[en] In 1986 the irradiation of the first prototypes of MOX fuels fabricated in Argentina started. The experiment's description, the results of the PlEs and the comparison with the output of the BACO code were published in 1996. In particular, Eddy current testings were performed before and after irradiation. The latter yielded wavelike signals whose amplitude variations can be easily correlated with the pellet distribution through the fuel rod and with the power profile. The present work attempts to give a thermomechanical interpretation of this experimental fact. The pellet and the cladding are simulated by a finite element scheme. Although the results are still preliminary, the tendency of the system to expand preferentially in the vicinity of the pellet's edge is well represented and the results correlate properly with the experimental observations. (author)

[en] When the residence time of nuclear fuel rods exceeds a given threshold value, several properties of the pellet material suffer changes and hence the posterior behaviour of the rod is significantly altered. Structural modifications start at the pellet periphery, which is usually referred to as rim zone. It is presently believed that these changes are a consequence of the localized absorption of epithermal neutrons by ²³⁸U, which effective cross section presents resonant peaks. Due to the chain of nuclear reactions that take place, several Pu isotopes are born especially at the rim. In particular, the fissile character of ²³⁹Pu and ²⁴¹Pu is the cause of the increased number of fission events that occur in the pellet periphery. For this reason, the power generation rate and the burnup adopt a non uniform distribution in the pellet, reaching at the rim values two or three times higher than the average [1]. The rim zone starts to form for a burnup threshold value of about 50-60 MWd/kgHM and its width increases as the irradiation progresses. The microstructure of this zone is characterized by the presence of small grains, with a typical size of 200 nm, and large pores, of some μm. Even though the rim zone is very thin, it has a significant effect on the mechanical integrity of the pellet, particularly when it makes contact with the cladding, and on the temperature distribution in the whole pellet, because of its low thermal conductivity [1,2]. The numerical codes designed to simulate fuel behaviour under irradiation must include the phenomena associated to high burnup if they aim at extending the prediction range, and this is the purpose with our DIONISIO code. But a detailed analysis of the phenomena that take place in this region demands the use of neutronic codes that solve the Boltzmann transport equations [3] in a number of energy intervals (groups), including adequate considerations in the region of the resonant absorption peaks of ²³⁸U. These cell codes predict with high precision the neutron flux, burnup and concentration of every isotope, fissile, fissionable or fertile, gaseous or solid, all of them as functions of radius and time. But this formidable task is not suitable to be included in a fuel performance code, which must attend the great number of thermomechanical and thermochemical processes within the fuel rod. To accommodate both requirements, a simplified treatment is adopted consisting of restricting the balance equations to more relevant nuclides and reducing the energy spectrum to a single group. The purpose is to obtain empirical expressions to represent, with the higher possible approximation degree, the absorption, capture and fission cross sections of these isotopes as functions of the initial enrichment in ²³⁵U, the average burnup and the radial coordinate. The curves obtained with a so drastic simplification demand a careful testing before incorporation in the general fuel behaviour code. This testing is performed via comparison with the reliable reactor codes. The first antecedent in this type of analysis is found in the RADAR model [4] which was validated against the WIMS [5,6] code. The TUBRNP model, included in the TRANSURANUS code [7] and the RAPID model [8] are also based on the same concept. In this work curves fitted for the cross sections of ²³⁵U, ²³⁶U, ²³⁸U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu are obtained from the predictions of the reactor cell codes HUEMUL [9] and CONDOR [10] for an average burnup ranging from fresh fuel to 120 MWd/kgHM and for an initial enrichment ranging from natural uranium to 12%. The final purpose is to extend the application range of the DIONISIO code [11,12,13] (originally designed to predict the fuel behavior in normal operation conditions) to the high burnup domain. The predictions of DIONISIO were compared with a large number of experimental data, obtaining an excellent agreement

[en] This paper summarizes the present status of a computer code that describes some of the main phenomena occurring in a nuclear fuel element throughout its life. Temperature distribution, thermal expansion, elastic and plastic strains, creep, mechanical interaction between pellet and cladding, fission gas release, swelling and densification are modelized. The code assumes an axi-symmetric rod and hence, cylindrical finite elements are employed for the discretization. Due to the temperature dependence of the thermal conductivity, the heat conduction problem is non-linear. Thermal expansion gives origin to elastic or plastic strains, which adequately describe the bamboo effect. Plasticity renders the stress-strain problem non linear. The fission gas inventory is calculated by means of a diffusion model, which assumes spherical grains and uses a finite element scheme. In order to reduce the calculation time, the rod is divided into five cylindrical rings where the temperature is averaged. In each ring the gas diffusion problem is solved in one grain and the results are then extended to the whole ring. The pressure, increased by the released gas, interacts with the stress field. Densification and swelling due to solid and gaseous fission products are also considered. Experiments, particularly those of the FUMEX series, are simulated with this code. A good agreement is obtained for the fuel center line temperature, the inside rod pressure and the fractional gas release. (author)

