Fuel Performance Code Research Articles

An application of the DEM-FFT method to predict the thermal conductivity of high burnup fragmented fuel

Understanding Fuel Fragmentation, Relocation and Dispersal (FFRD) phenomena is a major item to predict the behavior of fuel rods in the event of a Pressurized Water Reactor loss-of-coolant hypothetical scenario. In order to introduce an effective medium for fragmented fuel that could be used in fuel performance codes, the heat transfer properties of UO2 granular beds in He-Kr-Xe mixes was studied numerically. The Discrete Element Method was first used to compute the spatial arrangement of powder only submitted to gravity with a representative granulometric distribution. These generated geometries were then transformed into a 3D grid (voxelized) to perform numerical homogenization and compute their effective properties using a Fast Fourier Transform solver for heat transfer. Different hypotheses were made in the voxelation process about how the treatment of gas-solid and solid-solid interfaces, which in turn provided different estimates of the effective properties. Due to limits on the discretization and extended particle size distributions, a two-scale scheme was introduced and numerically tested against direct computations. For several beds, computed thermal conductivities were compared to analytical models for granular materials. Some heat transfer properties of fragmented fuel at several burnups were then computed, taking into account approximations for Knudsen and radiation size effects.

On the theory of the nucleation of gas bubbles at grain boundaries and incoherent inclusions

On the base of the critical analysis of two-dimensional models of the nucleation of gas filled bubbles at grain boundaries of helium-implanted specimens under the action of tensile stresses, a new model is developed within the framework of the Reiss theory of homogeneous nucleation in binary systems. This approach considers that gas bubbles are formed as a result of agglomeration in a binary system of vacancies and gas atoms at grain boundaries, avoiding significant simplifications of previous models based on the classical nucleation theory for single-component (unary) systems. The new model is extended to consider the nucleation of Xe bubbles at grain boundaries in UO2 under irradiation conditions and can be used for numerical analysis of experimental observations after the foreseen implementation in a fuel performance code. A similar approach can be applied to the nucleation and growth of gas bubbles on incoherent inclusions, such as those observed in irradiated ODS steels.

Fuel performance modelling of Cr-coated Zircaloy cladding under DBA/LOCA conditions

Chromium coatings are being developed for advanced technology fuel (ATF) claddings, offering negligible corrosion during normal operation, improved resistance to high-temperature steam oxidation, and superior high-temperature strength, the latter two being of utmost relevance during design basis accidents (DBAs). Demonstrating the improved response of Cr-coated Zircaloy requires the development or extension of fuel performance codes to coating simulations.In this work, material models and correlations for Cr-coated Zircaloy cladding have been derived or obtained from the literature and implemented into TRANSURANUS and the FRAPTRAN-TUmech suite. These extended tools have been used to simulate two complementary LOCA tests: QUENCH-L1 rod 4 (out-of-pile bundle test on fresh Zircaloy cladding) and IFA-650.10 (in-pile single rod test on high-burnup Zircaloy-UO2 fuel), enabling a gradual cross-verification of results between codes and a comparative performance analysis between coated and uncoated cladding.The results indicate negligible impact of coating properties other than creep on the burst time. While the superior high-temperature creep resistance of coated cladding slightly delays the burst time, additional burst data would be necessary to draw sound conclusions on the balloon size. Regarding the modelling approach, treating the coated cladding as a composite material through the definition of effective properties might result in worse performance relative to uncoated cladding, contradicting experimental observations. Therefore, the separate modelling of the coating and the cladding is recommended.

Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles

Tristructural isotropic (TRISO) particles are under consideration for use in several proposed advanced nuclear reactor concepts. The silicon carbide (SiC) layer in TRISO acts as a barrier to prevent the release of the fission products. However, despite remarkable retention, silver (Ag) release has been observed from intact particles, which requires investigation since the Ag isotope (110m Ag) has a long half-life. Previous work focused on developing a multiscale, mechanistic model for Ag diffusion accounting for temperature and microstructure effect and has been successfully validated. In this work, we expand the previous model to account for irradiation-enhanced Ag diffusivity in SiC and improve its accuracy over a wider grain size and temperature ranges relevant for advanced reactor conditions. A temperature, grain size, and flux dependent diffusivity is therefore derived using the mesoscale code MARMOT and implemented in the fuel performance code BISON. The irradiation-enhanced Ag diffusivity in SiC is compared against experimental data and validated using BISON against Ag release measurements from the Advanced Gas Reactor Fuel Development and Qualification Program (AGR-1 and AGR-2). Herein, we quantify the impact of SiC grain size, irradiation, and temperature on Ag release. In agreement with previous studies, we find accounting for SiC grain size improves agreement between BISON predictions and experimental observations for most cases. We also find that accounting for irradiation improves agreement for cases where Ag release was underestimated, but the impact was less significant than accounting for microstructure.

