maintained their primary rod-like geometry. Elemental species diffusion was found to occur in the Zircaloy rodlets indicating rapid diffusion kinetics at the temperatures achieved. No elemental species diffusion was found in the Silicon-Carbide rodlets. However, through-wall thickness cracks were observed due to displacement loading of the cladding from thermal expansion of the fuel pellets. This work presents the first experimental evidence of the performance of these fuel systems under transient nuclear heating and representative reactivity-initiated accident energy depositions.

Structure of the pellet-cladding interaction layer of a high-burnup Zr-Nb-O nuclear fuel cladding

Structure of the pellet/cladding-interaction (PCI) layer of a high-burnup nuclear fuel with niobium-bearing cladding was investigated using modern transmission electron microscopy (TEM) techniques. Morphology of the PCI consisted of uniform zirconia layer on the cladding side and finger-like features towards to the ceramic fuel. The PCI layer showed a complex microstructure that constitutes three distinct zirconia layers formed by two zirconia phases. From cladding to fuel, monoclinic, tetragonal, and another tetragonal phase were determined via TEM diffraction patterns and precession electron diffraction (PED) using ASTAR microscopy platform (AMP). Characterizations showed that monoclinic and tetragonal phases close to the cladding were crack-free, but the tetragonal layer adjacent to the fuel contained cracks and pores. Presence of the monoclinic layer suggested that its formation involved the gaseous diffusion of oxygen below threshold dose of fission product damage before the fuel/cladding contact occurred. Furthermore, high-resolution TEM revealed compound phase of (U,Zr)O2 at fringes of zirconia fingers on the fuel side.

Atomistic modelling of iodine-oxygen interactions in strained sub-oxides of zirconium

In water reactors, iodine stress corrosion cracking is considered the cause of pellet-cladding interaction failures, but the mechanism and chemistry are debated and the protective effect of oxygen is not understood. Density functional theory calculations were used to investigate the interaction of iodine and oxygen with bulk and surface Zr under applied hydrostatic strain (−2% to +3%) to simulate crack tip conditions in Zr to ZrO2, using a variety of intermediate suboxides (Zr6O, Zr3O, Zr2O and ZrO). The formation energy of an iodine octahedral interstitial in Zr was found to decrease with increasing hydrostatic strain, whilst the energy of an iodine substitutional defect was found to be relatively insensitive to strain. As the oxygen content increased, the formation energy of an iodine interstitial increased from 1.03 eV to 8.61 eV supporting the idea that oxygen has a protective effect. At the same time, a +3% tensile hydrostatic strain caused the iodine interstitial formation energy to decrease more in structures with higher oxygen content: 4.56 eV decrease in ZrO compared to 1.47 eV decrease for pure Zr. Comparison of the substitutional and interstitial energies of iodine, to the adsorption energy of iodine, in the presence of oxygen, shows the substitutional energy of iodine onto a Zr site is more favourable for all strains and even interstitial iodine is favourable between strains of +1-5%. Although substitutional defects are preferred to octahedral interstitial defects, in the ordered suboxides, a 3% tensile strain significantly narrows the energy gap and higher strains could cause interstitial defects to form.

Simplified Model of a High Burnup Spent Nuclear Fuel Rod under Lateral Impact Considering a Stress-Based Failure Criterion

The inventory of spent nuclear fuel (SNF) generated in nuclear power plants is continuously increasing, and it is very important to maintain the structural integrity of SNF for economical and efficient management. The cladding surrounding nuclear fuel must be protected from physical and mechanical deterioration, which causes fuel rod breakage. In this study, the material properties of the simplified beam model of a SNF rod were calibrated for a drop accident evaluation by considering the pellet–clad interaction (PCI) of the high burnup fuel rod. In a horizontal drop, which is the most damaging during a drop accident of SNF, the stress in the cladding caused by the inertia action of the pellets has a great effect on the integrity of the fuel rod. The failure criterion for SNF was selected as the membrane plus bending stress through stress linearization in the cross-sections through the thickness of the cladding. Because the stress concentration in the cladding around the vicinity of the pellet–pellet interface cannot be simulated in a simplified beam model, a stress correction factor is derived through a comparison of the simplified model and detailed model. The applicability of the developed simplified model is checked through dynamic impact simulations. The developed model can be used in cask level analyses and is expected to be usefully utilized to evaluate the structural integrity of SNF under transport and in storage conditions.