[en] A recently developed three-dimensional finite element model for pellet-cladding interaction (PCI) based in the Cohesive Zone Method, originally developed to treat discontinuities that progress in the continuum, in the 3.0 version of the DIONISIO code is presented. Given that the numerical technique used in DIONISIO is based on the finite element method to solve differential equations over defined domains, the development of a cohesive zone method provides a very robust formulation to manage the gap between pellet and cladding treated as a discontinuity. The approach resulted in a natural mode to tackle PCI considering specifically the cylindrical geometry of pellet and cladding, being innocuous when contact surfaces are apart, and presenting the expected behaviour through the entirety of contact and in the case that the surfaces were to disjoin any number of times during the simulation. Selected results of the model compared to special experiments designed to simulate the pellet-cladding thermomechanical interaction and under irradiation experiments taken from the IAEA database as well as a deep discussion about the advantages and shortcomings of this methodology are presented. Additionally, allowing the use of three-dimensional geometries, this model offers the possibility of performing simulations of asymmetric scenarios such as certain oblations or eccentricity of the pellet or the cladding, heating conditions regarding the position of the rod in a reactor core, pellet superficial defects as a result of fuel element assembly and ballooning of the cladding in LOCA cases, among others. (author)

[en] In the present work the behavior of fuel pellets for power reactors in the high burnup range (average burnup higher than 50 MWd/kgHM) is analyzed. For extended irradiation periods, a considerable Pu concentration is reached in the pellet periphery (rim zone), that contributes to local burnup, as long as a new microstructure develops, characterized by small grains and large pores as compared with those of the original material. In this region Xe is absent from the solid lattice (although it continues to be dissolved in the rest of the pellet). The porous microstructure in the pellet edge causes local changes in the mechanical and thermal properties, thus affecting the overall fuel behaviour. The evolution of porosity in the high burnup structure (HBS) is assumed to be determinant of the retention capacity of the fission gases released by the matrix. This is the reason why, during the latest years a considerable effort has been devoted to characterizing the parameters that influence porosity. Starting from several works published in the open literature, a model was developed to describe the behaviour and evolution of porosity at local burnup values ranging from 60 to 300 MWd/KgHM. The model is mathematically expressed by a system of non-linear differential equations that take into account the open and closed porosity, the interactions between pores and the free surface and phenomena like pore's coalescence and migration and gas venting. Interactions of different orders between open and closed pores, growth of pores radius by vacancies trapping, the evolution of the pores number density, the internal pressure and over pressure within the pores, the fission gas retained in the matrix and released to the free volume are analyzed. The results of the simulations performed in the present work are in excellent agreement with experimental data available in the open literature and with results calculated by other authors (author)

[en] During the latest ten years the Codes and Models Section of the Nuclear Fuel Cycle Department has been developing the DIONISIO code, which simulates most of the main phenomena that take place within a fuel rod during the normal operation of a nuclear reactor: temperature distribution, thermal expansion, elastic and plastic strain, creep, irradiation growth, pellet-cladding mechanical interaction, fission gas release, swelling and densification. Axial symmetry is assumed and cylindrical finite elements are used to discretized the domain. The code has a modular structure and contains more than 40 interconnected models. A group of subroutines, designed to extend the application range of the fuel performance code DIONISIO to high burn up, has recently been included in the code. The new calculation tools, which are tuned for UO₂ fuels in LWR conditions, predict the radial distribution of power density, burnup and concentration of diverse nuclides within the pellet. New models of porosity and fission gas release in the rim, as well as the influence of the microstructure of this zone on the thermal conductivity of the pellet, are presently under development. A considerable computational challenge was the inclusion of the option of simulating the whole bar, by dividing it in a number of axial segments, at the user's choice, and solving in each segment the complete problem. All the general rod parameters (pressure, fission gas release, volume, etc.) are evaluated at the end of every time step. This modification allows taking into account the axial variation of the linear power and, consequently, evaluating the dependence of all the significant rod parameters with that coordinate. DIONISIO was elected for participating in the FUMEX III project of codes intercomparison, organized by IAEA, from 2008 to 2011. The results of the simulations performed within this project were compared with more than 30 experiments that involve more than 150 irradiated rods. The high number of tests and the good quality of the predictions suggest that DIONISIO can be considered as a reliable prediction tool. The code is presently used for simulating the behavior of the fuel for the CAREM reactor (author)