Incorporation of uranium nitride fuel capability into the ENIGMA fuel performance code: Model development and validation

Uranium dioxide (UO2) is the primary fuel form for nuclear reactors but its moderate uranium density and low thermal conductivity have prompted the exploration of alternative materials. Uranium nitride (UN) has emerged as a promising candidate for a variety of reactors, offering higher uranium density and thermal conductivity. This paper details the development and implementation of UN fuel capabilities within the ENIGMA fuel performance code for Light Water Reactor (LWR) applications. The new UN capability in ENIGMA includes correlations for theoretical density at room temperature, thermal conductivity, specific heat capacity, enthalpy, thermal expansion strain, Young’s modulus, Poisson’s ratio, thermal creep strain rate, irradiation creep strain rate and emissivity. Additionally, it incorporates models for densification, solid fission product swelling, fission gas bubble swelling, and fission gas release, along with a modified RADAR model for determining the pellet radial power profile and helium generation. Validation of the UN model was conducted using data from the L414 pin irradiation in the JOYO fast reactor in Japan. Further validation efforts are planned using datasets from JOYO and the Siloé thermal reactor in France. The paper also outlines areas of future work to address experimental data gaps and enhance model accuracy to cover a broader range of cladding materials, manufacturing parameters (including porosity volume fraction) and operating conditions (including fuel temperatures and burnups). Although focused on LWR applications, the work outlined supports the use of UN fuel across various reactor systems.

Modeling and design of a separate effects irradiation test targeting fission gas release from Cr-doped UO2

Fission gas release (FGR) from nuclear fuel during operation can diminish heat transfer properties across the pellet-cladding gap and increase the fuel rod internal pressure, thereby posing a concern to fuel reliability and safety during an accident. Enlarging the fuel grain size, which has been shown to improve fission gas retention, can be achieved by doping the fuel feedstock prior to sintering. In this work, the BISON fuel performance code was used to predict FGR from undoped and chromia-doped UO2 (referred to as Cr-doped UO2) fuel specimens with different grain sizes and across various temperatures. The BISON models identified the irradiation conditions for which FGR is most significant, and a separate effects irradiation experiment in the High Flux Isotope Reactor (HFIR) was then developed targeting those conditions. The experiment leveraged the MiniFuel irradiation capability at Oak Ridge National Laboratory and consisted of 12 fuel specimens of varying grain size and Cr content. A coupling scheme between BISON FGR results and the ANSYS finite element thermal model used for experiment design was formulated to predict cumulative FGR from each fuel specimen based on expected irradiation temperature histories. The fuel samples were fabricated and characterized as a part of this work, and the fuel compositions modeled in BISON were representative of the specimens used in the experiment. This combined modeling and experimental effort aims to study the effect of fuel grain size and Cr content on FGR and to provide simulated BISON FGR results that can be used for future model validation activities.

Deep heterogeneous joint architecture: A temporal frequency surrogate model for fuel performance codes

Fuel performance codes, such as Transuranus, predict fuel behavior and are used to ensure the safe operation of nuclear reactors. These codes are moderately time-consuming and affordable in many applications but may be limited in others, primarily when many fuel rods must be evaluated simultaneously. This work presents how the temporal neural network techniques, Temporal Convolutional Networks, and a Fourier Neural Operator can be combined to form a deep heterogeneous joint architecture as a surrogate model for fuel performance modeling in time-critical situations. We train the model using realistic power histories and corresponding outputs generated using the fuel performance code Transuranus. The ultimate result is a surrogate model for use in time-critical situations that take milliseconds to evaluate for thousands of fuel rods and have a mean test error of unseen data around a few percent.

