Fission gas

The swelling mechanisms of U3Si2 under neutron irradiation in reactor conditions are not unequivocally known. The limited experimental evidence that is available suggests that the main driver of the swelling in this material would be fission gases accumulation at crystalline grain boundaries. The steps that lead to the accumulation of fission gases at these locations are multiple and complex. However, gradually, the gaseous fission products migrate by diffusion. Upon reaching a grain boundary, which acts as a trap, the gaseous fission products start to accumulate, thus leading to formation of bubbles and hence to swelling. Therefore, a quantitative model of swelling requires the incorporation of phenomena that increase the presence of grain boundaries and decrease grain sizes, thus creating sites for bubble formation and growth. It is assumed that grain boundary formation results from the conversion of stored energy from accumulated dislocations into energy for the formation of new ...

The fuel modelling code FUROM (FUel ROd Model), suitable for calculating the normal operation condition behaviour of PWR and WWER fuels, has been developed at AEKI for several years. The validation of the code has so far been based on the individual calculation of many relevant experiments. This, however, was a time consuming process that could give rise to errors both at the input and at the comparison stage. A new methodology is implemented to build up a uniform database from the IFPE data and run automated validation tasks depending on the model or phenomenon of interest. The general problems encountered and some results are presented here.

In this work, we present a model of fission gas behavior in U 3 Si 2 under light water reactor (LWR) conditions for application in engineering fuel performance codes. The model includes components for intra-granular and inter-granular behavior of fission gases. The intra-granular component is based on cluster dynamics and computes the evolution of intra-granular fission gas bubbles and swelling coupled to gas diffusion to grain boundaries. The inter-granular component describes the evolution of grain-boundary fission gas bubbles coupled to fission gas release. Given the lack of experimental data for U 3 Si 2 under LWR conditions, the model is informed with parameters calculated via atomistic simulations. In particular, we derive fission gas diffusivities through density functional theory calculations, and the resolution rate of fission gas atoms from intra-granular bubbles through binary collision approximation calculations. The developed model is applied to the simulation of an experiment for U 3 Si 2 irradiated under LWR conditions available from the literature. Results point out a credible representation of fission gas swelling and release in U 3 Si 2. Finally, we perform a sensitivity analysis for the various model parameters. Based on the sensitivity analysis, indications are derived that can help in addressing future research on the characterization of the physical parameters relative to fission gas behavior in U 3 Si 2. The developed model is intended to provide a suitable infrastructure for the engineering scale calculation of fission gas behavior in U 3 Si 2 that exploits a multiscale approach to fill the experimental data gap and can be progressively improved as new lower-length scale calculations and validation data become available.

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons are not satisfactory and have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that 1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, 2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and 3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons are not satisfactory and have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

Les gaz de fission produits pendant l'irradiation d'un combustible nucléaire peuvent provoquer du gonflement et la libération de gaz de fission, qui peut affecter la performance du combustible. La théorie de la diffusion effective, qui est généralement utilisée dans la modélisation de l’irradiation de base du combustible nucléaire, ne permet pas de prédire le dégagement de gaz de fission intra-granulaire pendant les essais de recuit post-irradiation. Nous discutons et analysons trois des mécanismes possibles: le mouvement dirigé vers la surface du grain de bulles de gaz intra-granulaires pressurisées dans un gradient de lacunes, le mouvement brownien de ces bulles via des mécanismes de diffusion en volume et en surface, un scénario de mouvement de bulles de gaz avec les dislocations via le mécanisme de montée de dislocation. Un nouveau modèle spatialisé, BEEP, a été mis au point pour la migration des bulles de gaz et leur interaction avec les défauts ponctuels. Les analyses ...

One objective of the FPT2 test of the PHEBUS FP Program was to study the degradation of an irradiated UO 2 fuel bundle and the fission product behaviour under conditions of low steam flow. The results of the post-irradiation examinations (PIE) at the upper levels (823 mm and 900 mm) of the test section previously reported are interpreted in the present paper. Solid state interactions between fuel and cladding have been compared with the characteristics of interaction identified in the previous separate-effect tests. Corium resulting from the interaction between fuel and cladding was formed. The uranium concentration in the corium is compared to analytical tests and a scenario for the corium formation is proposed. The analysis showed that, despite the rather low fuel burn up, the conditions of temperature and oxygen potential reached during the starvation phase are able to give an early very significant release fraction of caesium. A significant part (but not all) of the molybdenum was segregated at grain boundaries and trapped in metallic inclusions from which they were totally removed in the final part of the experiment. During the steam starvation phase, the conditions of oxygen potential were favourable for the formation of simple Ba and BaO chemical forms but the temperature was too low to provoke their volatility. This is one important difference with out-of-pile experiments such as VERCORS for which only a combination of high temperature and low oxygen potential induced a significant barium release. Finally another significant difference with analytical out-of-pile experiments comes from the formation of foamy zones due to the fission gas presence in FPT2-type experiments which give an additional possibility for the formation of stable fission product compounds.