Simulation of dynamic characteristics of NHR200-II fuel assembly

NHR200-Ⅱ is a 200 MW nuclear heating reactor. The integrity of the fuel assembly should be ensured during the normal operations and design basis accidents. In the latter cases (e.g. earthquake), its deformation or damage should not prevent the insertion of control rods to shut down the reactor. Meanwhile, during the operating time, there are two aspects that attention should be paid, (i) the mechanical performance of internal core materials might be worsened by the increase of burnup and (ii) dynamic loads involving hydraulic erosion, flow-induced vibration might damage the assemblies. Therefore, the investigation of the dynamic characteristics of the fuel assembly with consideration of burnup effects is necessary. The fuel rod with pellets and cladding was approximated by the beam elements with a circular cross-section applied with calibrated material properties. Connector elements were used to replace the three-arc springs and dimples on the grid and thus simplify the geometry of the spacer grid. The agreements between the detailed and simplified models indicate the equivalence of these two methods. Then two simulations were performed to evaluate the dynamic characteristics of NHR200-II fuel assembly at the beginning of life. The modal simulation of the fuel bundle and channel suggests that the channel could be assumed stationary to the fuel bundle during the dynamic analysis. The seismic simulation conducted in the specific boundary conditions indicates that the integrity of spacer grids can be guaranteed and there was no buckling of the gird. No permanent deformation of the channel was observed, and its relative displacements were sufficiently small to ensure the insertion of control rods. Finally, burnup effects on the structural behavior of the fuel rod are discussed. The initial conditions of NHR200-II fuel assembly and its material properties at different burnup times are provided by referring to two irradiated tests. Various sensitivity analyses had been conducted to evaluate the effects of material properties and pellet-cladding interaction on the resulting failure limit and stiffness of the fuel rod.

Atomistic Level Molecular Dynamics Analysis and its Integrity in the Interim of Irradiation

In this work, molecular dynamics (MD) simulation was utilized in relation to access the thermal conductivity of UO2, PuO2 and (U, Pu)O2 in temperature range of 500–3000 K. Diffusion study on mixed oxide (MOX) was also performed to assess the effect of radiation damage by heavy ions at burnup temperatures. Analysis of the lattice thermal conductivity of irradiated MOX to its microstructure was carried out to enhance the irradiation defects with how high burnup hinders fuel properties and its pellet-cladding interaction. Fission gas diffusion as determined was mainly modelled by main diffusion coefficient. Degradation of diffusivity is predicted in MOX as composition deviate from the pure end members. The concentration of residual anion defects is considerably higher than that of cations in all oxides. Depending on the diffusion behavior of the fuel lattice, there was decrease in the ratio of anion to cation defects with increasing temperature. Besides, the modern mixed oxide fuel releases fission gas compared to that of UO2 fuel at moderate burnups.

Fuel rod nonlinear vibrations to detect and characterize Pellet-Cladding Interaction