Modeling oxygen transport in Cr doped UO2 fuel with the TAF-ID during power transients

This paper proposes an efficient coupling between an oxygen redistribution model and the Thermodynamics for Advanced Fuels-International Database (TAF-ID) to describe the behavior of fission products, actinides, and dopant in a uranium dioxide matrix. The formulation of oxygen migration in the fuel is a function of the oxygen (and uranium) chemical potential gradient. The proposed formulation does not require the introduction of the Soret term and relies solely on the known mobilities of uranium and oxygen in the fuel. This coupling is applied to the simulation of a power ramp on a chromia-doped fuel rod using the PLEIADES/ALCYONE fuel performance code. The simulations successfully reproduce most of the available experimental observations on the fission products (Cs, Mo, Ru) and on the chromium dopant.

Sensitivity analysis and fuel performance of a control rod withdrawal in the generic fluoride-salt-cooled high temperature reactor

The purpose of this study is to develop boundary conditions with respect to input parameter uncertainty for fuel performance assessment of a control rod withdrawal transient in a fluoride-salt-cooled high-temperature reactor (FHR). Sensitivity analysis was performed to determine how certain key input parameters impacted the uncertainty of the outputs needed for the boundary conditions. The generated boundary conditions can be used in fuel performance codes to obtain failure probabilities and fuel performance envelopes. To accomplish this, we utilized a KP-AGREE model of the generic fluoride-salt-cooled high-temperature reactor (gFHR) and a pebble bed benchmark developed by Kairos Power. The key boundary conditions that were developed included the total power, maximum fuel region temperature, maximum fuel kernel temperature, and maximum moderator (graphite) temperature within the core. The sensitivity analysis was run until the Sobol indices and output converged within a 95 % confidence level; it was found that the coolant inlet temperature consistently had the highest impact on the uncertainty of the key core temperature values. The maximum bound of these temperatures also did not exceed 1123.5 °C for the fuel kernel and 973.7 °C for the moderator after the transient, demonstrating that there is a large margin between the predicted fuel and moderator temperature and accepted safety limits, given a 1650 °C limit for Tristructural Isotropic (TRISO) fuel. These boundary conditions were implemented into a BISON model to demonstrate their use in fuel performance and obtain a brief failure profile of the fuel. The maximum magnitude of the hoop stress on the SiC layer was −12.02 MPa, resulting in a failure probability of 4.90·10−19. Thus, the fuel performance simulations also emphasized the robustness of the TRISO fuel design utilized in the gFHR, as the predictions reveal substantial margin of the TRISO particles relative to temperature limits during the control rod withdrawal. The limiting temperature is expected to be the reactor vessel temperature and the fuel is not expected to be challenged at all by mechanical failure of the SiC layer during these events.

Integral-scale validation of the SCIANTIX code for Light Water Reactor fuel rods

Mechanistic multi-scale modelling holds the potential to inform fuel performance codes by incorporating high-fidelity models, algorithms, parameters, and material properties. In this context, meso-scale codes emerge as valuable tools for developing detailed models and performing separate verification and validation steps. This work focuses on SCIANTIX, an open-source 0D meso-scale code designed to describe the behaviour of gaseous and volatile fission products in nuclear oxide fuel. The code predominantly employs engineering physics-based behavioural models featuring computational times that align with typical fuel performance code requirements. Given the numerical foundation of the code, it is applicable to both stationary and transient conditions. Following a recent work outlining the standalone SCIANTIX (version 2.0) performance and its separate-effect validation database, we present its performance when coupled with fuel performance codes to simulate light water reactor fuel rods. The experiments selected for the comparative analysis constitute an initial integral validation database. The comparison focuses on conventional engineering quantities of interest, such as integral fission gas release, demonstrating the satisfactory performance of the code. Additionally, it highlights the potential advantages of multi-scale modelling over conventional semi-empirical approaches.