ABSTRACTThe Knudsen Effusion Mass Spectrometer (KEMS) and the mechanistic MFPR (Module for Fission Product Release) code are tools which seem particularly interesting to support studies of the Instant Release Fraction (IRF) of Cs from spent nuclear fuel in a final repository. With KEMS, the thermal release of 137Cs and 136Xe were analysed by annealing up to total vaporization (2500K) of high burn-up (60 GWd/tU) Spent Nuclear Fuel (SNF) samples. Powder samples from the centre of the fuel, without high burn-up structure, were used. To determine the IRF, samples were analysed before and after being submitted to corrosion experiments in bicarbonated aqueous media.MFPR was applied to determine the localization of Cs and fission gases in the SNF at the end of irradiation; the results are compared and supported by dedicated thermodynamics calculations performed for equilibrium conditions at various temperatures and fuel oxygen potentials by the non-ideal thermodynamic MEPHISTA (Multiphase ...

Analysis of advanced gas-cooled reactor (AGR) simulated used nuclear fuels (SIMFuels) has been carried out using micro-Raman spectroscopy in order to understand the effect lanthanide species (e.g. Nd, Y, Ce), representative of fission products generated during fuel burnup, have on the structure of the UO 2 matrix in spent AGR fuel. Results show a decrease in perfect fluorite character with increasing burnup as well as the development of a broad lattice distortion peak between 500 and 650 cm-1. Peak analysis of this broad band reveals in it comprised of three overlapping peaks at 534 cm-1 , 574 cm-1 and 624 cm-1. The peak at 534 cm-1 has been examined and is suggested to be due to a local phonon mode associated with oxygen-vacancy-induced lattice distortion as a result of lanthanide 3+ ion incorporation into the UO 2 bulk matrix.

Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swe...

This article presents a synthesis of the investigations at the atomic scale of the transport properties of defects and fission gases in uranium dioxide, as well as of the transfer of results from the atomic scale to models at the mesoscopic scale, performed during the F-BRIDGE European project (2008-2012). We first present the mesoscale models used to investigate uranium oxide fuel under irradiation, and in particular the cluster dynamics and kinetic Monte Carlo methods employed to model the behaviour of defects and fission gases in UO 2 , as well as the parameters of these models. Second, we describe briefly the atomic scale methods employed, i.e. electronic structure calculations and empirical potential methods. Then, we show the results of the calculation of the data necessary for the mesoscale models using these atomic scale methods. Finally, we summarise the links built between the atomic and mesoscopic scale by listing the data calculated at the atomic scale which are to be used as input in mesoscale modelling. Despite specific difficulties in the description of fuel materials, the results obtained in F-BRIDGE show that atomic scale modelling methods are now mature enough to obtain precise data to feed higher scale models and help interpret experiments on nuclear fuels. These methods bring valuable insight, in particular the formation, binding and migration energies of point and extended defects, fission product localization, incorporation energies and migration pathways, elementary mechanisms of irradiation induced processes. These studies open the way for the investigation of other significant phenomena involved in fuel behaviour, in particular the thermochemical and thermomechanical properties and their evolution in-pile, complex microstructures, as well as of more complex fuels.

The swelling mechanisms of U3Si2 under neutron irradiation in reactor conditions are not unequivocally known. The limited experimental evidence that is available suggests that the main driver of the swelling in this material would be fission gases accumulation at crystalline grain boundaries. The steps that lead to the accumulation of fission gases at these locations are multiple and complex. However, gradually, the gaseous fission products migrate by diffusion. Upon reaching a grain boundary, which acts as a trap, the gaseous fission products start to accumulate, thus leading to formation of bubbles and hence to swelling. Therefore, a quantitative model of swelling requires the incorporation of phenomena that increase the presence of grain boundaries and decrease grain sizes, thus creating sites for bubble formation and growth. It is assumed that grain boundary formation results from the conversion of stored energy from accumulated dislocations into energy for the formation of new ...

