[en] This paper is concerned with the evolution of the microstructure of cementitious materials subjected to high temperatures and subsequent resaturation in the particular context of long-term storage of radioactive wastes, where diffusive and convective properties are of primary importance. Experimental results obtained by mercury intrusion porosimetry (MIP) are presented concerning the evolution of the pore network of ordinary portland cement (OPC) paste heated at temperatures varying between 80 and 300 deg. C. The consequences of heating on the macroscopic properties of cement paste are evaluated by measures of the residual gas permeabilities, elastic moduli and Poisson's ratio, obtained by nondestructive methods. Resaturation by direct water absorption and water vapour sorption are used to estimate the reversibility of dehydration. The results provide some evidence of the self-healing capacity of resaturated cement paste after heating at temperatures up to 300 deg. C

[en] The behaviour of advanced cladding materials under challenging conditions needs to be fully characterized and understood. The Studsvik Cladding Integrity Project (SCIP) is aimed at studying the dominant failure mechanisms of LWR cladding under pellet-clad mechanical interaction loadings (i.e., pellet-cladding interaction, PCI; hydride embrittlement, HE; and delayed hydride cracking, DHC). Besides the experimental work, analytical activities have been launched within the project, both to support test interpretation and to validate the available models against the experimental data base obtained. This paper summarizes the main outcomes resulting from a recent code benchmark exercise set up under the frame of the SCIP project. Further than assisting test interpretation, the main objective was to check the code capability to model ramp scenarios. Nonetheless, as the fuel rods had been previously irradiated in commercial reactors up to high burn-up values and subsequently characterized experimentally, the code results for the base irradiation period were also examined. The comparison exercise was structured in two steps: a preliminary phase, devoted to tune the code to the experimental scenarios; and, the true benchmark phase. Each of them consisted in the modelling of four ramp tests dealing with at least three type of cladding materials and denoted by the rodlets names: KKL-4, M5-H1, Z2 and Z3 in the former, and KKL-1, M5-H2, O2 and Z4 in the latter. Both the tests specifications (i.e., rodlet design data, in-reactor power history, power profiles, etc.) and the guidelines for reporting the results were defined. Four codes have been used: ALCYONE v1.1, FALCON-PSI, FRAPCON-3 v3.3 and STAV7.3. A set of hypotheses and approximations were made in each of the codes regarding both the boundary conditions (i.e., power histories, inlet coolant temperature, re-fabrication, etc.) and the fuel and clad modelling (i.e., densification, rim porosity, materials properties, etc.). Their predictions have been compared to data in terms of cladding oxidation, diameters and elongation. Predictability of clad oxidation was certainly scattered and while some codes showed reasonable accuracy, other results were notably deviated. As for diameters, most of the codes were capable of qualitatively capturing the axial profile, and showed consistency between diameters and hoop stress and strain predictions. Elongation estimates were generally poor, and were rather far from measurements in most cases (even the trends observed were just vaguely followed by the codes). Additionally, some other variables (i.e., fuel temperature, contact pressure, hoop and axial stresses, etc.) were code-to-code compared and some systematic tendencies of specific codes were noted. In summary, the benchmark has highlighted some drawbacks of current mechanical modelling and areas where further model development is necessary have been identified, both generally and in specific codes. An improved code performance would mean a more effective use of analysis as a tool to understand rod failure mechanisms. This benchmark was coordinated by CIEMAT under the sponsorship of the Nuclear Regulatory Body of Spain (CSN). (authors)

[en] In this paper, three power ramp tests performed on high burn-up Recrystallized Zircaloy2 - UO₂ BWR fuel rods (56 to 63 MWd/kgU) within the SCIP project have been simulated with METEOR and ALCYONE 3D. Two of the ramp tests are of staircase type up to Linear Heat Rates of 420 and 520 W/cm and with long holding periods. Failure of the 420 W/cm fuel rod was observed after 40 minutes. The third ramp test consists in a more standard ramp test with a constant power rate of 80 W/cm/min till 410 W/cm and a short holding time. The tests were first simulated with the METEOR 1D fuel rod code leading accurate results in terms of profilometry and fission gas releases. The behaviour of a fuel pellet fragment and of the cladding piece on top of it was then investigated with ALCYONE 3D. The size and the main characteristics of the ridges after base irradiation and power ramp testing were recovered. Finally, the failure criteria validated for PWR conditions and fuel rods with low to medium burn ups were used to analyze the failure probability of the KKL rodlets during ramp testing