Advances from R2CA project on reactor simulations for burst rod number evaluation during LOCA

In the frame of the “Reduction of Radiological Consequences of design basis and extension accidents” European Union’s Horizon 2020 project, a specific effort was focused on the update and the development of methodologies to assess the radiological consequences of loss-of-coolant accidents. Evaluation of the radiological consequences associated to this accident is strongly linked to the prediction of the rod burst ratio and new approaches were investigated through research and development associated to accident simulation software. Complementarily to approaches chaining system thermalhydraulic simulation to fuel performance code, approaches with integral codes ASTEC, DRACCAR and ATHLET-CD were developed respectively by ENEA, IRSN and HZDR. These applications were demonstrated on LOCA simulation for PWR and highlighted the capabilities of the tools. The ASTEC PWR model was extended by ENEA, which proposes a 2D core model with an increased number of representative fuel rods. HZDR and IRSN proposes approaches based on 3D core model which are able to capture distinctive fuel assembly responses and predict the rod burst ratio according to the distribution of rod behaviors within the core. This work highlighted several possible core models to predict the rod burst ratio and which can be used in integral tools abled to couple thermal hydraulics and thermomechanics. Advanced core models simulating the thermomechanical response of several representative rods per fuel assemblies and using 3D thermal hydraulics model are recommended due to the heterogeneities on power distribution, which impacts flow distribution. The need to model RPV in 3D was underlined by ATHLET-CD large break loss-of-coolant demonstrative case, which exhibits non-symmetric core heat-up.In addition, demonstrative cases proposed by ENEA and IRSN compared predictions obtained with the specific burst criteria developed in the frame of R2CA project and more classical one devoted to the core coolability assessment. The strong sensitivity of the rod burst ratio prediction to the burst criteria was highlighted. The need to handle uncertainties within a global methodology for rod burst ratio evaluation was underlined.Finally, throughout the development of loss-of-coolant accident applications with ASTEC, DRACCAR and ATHLET-CD, some prospects of development were identified for the codes and should bring advances for LOCA simulation and RBR predictions.

Assessing the impact of DIONISIO-SubChanFlow code coupling in nuclear fuel performance simulations

Realistic simulation of nuclear fuel performance requires not only validated models capable of describing the thermomechanical phenomena that take place within the fuel under irradiation conditions, but a detailed description of the thermal hydraulics of the channel surrounding the fuel rods, which provides the boundary conditions of the system. In this work, the main results and outlooks of coupling the thermal hydraulics code SubChanFlow with the fuel performance code DIONISIO are presented. To achieve this, an internal coupling was implemented, wherein DIONISIO is used as a master code controlling SubChanFlow as a thermal hydraulics subroutine replacing the simplified version already embedded in DIONISIO. Several tests were conducted to ensure the performance and quality of the coupling under normal operation conditions as a first approach. In addition, it was observed that the coupling demonstrated a significant improvement in the description of the cladding temperature and related variables, such as oxide thickness and hydrogen uptake, when compared with experimental data.

A study on the impact of using a subchannel resolution for modeling of large break loss of coolant accidents

The nuclear industry is investigating the feasibility of transitioning from 18- to 24-month fuel cycles because of the positive impact it would have on the operational costs for the current fleet of light-water reactors. A challenge to making this change is the increased risk of fuel fragmentation, relocation, and dispersal (FFRD) due to the known potential for ceramic fuel to pulverize into fine particles at the higher discharge burnups. Previous work has been performed by the Nuclear Energy Advanced Modeling and Simulation program to assess FFRD risk in high-burnup cores using the BISON fuel performance code and a coarse mesh thermal hydraulics (T/H) solution for a loss-of-coolant accident (LOCA) using the TRACE system T/H code. Because of the importance of the T/H solution for FFRD assessment, this study seeks to investigate the impact of using higher-fidelity subchannel techniques for modeling of the LOCA transient. CTF was used to model a subregion of a high-burnup core that was depleted by the Virtual Environment for Reactor Applications (VERA) multiphysics core simulator. Both coarse-mesh and pin-resolved models were created in CTF, and a consistent coarse-mesh TRACE model was also developed to allow for benchmarking the code results. A large-break loss-of-coolant accident (LBLOCA) reflood transient was simulated using these three models, and results were compared. Results showed some consistent differences between the CTF and TRACE coarse models, including a higher peak cladding temperature (PCT) prediction in CTF and later quenching in CTF; however, the transient clad temperature behavior was similar, and these differences are likely due to post-critical heat flux heat transfer modeling differences and minimum film boiling temperature model differences. The pin-resolved results indicate that the PCT in the lumped model is often under-predicted by as much as 70 °C and that PCT occurs at a different location than the high-power pin in the assembly. The lumped model predicts a difference of 10 °C or less between the average and hot pins in the assembly, whereas the pin-resolved model predicts a range of over 100 °C. These results indicate that higher-fidelity T/H results may have an impact on predicted core behavior during LOCA, which may be important to consider when assessing FFRD risk.