Physics-based meso-scale models of fission gas behaviour for fuel performance codes currently consider only the average grain size as physical parameter to describe the fuel microstructure. Nevertheless, information on the grain-size distribution is available for several metallographically characterized fuel samples. To extend the current modelling approach, we present new experimental data and develop a methodology to treat the fuel grain-size distribution based on multi-grain analysis. This is applied to describe helium behaviour from infused and annealed polycrystalline UO 2 samples. The methodology consists of three steps: (1) Acquisition of the empirical grain distribution from sections of the polycrystals, (2) Calculation of the gas distribution factor pertaining to each grain size class after helium infusion, and (3) Simulation of the experimental annealing histories with input parameters weighted by the gas distribution factors. To perform the multi-grain analysis, we used the SCIANTIX code, which allows calculating the helium kinetics in a single fuel grain. The application of this methodology is promising because it can represent the dynamics of helium release more consistently with the physics of the fuel microstructure compared to the state-of-art approaches, and it can be suitable for application in fuel performance codes for different enhanced accident tolerant fuel concepts as outlined in our conclusions.

In this study, smaller-grained (hundred nano-meter size grain) and larger-grained (micro-meter size grain) U-10Mo specimens have been irradiated (implanted) with 250 keV Xe + beam and were in situ characterized by TEM. Xe bubbles were not seen in the specimen after an implantation fluence of 2×10 20 ions/m 2 at room temperature. Nucleation of Xe bubbles happened during heating of the specimen to a final temperature of 300 ˚C. By comparing measured Xe bubble statistics, the nucleation and growth behaviors of Xe bubbles were investigated in smaller-grained and larger-grained U-10Mo specimens. A multi-atom kind of nucleation mechanism has been observed in both specimens. X-ray Absorption spectroscopy showed the edge position in the bubbles to be the same as that of Xe gas. The size of Xe bubbles has been shown to be bigger in larger-grained specimens than in smaller-grained specimens at the same implantation conditions.

This article presents a synthesis of the investigations at the atomic scale of the transport properties of defects and fission gases in uranium dioxide, as well as of the transfer of results from the atomic scale to models at the mesoscopic scale, performed during the F-BRIDGE European project (2008-2012). We first present the mesoscale models used to investigate uranium oxide fuel under irradiation, and in particular the cluster dynamics and kinetic Monte Carlo methods employed to model the behaviour of defects and fission gases in UO 2 , as well as the parameters of these models. Second, we describe briefly the atomic scale methods employed, i.e. electronic structure calculations and empirical potential methods. Then, we show the results of the calculation of the data necessary for the mesoscale models using these atomic scale methods. Finally, we summarise the links built between the atomic and mesoscopic scale by listing the data calculated at the atomic scale which are to be used as input in mesoscale modelling. Despite specific difficulties in the description of fuel materials, the results obtained in F-BRIDGE show that atomic scale modelling methods are now mature enough to obtain precise data to feed higher scale models and help interpret experiments on nuclear fuels. These methods bring valuable insight, in particular the formation, binding and migration energies of point and extended defects, fission product localization, incorporation energies and migration pathways, elementary mechanisms of irradiation induced processes. These studies open the way for the investigation of other significant phenomena involved in fuel behaviour, in particular the thermochemical and thermomechanical properties and their evolution in-pile, complex microstructures, as well as of more complex fuels.

A model to predict fission gas behavior in irradiated uranium dioxide fuel during the steadystate operation of a nuclear reactor is developed. The basic physical phenomena encountered in analyzing the disposition of fission gas have been retained, but in a simplified form for ease of calculation. The analysis includes treatment of intragranular, grain face, and grain edge gas, and release to open spaces. The code is utilized to obtain comparison with experimental data and to perform fuel behavior studies. The sensitivity studies indicate the importance of grain face and grain edge bubble treatments in modeling fission gas. It is found that representation of release in different sections of the fuel pin is possible in a simple way by assuming evenly spaced bubbles on the edge, and that grain edge bubble interlinkage is a necessary condition for release to the open spaces. The sensitivity studies show that fission gas swelling is mainly due to grain edge bubbles. Grain face bubbles, although large in size, are few in number and contribute little to swelling. Intragranular swelling is intermediate between these two values. The code is successfully used to analyze the Westinghouse fission gas release data from the Zorita, Spain, light water reactor and data from the U.K. reactor DIDO. This success in modeling experiments suggests that the present code can be used in predicting fuel element performance, which is necessary in nuclear fuel design, safety analysis, and interpretation of experimental data on fuel element behavior.