[en] CEA is developing the ALCYONE fuel performance code for PWR fuel in the PLEIADES software environment. It is dedicated to normal, off-normal and accident conditions such as RIA and LOCA. CEA’s participation to the IAEA FUMAC CRP led to an improvement of the fuel modelling in LOCA conditions. Specific developments of the fuel performance code for the LOCA conditions have been done regarding cladding behaviour modelling, fission gas release and stress evaluation in the pellet before and during the tests. The improved code has been used to simulate some of the experiments of interest of the FUMAC project (IFA650.10 and Studsvik 192 LOCA test), the paper summarizes the results. For IFA650.10, the cladding outer temperature profile calculated with the SOCRAT code and provided to the participants of the FUMAC CRP has been used. The results obtained with ALCYONE are in a good agreement with the experimental data. In terms of uncertainty quantification, it seems that the uncertainty on the determination of the boundary conditions like cladding outer temperature results in a large uncertainty on the cladding deformation and the burst time. Recent developments have also been done in ALCYONE to improve the modelling of fuel behaviour in RIA conditions, in particular about fuel mechanical behaviour and the consequences on fission gas release. (author)

[en] This paper presents a service life model for concrete leached by pure water. The tool is based on the generalized transport equation of calcium. The equation is solved numerically using the CAST3M mixed-hybrid finite element code. This model simulates the evolution of concrete degradation in 2D, with the possibility to account for the influence of cracks. Within this model, a second transport equation has been implemented to determine the radionuclides (RN) diffusion. The radioactive decay and the retention of the RN on the concrete are taken into account. In this paper, calculations are performed to evaluate the release of various RN from a (cracked) concrete container leached by pure water. Significant results are obtained regarding the impact of retention, leaching and cracking on the RN release. Further work will consist in integrating other types of degradation in the service life model such as carbonation and sulphate attack. In conclusion: in this study, a service life model allowing to perform radionuclide release calculations through a concrete submitted to leaching and cracking has been presented. The release of four radionuclides (¹³⁷Cs ²⁴¹Am, ²²⁶Ra and ²⁴³Am) have been studied in the framework of these calculations. The K_d values of the studied radionuclides correspond to a low level (¹³⁷Cs), a mean level (²²⁶Ra) and a high level (²⁴³Am) of retention. Calculations carried out have shown that the retention of the radionuclides on the matrix decreases considerably the released quantities. More precisely, the reduction in the released quantity due to the retention increases with the retention level of the radionuclide is high, but also with the shortening of its life time. For a given radionuclide, the couple (value of K_d/radioactive half-life) is thus predominant with regards to the released quantity through the concrete container. In addition, it should be noted that leaching causes a significant increase in the released quantity when the radionuclide life time is large compared to the duration necessary to degrade all the concrete thickness. This increase is related to the modified transport properties (diffusion coefficient, porosity) that originate from leaching of concrete. It reaches 10 orders of magnitude in the case of ²⁴³Am between configuration 2 (case with retention and leaching) and configuration 1 (case with retention and without leaching). Cracking can also have a significant impact if the duration necessary to degrade all the concrete thickness is higher than the RN life time. In this case, cracking represents a propagation vector for leaching and induces an increase in the degraded surface, in particular in the depth of material (near the crack tip). This increase in the degraded surface can in turn induce a significant increase in the released quantities. For example, the released quantity of ¹³⁷Cs is multiplied by a factor 4 when comparing the results obtained in configuration 3 (case with retention, leaching and cracking) and configuration 2 (case with retention, leaching and without cracking). This increase, corresponding to a low density of skin cracking, may become more important for greater crack depths and densities. Concerning the diffusion in the crack (not taken into account in the simplified model), the results obtained with the Diffu-Ca code emphasize its weight when the RN life time is important and allows the release through the crack to occur in a significant way. Further work will consist in studying the influences of greater crack depths and densities on the released quantities. It is also envisaged to integrate other types of degradation in the service life model, such as carbonation and sulphate attack. (authors)