ALCYONE: the fuel performance code of the PLEIADES platform dedicated to PWR fuel rods behavior

The aim of this contribution is to deeply introduce the ALCYONE Fuel Performance Code. ALCYONE is being implemented within the PLEIADES numerical framework and is dedicated to the multiphysics behavior of fuel rods in pressurized water reactors (PWRs). It has been a joint development between CEA, EDF and Framatome since 2005 and has been used since then for numerous studies covering normal, off-normal and accidental conditions. However, a detailed description of ALCYONE’s physical and numerical foundations has never been published. In this sense, this paper is intended to be the reference presentation paper for ALCYONE. On the first hand, it provides a comprehensive description of the modeling features and associated multiphysics and multiscale computational schemes. On the other hand, focuses are made on advanced modeling features showing enhanced capabilities of simulation of ALCYONE.

Advanced interpretation of the SPHERE irradiation experiment with neutronics and fuel performance codes

The SPHERE experiment aimed at studying the behaviour of Minor Actinide-bearing Driver Fuel (U,Pu,Am)O2-x by comparing sphere-packed and pelletized fuels. The irradiation experiment was performed in the High Flux Reactor at Petten from August 2013 to April 2015, and was followed by post-irradiation examinations up to mid-2017. The present work consists in a new analysis of the SPHERE experiment, focusing on the pelletized fuel, by the means of both neutronics and fuel performance codes. This study is performed in the frame of the European Project PATRICIA. The adopted methodology and the main results achieved, assessed in particular against inert gas-related experimental data, are presented in the paper.

Thermodynamic and thermoelastic properties of hypostoichiometric MOX fuels with molecular dynamics simulations

Employing the Cooper-Rushton-Grimes (CRG) interatomic empirical potential with new parameters for U3+ interactions in U1-y,PuyO2-x, classical Molecular Dynamics simulations were carried out for the first time to compute the enthalpy, the lattice parameter, the elastic constants and derived properties of hypostoichiometric U1-y,PuyO2-x compounds (with y in the range 0-1, x between 1.92-2.0, and from 1000 K to melting points). As hypostoichiometry results from adding oxygen vacancies to U1-y,PuyO2, this study investigates how oxygen vacancies affect the thermodynamic and thermomechanical properties of MOX compounds. The most pronounced effect appears at high temperatures, where the so-called Bredig transition occurs. We observed that the effects of this transition on the properties are softened by the presence of oxygen vacancies. The analytical law of the heat capacity for stoichiometric mixed-oxide fuels U1-y,PuyO2 developed by Bathellier et al. [9] was modified by introducing the deviation from stoichiometry as a variable. It was fitted on Molecular Dynamics data of heat capacity and linear thermal expansion coefficient and a good agreement was found. To model the thermal evolution of mechanical properties up to the melting point, a new law was developed. The laws may be used in Fuel Performance Codes.

Uranium-zirconium hydride nuclear fuel performance in the NaK-cooled MARVEL microreactor

The Microreactor Applications Research Validation and EvaLuation (MARVEL) Project is producing a high temperature NaK-cooled nuclear reactor test bed at the Idaho National Laboratory to ultimately improve integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. The computational software package chosen for high fidelity fuel performance modeling, BISON, has been upgraded to model the performance of the MARVEL reactor's selected fuel system: 304 stainless steel-clad U-ZrHx. In this manuscript, we begin by describing the MARVEL fuel element. The known properties and relationships of the fuel element constituents are recapitulated from data available in the literature. These historical relationships assume a particular U-ZrH microstructure which previously was not well-understood; therefore, the fuel's microstructure is experimentally confirmed using electron microscopy and X-ray diffraction. This information is then used to analyze the thermomechanical performance of the MARVEL fuel element during a hypothetical extreme accident scenario (a loss of flow accident during a loss of coolant pressure accident) using the nuclear fuel performance code BISON. The analysis shows that the primary phenomenon which could compromise the structural integrity of the MARVEL fuel element is the same as for TRIGA reactors: high temperature internal gas overpressurization-induced rupture of the cladding arising from excessive hoop stresses. Despite modeling the most severe MARVEL accident scenario, the highest anticipated temperature in the fuel, about 718 °C, results in hoop stresses of less than 10 MPa. Based on this analysis, a conservative steady state fuel meat temperature limit of 900 °C precludes cladding damage. The newly incorporated fuel performance simulation capabilities developed in this work may also be extended to model this fuel system under other conditions or reactors.