Among the applications of the multiscale modelling approach in nuclear fuel rod performance, the coupling of integral thermo-mechanical fuel performance codes with lower-length meso-scale modules is of great interest. This strategy allows to overcome correlation-based approaches with mechanistic ones and test their application in accidental conditions. In this work, we explore the coupling between the TRANSURANUS fuel performance code and two meso-scale modules for fission gas/product behaviour: MFPR-F and SCIANTIX. These modules, coupled within TRANSURANUS, are assessed against the IFA-650.10 loss-of-coolant accident test to analyse their overall impact and highlight future developments toward mechanistic modelling of fission gas during accident scenarios.

Gaseous fission product transport and release has a large impact on fuel performance, degrading fuel and gap properties. While gaseous fission product behavior has been investigated with bulk reactor experiments and simplified analytical models, recent improvements in experimental and modeling approaches at the atomistic and mesoscales are beginning to reveal new understanding of the unit mechanisms that define fission product behavior. Here, existing research on the basic mechanisms of fission gas release during normal reactor operation are summarized and critical areas where work is needed are identified. This basic understanding of the fission gas behavior mechanisms has the potential to revolutionize our ability to predict fission product behavior and to design fuels with improved performance. In addition, this work can serve as a model on how a coupled experimental and modeling approach can be applied to understand the unit mechanisms behind other critical behaviors in reactor materials.

The TRANSURANUS nuclear fuel performance code of the European Commission, Joint Research Centre has been extended with new best-estimate and conservative models to support deterministic and probabilistic safety analyses of Westinghouse's PWR and VVER fuel rod designs. The recent model refinements and code validation analyses have been carried out in the EU-funded project ESSANUF aiming at the development of an alternative nuclear fuel, as well as licensing methods and methodologies for the Russian-designed pressurized water reactor VVER-440. The Westinghouse-specific developments of TRANSURANUS have covered fission gas release, fuel swelling, cladding corrosion and hydrogen uptake, rod growth, cladding creep, rod burst, inner cladding steam oxidation and oxide breakaway. The code validation has been based on two major data sources: (1) an extensive database of Westinghouse consisting of pool-side measurements and postirradiation examinations (PIE) for a large number of commercia...

Modeling of the thermo-mechanical behavior of the nuclear fuel rod allows to assess the temperatures and stresses that can be used to determine if the fuel rod will operate without failures during proposed steady state conditions and permitted power ramps and also during unexpected power excursions. The paper presents long-term effort to establish closed chain of computer codes, which allow computer modeling of the thermo-mechanical fuel rod behavior from the one dimensional (1D) representation to the 3D detailed problem description. The whole complex uses the same finite element method (FEM) based mathematical approach. The principle of this modular approach is to handle boundary and initial conditions by the similar transfer independent on the complexity of solved problem. We have combined the integral code to the full 2D (r-z) global model of fuel pin ending at 2D (r-z, r-φ) and 3D local detailed submodels. The system of codes is connected (coupled) by data transfer. All the subm...

The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code with a recently implemented physics-based model for fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO 2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, significantly higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

BISON is a nuclear fuel performance application built using the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element library. One of its major goals is to have a great amount of flexibility in how it is used, including in the types of fuel it can analyze, the geometry of the fuel being modeled, the modeling approach employed, and the dimensionality and size of the models. Fuel forms that can be modeled include standard light water reactor fuel, emerging light water reactor fuels, tri-structural isotropic fuel particles, and metallic fuels. BISON is a platform for research in nuclear fuel performance modeling while simultaneously serving as a tool for the analysis of nuclear fuel designs. Recent research in BISON includes techniques such as the extended finite element method for fuel cracking, exploration of high-burnup light water reactor fuel behavior, swelling behavior of metallic fuels, and central void formation in mixed-oxide fuel. BISON includes integrated documentation for each of its capabilities, follows rigorous software quality assurance procedures, and has a growing set of rigorous verification and validation tests.