Pellet-Cladding Interaction (PCI) in a Light Water Reactor is one of the major concerns to guarantee clad integrity while attempting at increasing the flexibility of PWR nuclear reactor operations to follow power grid demand. In order to forbid operations leading to clad failure, modeling capability to simulate the mechanism has improved through the years. Code development needs detailed and precise experimental data. Those data result from dedicated irradiation programs, named “power ramp tests”, carried out in experimental devices in Material Testing Reactors (e.g. in France ISABELLE in OSIRIS in the past (Alberman et al., 1997), the ADELINE loop in the Jules Horowitz Reactor JHR in the next future (Cheymol et al., 2011)). Those irradiation devices are highly instrumented to collect the most relevant information with the highest possible accuracy. In parallel with information gained thanks to post-irradiation examination programs, research is working on innovative methodologies to detect and characterize PCI kinetics during the tests. In this frame, we investigate the technological feasibility to measure the effects of PCI, on the vibrations of a nuclear fuel rod externally submitted to the turbulent axial flow rate excitation. We designed and realized the out-of-pile IMPIGRITIA set-up to reproduce, in controlled laboratory conditions, the mechanical interaction originating in the nuclear reactor. A single PWR rod test section, reproducing the main relevant geometrical and material characteristics of ADELINE experimental system, is submitted to the hydraulic excitation representative of the real case. The local closure of the gap is obtained by means of a remotely controlled expansion system. Measurements of the transverse displacement of the sample rod are collected by Laser Doppler Vibrometry. In this paper we introduce the design of the mock-up and the associated measurement method, then we present the two experimental phases and their results: the first one in air, to show the feasibility of the measurement in air in controlled conditions; the second one under turbulent flow rate to state on the feasibility of the passive detection. At the end, conclusions and perspectives for the work are discussed.

Mechanical failure of high-burnup fuel rods with stress-relieved annealed and recrystallized M-MDA cladding under reactivity-initiated accident conditions

ABSTRACT To evaluate the effects of the hydride morphology and initial temperature of fuel cladding on the pellet-cladding mechanical interaction failure under reactivity-initiated accident (RIA) conditions, RIA-simulated experiments were performed on high-burnup fuels with stress-relieved annealed (SR) and recrystallized (RX) M-MDATM cladding at room and high (~280°C) temperatures. The results demonstrated that the failure-limit trend of RX-cladded fuels being lower than that of SR-cladded fuels for a similar hydrogen content holds up to at least about 700 wtppm. The observation of the fracture surfaces of failed RX cladding suggests a contribution of radially oriented hydrides to the crack formation and/or penetration, which coincides with the aforementioned failure-limit trend. The temperature effect, namely the failure-limit rise at a high temperature, is evident irrespective of the hydride morphology, while the degree of the temperature effect decreases as the hydrogen content increases.

Nusselt correlation development in unsteady laminar gas flows for CABRI multiphysic simulations with CATHARE2

CABRI is an experimental pulse reactor, which aims at studying and thus at better understanding RIA (Reactivity Initiated Accident) effects on nuclear fuels. The restart of the CABRI reactor in 2015 offers the opportunity to validate tools involving multiphysic calculation schemes on reactivity insertions (RI) transients at a system scale. The reactivity insertion in CABRI is mastered by the depressurization of a neutron absorber (3He), contained into transient rods. This enables the realization of various types of transients. As this depressurization highly influences the reactivity insertion kinetics, it is essential to accurately simulate the 3He density evolution which depends on the heat exchanges inside the circuit. The challenge is thus to manage the simulation of the various CABRI complex transients. Previous work allowed to model with CATHARE2 the whole reactor with its specific phenomena as well as phenomena generally involved in RIA, for instance strong neutronic feedback effects, effect of the pellet-clad interaction on the gap conductance and transient heat exchanges between fuel rods and water. These studies also revealed that some CABRI transients were still difficult to model with CATHARE2 and this paper shows the necessity to catch the transient heat exchanges in the transients rods for the simulation of the power pulse in CABRI reactor. In order to achieve this objective, a correlation for the Nusselt evolution in an unsteady compressible laminar 3He flow, inside a closed tube, has been realized from numerical resolution of an analytical model and implemented in the scientific calculation tool CATHARE2. This paper describes this analytical model and its numerical resolution. After having been validated on results of the literature in steady flow inside open tube, the correlation established for the evolution of the Nusselt in a transient rods’ tube is given. Finally, Best-Estimate simulation results of CATHARE2 are compared with experimental data obtained on CABRI commission tests: core power and 3He pressure evolution.