Synchrotron micro-computed tomography analysis of neutron-irradiated U-Mo fuel

The three-dimensional (3D) microstructure of neutron-irradiated uranium-10 wt.% molybdenum (U-10Mo) fuel with a burn-up of 9.8 × 1021 fissions/cm3 was characterized using a novel, multi-modal synchrotron micro-computed tomography approach combining propagation-based phase-contrast enhanced and absorption contrast techniques. The porosity development, porosity interconnectedness, swelling, composition, local thickness of the zirconium (Zr) diffusion barrier, and the influence of the fuel–cladding interaction on the local composition and pore morphology, were uniquely determined in 3D. Two cuboids were produced using a focused ion beam-scanning electron microscope at the Zr diffusion barrier–fuel interface and in the bulk fuel. The bulk fuel sample swelled by 53.3 [+9.7/−3.1]%, while the fuel near the Zr–fuel interface swelled by 63.3 [+14.7/−7.2]%. The average local thickness of the Zr diffusion barrier decreased by 53 %, compared to the expected pre-irradiated thickness. Four pore morphology regions were identified initiating parallel to the fuel–Zr interaction region: (1) an interaction layer of suppressed porosity, (2) a layer of elongated and interconnected porosity, (3) a transition zone of low porosity, and (4) a layer of unoriented porosity representative of the bulk fuel behavior. The increase in porosity near the diffusion barrier corresponded to a higher U concentration compared to that in the bulk fuel. The interconnected porosity in the fuel near the diffusion barrier was extensive and oriented parallel to the diffusion barrier, while the bulk fuel had more compact and isolated pore networks. The interaction layer, despite having suppressed porosity, was nearly 100 wt.% U. Porosity suppression at the diffusion barrier corresponds to the expected reduction in radiation-driven diffusion of Xe at the interface despite the anticipated increase in fission product nucleation originating from a higher U concentration. The novel 3D insights of the porosity, swelling, and compositional variations characterized herein can improve the fidelity of fuel performance codes for proliferation-resistant fuels for research and test reactors.

Predicted thermophysical properties of UN, PuN, and (U,Pu)N

Molecular dynamics and density functional theory simulations are used to predict the lattice and electronic contributions of thermophysical properties for UN, PuN, and mixed (U,Pu)N systems. The properties predicted include the lattice parameter, linear thermal expansion, enthalpy, and specific heat capacity, as a function of temperature. The simulation predictions for high temperature specific heat capacity are compared against experimental measurements to understand the behavior, and why differences in the experimental measurements are observed. The influence of adding U vacancies, N interstitials, and Pu to UN is also examined. For this, a new PuN potential parameter set is developed and used with the Kocevski UN potential, enabling the dynamics of mixed (U,Pu)N systems to be studied. How defects impact the thermophysical properties is important for understanding fuel behavior under different reactor conditions, and these mechanistic predictions can be used to support fuel performance codes where data is scarce.

AI-driven non-intrusive uncertainty quantification of advanced nuclear fuels for digital twin-enabling technology

In response to the urgent need to establish AI/ML-integrated Digital Twin (DT) technology within next-generation nuclear systems, advancements in modeling methods and simulation codes are necessary. The increased complexity of models demands significant computational resources to quantify their uncertainties. To address this challenge, a data-driven non-intrusive uncertainty quantification method via polynomial chaos expansion is introduced as an efficient strategy within the finite element analysis-based fuel performance code BISON. Models of UO2 and U3Si2 fuels, alongside SiC/SiC cladding material, were prepared to demonstrate the proposed method. The impact of four independent uncertain input variables on the system output was quantified, requiring fewer than 100 BISON simulations for each model. This approach not only accelerates the modeling and simulation task but also enhances the reliability in the development of DT-enabling technologies.