U 3 Si 2 fuel is a potential candidate replacement for UO 2 in light water reactor fuel rods due to its higher uranium density and thermal conductivity. Its lower melting temperature and significant gaseous swelling may be of concern. The aggressive development schedule of the United States Department of Energy's (DOE) Advanced Fuels Campaign (AFC) to have potential accident tolerant fuels (ATF) concepts in lead test rods or assemblies as quickly as possible has necessitated the need for advanced modeling and simulation techniques to investigate the behavior of these materials under normal operating and accident conditions. This report presents the modeling capabilities for U 3 Si 2 fuel at the engineering (continuum) scale developed as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) high impact problem (HIP). In this report, the models incorporated into the BISON fuel for U 3 Si 2 are described and comparisons to conventional UO 2 for separate effects cases are presented.

The behaviour of the fission gas plays an important role in the fuel rod performance. In a previous work, we presented a physics-based model describing intra-and inter-granular behaviour of radioactive fission gas. The model was implemented in SCIANTIX, a mesoscale module for fission gas behaviour, and assessed against the CONTACT 1 irradiation experiment. In this work, we present the multi-scale coupling between the TRANSURANUS fuel performance code and SCIANTIX, used as mechanistic module for stable and radioactive fission gas behaviour. We exploit the coupled code version to reproduce two integral irradiation experiments involving standard fuel rod segments in steady-state operation (CONTACT 1) and during successive power transients (HATAC C2). The simulation results demonstrate the predictive capabilities of the code coupling and contribute to the integral validation of the models implemented in SCIANTIX.

The International Atomic Energy Agency (IAEA) sponsored the Coordinated Research Project (CRP) on Fuel Modeling under Accident Conditions (FUMAC) to coordinate and support research on nuclear fuel modelling under accident conditions in member countries. The focus of the FUMAC CRP (20152018) has been on loss-of-coolant accidents (LOCA). Various institutions performed fuel performance simulations of selected experiments using different fuel performance codes (e.g., FRAPCONFRAPTRAN, TRANSURANUS, ALCYONE, DIONISIO, SOCRAT, FTPAC, BISON, RAPTA) and system codes (e.g SOCRATE, ATHLET). One of the results of the FUMAC CRP is a comprehensive code-to-code benchmark of selected results, and a comparison of simulations with experimental data as well. This paper represents an overview of the current state-of-the-art of nuclear fuel simulation capabilities for LOCAs and paves the way to further analyses and future developments. More precisely, we discuss the results of the simulation of a subset ...

In-reactor tests and post-irradiation examinations (PIEs) were performed for standard and large-grained pellets with and without additives being soluble in a matrix and/or precipitated in a grain boundary, to confirm the effects of large grain structure on decreasing fission gas release (FGR) and swelling and to evaluate the influence of the additives in the matrix/grain boundary on them. The standard and large-grained pellets were loaded into smail-diameter rods equipped with a pressure gauge. These rods were irradiated to about 60 GWd/t U at a linear heat rate of about 30-40 kW/m in the Halden reactor and then subjected to PIEs. Large-grained pellets showed a smaller FGR compared with standard pellets. Post-irradiation annealing tests suggested that swelling during transient power was decreased for large-grained pellets, except for those with additive enhancing cation diffusion.

Advanced microstructure investigations of the high-burnup structure (HBS) in UO 2 produced by high-dose 84 MeV Xe ion irradiation are reported. Spark plasma sintered micro-grained UO 2 was irradiated to 1357 dpa at 350 • C. The characteristic nano-grains and micro-pores of the HBS were formed. The grain size and grain boundary misorientation distributions of the HBS were measured using transmission electron microscopy based orientation imaging microscopy. Grain polygonization due to accumulation of radiation-induced dislocations was found to be the mechanism of nano-crystallization. The morphology of Xe bubbles was quantitatively investigated. This study provides crucial references for advanced fuel performance modeling of high-burnup UO 2 .

The neutronics performance and safety characteristics of Uranium mononitride (UN) fuel for System-Integrated Modular Advanced Reactor (SMART) has been investigated to discern the potential for non-proliferation, waste, and accident tolerance benefits of UN fuel. The neutronic evaluation of UN fuel for SMART reactor has been carried out under normal operation using OpenMC and compared with Uranium dioxide (UO 2) in terms of fuel cycle length, reactivity coefficients, Fuel depletion (burnup), thermal flux, and fission product activity. The power peaking factor (PPF) has been compared at the beginning of the fuel cycle (BOC), mid of the fuel cycle (MOC), and at the end of the fuel cycle (EOC). Results indicate that the UN fuel can be operated beyond the designed fuel cycle length of the SMART reactor, which induces the positive reactivity at the end of the fuel cycle of about 4625 pcm. However, the UO 2 showed negative reactivity after three years. The total fission product activity at the end of the fuel cycle (3.5 years) for UO 2 and UN has been founded 1.003 × 10 20 Bq and 1.023 × 10 20 Bq, respectively.