[en] The version 2.0 of the DIONISIO code has been recently developed with the purpose of improving the accuracy of the simulation of the whole fuel rod. To this end, the rod is divided into a number of axial segments. The local values of linear power and coolant temperature are given as input data to DIONISIO 1.0 which is executed in each segment obtaining as outputs the local values of temperature, stress, strain, among other physical variables. Then, the general rod parameters (internal rod pressure, amount of fission gas released, pellet stack elongation, etc.) are evaluated at the end of every time step, conveniently combining the results of all the axial segments. The new code architecture allows taking into account the axial variation of the linear power and, consequently, evaluating the dependence of all the significant rod parameters with the longitudinal coordinate. Moreover, new calculation tools designed to extend the application range of the code to high burn up have also been incorporated to DIONISIO 2.0 in recent times and are the subject of other presentation. With these improvements, the code results are compared with some experiments published in the IAEA data base, covering more than 380 fuel rods irradiated up to average burnup levels of 40-60 MWd/ng. The results of these comparisons, which are presented here, reveal the good quality of the simulations. (author)

[en] Due to the extension of the permanence time of the nuclear fuels within the reactor, physical and chemical modifications take place in the fuel material, especially in the external ring of the fuel pellet. The codes for simulating the rod behaviour during reactor operation need adequate models to describe these phenomena and be capable of making accurate predictions in the whole burnup range. A complex group of subroutines has been included in DIONISIO to represent the radial distribution in the pellet of the power density, burnup, porosity and concentration of diverse nuclides, particularly those capable of undergoing fissions, in terms of overall parameters like initial enrichment and average burnup. In this work we summarize the models recently developed, related to the high burnup scenario. On the one hand, empirical expressions representing the absorption and capture cross sections of several uranium and plutonium isotopes, as functions of the initial enrichment in "2"3"5U, average burnup and radial coordinate are presented. In addition, new models that give the distribution of porosity, fission gas retention and release in the pellet edge are described. Moreover, an empirical formula that relates the thermal conductivity of the fuel material with the burnup and the content of gadolinium, usually added as burnable poison, is presented. Subroutines corresponding to each of these models have been incorporated to the DIONISIO code. With these improvements the code was used to simulate data provided by the FUMEX I/II/III NEA data bank. The results presented here make evident the good agreement between experiments and simulations. (author)

[en] A mathematical and numerical model for the description of different aspects of microtumor development is presented. The model is based in the solution of a system of partial differential equations describing an avascular tumor growth. A detailed second-order numeric algorithm for solving this system is described. Parameters are swiped to cover a range of feasible physiological values. While previous published works used a single set of parameters values, here we present a wide range of feasible solutions for tumor growth, covering a more realistic scenario. The model is validated by experimental data obtained with a multicellular spheroid model, a specific type of in vitro biological model which is at present considered to be optimum for the study of complex aspects of avascular microtumor physiology. Moreover, a dynamical analysis and local behaviour of the system is presented, showing chaotic situations for particular sets of parameter values at some fixed points. Further biological experiments related to those specific points may give potentially interesting results

[en] Several models were integrated to the DIONISIO code within the framework of the IAEA Research Project “Fuel Modeling in Accident Conditions (FUMAC)”, to take account of accidental conditions, in particular the loss of coolant accidents (LOCA). A specially designed thermal-hydraulic subroutine provides a simplified description of the rod environment in normal or accidental conditions. The heat transfer coefficients that account for the different coolant regimes, in single or double phases, are activated as the corresponding conditions occur. The simulation of a considerable number of experiments has shown that, despite its simplicity this subroutine gives adequate predictions of the conditions in a vertical cooling channel, quite similar to those given by the thermal-hydraulic codes. The description of the fuel rod atmosphere is improved with the incorporation of this subroutine since it provides fairly realistic boundary conditions for the simulation of the fuel rod behavior, without requiring the intervention of external specific codes. Models of high temperature oxide growth (ZrO2) and hydrogen capture and release by the cladding in steam were also included. Moreover, the model of cladding creep predicts the conditions for ballooning and eventually, those for catastrophic failure (burst) and its localization. The calculation scheme makes a partition of the rod length into a number of segments defined by the user. In each segment the local conditions are considered to calculate, with the synchronous work of all the subroutines, the physical and chemical parameters in one representative pellet. Then, a description of the whole rod is obtained by coupling all the segments. This strategy has yielded accurate simulations of a wide variety of cases, either in normal or LOCA type conditions. Exhaustive comparisons were carried out with several thermal-hydraulic codes (COBRA-IV, RELAP5-Mod3.1, SOCRAT, ATHLET-Mod 1.1) and with a number of experiments like those of the IFA–650 series (-1,-2,-9,-10,-11), PUZRY, QUENCH-L0/L1 (for which a new working scheme was specially developed in DIONISIO), CORA-15, IAEA-SPE-4, among others. (author)