Optimal operation strategy of LWR based on the PCI online prognosis model

Abstract The pellet-cladding interaction (PCI) is one of the major issues in fuel rod during the power change operation in light water reactors (LWR). The frequent power transient will lead to the PCI failure, which results in the radioactive leakage. Unfortunately, the failure cannot be detected on time during operation. For solving this problem, the PCI results is calculated and predicted before the power transient, and the results is used for optimizing the operation strategy. As everyone knows, the calculation of the fuel is a cumbersome work, which is not fit for online evaluation. At first, we present a PCI online prognosis model in this paper by the method of radial basis function neural network (RBFNN). Based on this online model, an optimal operation strategy is present then. The PCI evaluation results is treated as the feedback of the strategy, which will go to the power control system to decide to carry out the power plan or to modify it. The presented operation strategy used the fast online prognosis model to predict the PCI status, which is used as a feedback signal to the power control system. The optimal operation strategy is tested by the experimental data from the references, and the results demonstrate its effectiveness.

A new PCI analysis methodology based on artificial neural network

ABSTRACT Pellet-Cladding Interaction (PCI) is one of the most important issues in fuel rods of a Pressure Water Reactor (PWR) which may cause the failure of fuel rods, and it should be analyzed under operating transients to ensure the fuel rods’ integrity. The traditional PCI analysis method simulates the detailed physical phenomena that occurred in fuel rods and it considers the combined effects of irradiation, thermodynamics, and mechanics, which consumes a large amount of time and computational resources. A two-stage artificial neural network model has been developed by China General Nuclear Power Corporation (CGN), which provided a new solution for the PCI analysis. This approach employs two data-driven models basing on artificial neural networks. It takes a very small amount of time and computational resources, while the accuracy is comparable to the traditional method.

Initial development of an RIA envelope for dispersed nuclear fuel

A reactivity insertion accident (RIA) is a design basis accident in which reactivity is rapidly injected as a result of a control rod ejection or blade drop scenario. The resultant power increase can result in fuel rod failure and subsequent release of radioactive material into the reactor’s primary system. For light water reactors, failure under RIAs is typically a result of fuel melt, pellet-cladding interaction, or rod over-pressurization. Fuel melt occurs when the fuel system is unable to transport heat out of the fuel system. Pellet-cladding interaction occurs when an aggressive fuel pellet expansion pushes on an embrittled cladding beyond its strain limit. Rod over-pressurization occurs when a fuel rod exceeds the critical heat flux and departs from nucleate boiling, causing the rod internal pressure to rapidly increase, exceeding the systems pressure, and balloon until it bursts. To prevent failure, regulators have developed safety requirements that limit the injected enthalpy based on the state of the fuel system. However, dispersed fuel could enable the removal or relaxation of regulator-imposed safety criteria. Dispersed fuel embeds fuel particles in a highly conductive metal. Dispersed particles can have a particle radius of 1 µm to greater than 100 µm. Traditional fuel systems are plagued by poor thermal conductivity, especially at higher burnups, thus reducing the ability of the fuel system to respond to an RIA. However, dispersed fuel could improve the ability of the fuel system to transport injected energy away from the fissile material and into the coolant more efficiently. Additionally, dispersing the fuel particles mitigates hard contact that typically occurs in the traditional zirconium-uranium dioxide fuel system, thus removing pellet-cladding mechanical interaction as a potential failure mechanism. This paper evaluates fuel particles dispersed in a metal matrix using the BISON fuel performance code to develop an initial failure threshold based on melt temperature of the fuel particle and/or cladding material as a function of fission density for beginning-of-life and end-of-life conditions. BISON results show that (1) reducing the size of the particles allows fission density to increase as a function of time, (2) particle-to-particle proximity must be considered to evaluate the limiting conditions, and (3) improving the fuel particle thermal conductivity improves thermal performance.