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANS-URANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

In this work, we propose a new mechanistic model for the treatment of helium behaviour which includes the description of helium solubility in oxide fuel. The proposed model has been implemented in SCIANTIX and validated against annealing helium release experiments performed on small doped fuel samples. The overall agreement of the new model with the experimental data is satisfactory, and given the mechanistic formulation of the proposed model, it can be continuously and easily improved by directly including additional phenomena as related experimental data become available.

The modelling of fission gas behaviour is a crucial aspect of nuclear fuel performance analysis in view of the related effects on the thermo-mechanical performance of the fuel rod, which can be particularly significant during transients. In particular, experimental observations indicate that substantial fission gas release (FGR) can occur on a small time scale during transients (burst release). To accurately reproduce the rapid kinetics of the burst release process in fuel performance calculations, a model that accounts for non-diffusional mechanisms such as fuel micro-cracking is needed. In this work, we present and assess a model for transient fission gas behaviour in oxide fuel, which is applied as an extension of conventional diffusion-based models to introduce the burst release effect. The concept and governing equations of the model are presented, and the effect of the newly introduced parameters is evaluated through an analytic sensitivity analysis. The model is assessed for application to integral fuel rod analysis by implementation in two structurally different fuel performance codes: BISON (multi-dimensional finite element code) and TRANSURANUS (1.5D code). Model assessment is based on the analysis of 19 light water reactor fuel rod irradiation experiments from the OECD/NEA IFPE (International Fuel Performance Experiments) database, all of which are simulated with both codes. The results point out an improvement in both the quantitative predictions of integral fuel rod FGR and the qualitative representation of the FGR kinetics with the transient model relative to the canonical, purely diffusion-based models of the codes. The overall quantitative improvement of the integral FGR predictions in the two codes is comparable. Moreover, calculated radial profiles of xenon concentration after irradiation are investigated and compared to experimental data, illustrating the underlying representation of the physical mechanisms of burst release.

The microstructural change of UO, fuel pellets (burnup: 6-83 GWd/t), base irradiated under LWR conditions, has been studied by detailed postirradiation examinations. The lattice parameter near the fuel rim in the irradiated UO, increased with burnup and appeared to become constant beyond about 50 GWd/t. This lattice dilation was mainly due to the accumulation of radiation induced point defects. Moreover, the dislocation density in the UO, matrix developed progressively with burnup, and eventually the tangled dislocations organized many sub-grain boundaries in the highest bumup fuel of 83 GWd/t. This sub-grain structure induced by accumulated radiation damage was compatible in appearance with SEM fractography results which revealed subdivided grains of sub-micron size in as-fabricated grains. The influence of burnup on "Kr release from the UOz fuels has been examined by means of a postirradiation annealing technique. The higher fractional release of high burnup fuels was mainly due to the burnup dependence of the fractional burst release evolved on temperature ramp. The fractional burst release was represented in terms of the square root of burnup from 6 to 83 GWd/t.

A model to predict fission gas behavior in irradiated uranium dioxide fuel during the steadystate operation of a nuclear reactor is developed. The basic physical phenomena encountered in analyzing the disposition of fission gas have been retained, but in a simplified form for ease of calculation. The analysis includes treatment of intragranular, grain face, and grain edge gas, and release to open spaces. The code is utilized to obtain comparison with experimental data and to perform fuel behavior studies. The sensitivity studies indicate the importance of grain face and grain edge bubble treatments in modeling fission gas. It is found that representation of release in different sections of the fuel pin is possible in a simple way by assuming evenly spaced bubbles on the edge, and that grain edge bubble interlinkage is a necessary condition for release to the open spaces. The sensitivity studies show that fission gas swelling is mainly due to grain edge bubbles. Grain face bubbles, although large in size, are few in number and contribute little to swelling. Intragranular swelling is intermediate between these two values. The code is successfully used to analyze the Westinghouse fission gas release data from the Zorita, Spain, light water reactor and data from the U.K. reactor DIDO. This success in modeling experiments suggests that the present code can be used in predicting fuel element performance, which is necessary in nuclear fuel design, safety analysis, and interpretation of experimental data on fuel element behavior.

When assessing the radiological consequences of postulated accident scenarios, it is of primary interest to determine the amount of radioactive fission gas accumulated in the fuel rod free volume. The state-ofthe-art semi-empirical approach (ANS 5.4e2010) is reviewed and compared with a mechanistic approach to evaluate the release of radioactive fission gases. At the intra-granular level, the diffusiondecay equation is handled by a spectral diffusion algorithm. At the inter-granular level, a mechanistic description of the grain boundary is considered: bubble growth and coalescence are treated as interrelated phenomena, resulting in the grain-boundary venting as the onset for the release from the fuel pellets. The outcome is a kinetic description of the release of radioactive fission gases, of interest when assessing normal and off-normal conditions. We implement the model in SCIANTIX and reproduce the release of short-lived fission gases, during the CONTACT 1 experiments. The results show a satisfactory agreement with the measurement and with the state-of-the-art methodology, demonstrating the model soundness. A second work will follow, providing integral fuel rod analysis by coupling the code SCIANTIX with the thermo-mechanical code TRANSURANUS.

The description of intra-granular fission gas behavior during irradiation is a fundamental part of models used for the calculation of fission gas release and gaseous swelling in nuclear fuel performance codes. The relevant phenomena include diffusion of gas atoms towards the grain boundaries coupled to the evolution of intra-granular bubbles. While intra-granular bubbles during normal operating conditions are limited to sizes of a few nanometers, experimental evidence exists for the appearance of a second population of bubbles during transients, characterized by coarsening to sizes of tens to hundreds of nanometers and that can significantly contribute to gaseous fuel swelling. In this work, we present a model of intra-granular fission gas behavior in uranium dioxide fuel that includes both nanometric fission gas bubble evolution and bubble coarsening during transients. While retaining a physical basis, the developed model is relatively simple and is intended for application in engineering fuel performance codes. We assess the model through comparisons to a substantial number of experimental data from SEM observations of intra-granular bubbles in power ramp tested uranium dioxide samples. The results demonstrate that the model reproduces the coarsening of a fraction of the intra-granular bubbles and correspondingly, predicts gaseous swelling during power ramps with a significantly higher accuracy than is allowed by traditional models limited to the evolution of nanometric intra-granular bubbles.

The role of uncertainties in fission gas behavior calculations as part of engineering-scale nuclear fuel modeling is investigated using the BISON fuel performance code with a recently implemented physics-based model for fission gas release and swelling. Through the integration of BISON with the DAKOTA software, a sensitivity analysis of the results to selected model parameters is carried out based on UO 2 single-pellet simulations covering different power regimes. The parameters are varied within ranges representative of the relative uncertainties and consistent with the information in the open literature. The study leads to an initial quantitative assessment of the uncertainty in fission gas behavior predictions with the parameter characterization presently available. Also, the relative importance of the single parameters is evaluated. Moreover, a sensitivity analysis is carried out based on simulations of a fuel rod irradiation experiment, pointing out a significant impact of the considered uncertainties on the calculated fission gas release and cladding diametral strain. The results of the study indicate that the commonly accepted deviation between calculated and measured fission gas release by a factor of 2 approximately corresponds to the inherent modeling uncertainty at high fission gas release. Nevertheless, significantly higher deviations may be expected for values around 10% and lower. Implications are discussed in terms of directions of research for the improved modeling of fission gas behavior for engineering purposes.

Fission gas behavior of U0 2 fuel pellets with controlled microstructure irradiated to 23 GW d/t in a test reactor has been studied by using a postirradiation annealing experiment. Four types of fuel pellets with or without additives were examined: (1) un-doped standard (grain size: 1611m), (2) un-doped large grained (43 pm), (3) 0.7 wt?r;' Nb 20 5-doped large grained (110 pm), (4) 0.2 wt?.;' Ti0 2-doped large grained (85 pm) fuels. The annealing was conducted at 1,600 or 1,800oC for 5 h in reducing or oxidizing atmospheres. Fission gas release and bubble swelling caused by the high temperature annealing for the two un-doped fuels were reduced to about 1/3-1/2 by increasing the grain size from 16 to 4311m, which roughly corresponded to the ratio of their grain sizes. By contrast, the performance of the two large grained fuels doped with Nb20 5 or Ti02 was roughly equivalent to, or rather inferior to that of the standard fuel, despite their large grain sizes of 110 and 85 pm. The fission gas behavior of un-doped fuels was aggravated by increasing the oxygen potential in the annealing atmosphere, while that of additive doped fuels did not depend on it. The effects of grain size. additive doping and oxygen potential on fission gas release and bubble swelling were discussed in connection with the diffusivities of fission gas atoms and cation vacancies.

Investigations of fuel behaviour are carried out in close connection with experimental research operation feedback and computational analyses. In this connection, OECD NEA sets up the "public domain database on nuclear fuel performance experiments for the purpose of code development and validation-International Fuel Performance Experiments (IFPE) database", with the aim of providing a comprehensive and well-qualified database on UO 2 fuel with Zircaloy cladding for model development and code validation. The aim of the activity is to assess the capability of TRANSURANUS (version "v1m1j08") code in predicting the phenomenon of the pellet clad interaction (PCI). The objective is pursued using two experimental databases carried out in the research reactor R2 at Studsvik and belonging to the IFPE database. The experiments address the behavior of standard-type BWR and PWR fuel rods, including preceding base irradiation, during the overpower ramping. Pre-, during-, and post-irradiation, non destructive and destructive examinations were executed, in order to determine and understand the behavior of the fuel rods, but also to provide suitable data, useful for code validation.

for single crystals, with the Boltzmann factor 1/kT in (eV 1). The correlation for Henry's constant in powder samples is of interest for the analysis of helium behaviour in the fuel after the pulverization occurring during LOCA-like temperature transients, while the correlation for Henry's constant in single-crystals is usable in meso-scale models describing helium behaviour at the level of fuel grains. The current lack of data for this fundamental property, especially for poly-crystalline samples, calls for new experiments.

The oxygen potentials and lattice parameter of oxide fuels (U02 and U0,-2 %Gd,O,) irradiated in commercial BWRs have been measured by using a solid electrolyte galvanic cell technique and X-ray diffractometry. The fuel burn-up was 13-30 GWd/t for the former measurements and 6-35 GWd/t for the latter. The oxygen potentials at 750"C for irradiated U02 ranged from-420 kJ/mol at the fuel rim to-480 kJ/mol at the fuel center. The corresponding values for irradiated U0 2-2 %Gd 2 0 3 ranged from-460 to-540 k]/mol. The lattice parameter near the fuel rim in the irradiated U0 2 increased with burn-up, and was distinctly larger than that at the fuel center. This lattice dilatation was mainly due to the accumulation of radiation defects, not to a decrease in 0/M ratio, and was almost recovered after the annealing at 1,600"C for 5 h.

High-resolution TEM HRTEM observations and nano-area EDX analyses were carried out on small intragranular bubbles in the outer region of high burnup UO pellets. Sample was prepered from the outer region of UO pellet, which 2 2 had been irradiated to the pellet average burnups of 49 GWdrt in a BWR. HRTEM observations and element analyses were Ž . Ž . made with a 200 KV cold-type field emission TEM Hitachi FE-2000 having an ultra-thin window EDX Noran Voyager . Lattice image and nano-area EDX results indicate the presence of 4-8 nm size Xe-Kr bubbles along with fission products of five metal particles, Mo-Tc-Ru-Rh-Pd. Nano-diffraction patterns from bubbles show two different new patterns besides matrix UO . From the XerU proportion obtained by nano-area EDX peak and nano-diffraction patterns from bubbles, it was 2 concluded that Xe in the small bubbles was present in a solid or near solid state at very high pressure. Furthermore, from the results of high resolution images and diffractions obtained from recrystallized grains in rim structure region, neighboring recrystallized grains were clarified to be present with high angle grain boundaries. q 1997 Elsevier Science B.V.

Since the 90's, EDF and AREVA-NP have irradiated, up to very high burnups, lead assemblies housing M5 ® cladded fuels. Post-irradiation examination of high burnup UO 2 pellets show an increase in the fission-gas release rate, an increase in fuel swelling, and formation of fission-gas bubbles throughout the pellets. Xenon abundances were quantified, and phenomena leading to this bubble formation were identified. All examinations provided valuable data on the complex state of the fuel during irradiation. They show the good behavior of these fuels, exhibiting various microstructures at very high burnups, none of which is likely to lead to problems during irradiation.