[en] After various thermal treatments (up to 450 deg C), residual thermo-hydro-mechanical (T-H-M) properties of two OPC high performance concretes (HPC) were analyzed in the context of surface long-term storage. Materials were prepared with silico-calcareous aggregates (standard HPC) and hematite aggregates (heavy HPC). The initial micro structural (porosity =10%) and transport (gas permeability ∼ 10^-19 m²) properties are similar for both concretes. As far as the mechanical aspect is concerned, heavy HPC shows a higher compressive strength and elastic modulus than standard HPC (78 and 63 MPa, 81 and 49 GPa, respectively). Heavy HPC is also characterized by a higher thermal conductivity (7.3 W m^-1 K^-1 compared to 2.7 W m^-1 K^-1 for standard concrete). Results analysis show that thermo-hydro-mechanical damages are smaller for heavy HPC. Between 60 and 250 deg C, the elastic modulus and the compressive strength of standard HPC decrease by 40% and 16%, respectively. For heavy HPC, these parameters respectively decrease by 10% and 4% A similar trend was observed for thermal conductivity evolution. Gas permeability and porosity data confirm the good behavior of heavy HPC. As a conclusion, hematite HPC seems to provide more interesting T-H-M residual properties than standard HPC. Limited thermal expansion and thermal gradients induced by hematite are probably responsible of this behavior. Copyright (2001) Material Research Society

[en] The present study describes the in-situ electrochemical modifications which affect irradiated PWR UO_2 fuels in the course of a power ramp, by means of in-situ oxido-reduction indicators such as chromium or neo-formed chemical phases. It is shown that irradiated fuels (of nominal stoichiometry close to 2.000) under temperature gradient such as that occurring during high power transients are submitted to strong oxido-reduction perturbations, owing to radial migration of oxygen from the hot center to the cold periphery of the pellet. The oxygen redistribution, similar to that encountered in Sodium Fast Reactors fuels, induces a massive reduction/precipitation of the fission products Mo, Ru, Tc and Cr (if present) in the high temperature pellet section and the formation of highly oxidized neo-formed grey phases of U_4O_9 type in its cold section, of lower temperature. The parameters governing the oxidation states of UO_2 fuels under power ramps are finally debated from a cross-analysis of our results and other published information. The potential chemical benefits brought by oxido-reductive additives in UO_2 fuel such as chromium oxide, in connection with their oxygen buffering properties, are discussed.

[en] This paper focuses on the PLEIADES fuel performance software environment and its application to the modeling of pellet-cladding interaction (PCI). The PLEIADES platform has been under development for 10 yr; a unified software environment, including the multidimensional finite element solver CAST3M, has been used to develop eight computation schemes now under operation. Among the latter, the ALCYONE application is devoted to pressurized water reactor fuel rod behavior. This application provides a three-dimensional (3-D) model for a detailed analysis of fuel element behavior and enables validation through comparing simulation and post-irradiation examination results (cladding residual diameter and ridges, dishing filling, pellet cracking, etc.). These last years the 3-D computation scheme of the ALCYONE application has been enriched with a complete set of physical models to take into account thermomechanical and chemical-physical behavior of the fuel element under irradiation. These models have been validated through the ALCYONE application on a large experimental database composed of approximately 400 study cases. The strong point of the ALCYONE application concerns the local approach of stress-corrosion-cracking rupture under PCI, which can be computed with the 3-D finite element solver. Further developments for PCI modeling in the PLEIADES platform are devoted to a new mesh refinement method for assessing stress-and-strain concentration (multigrid technique) and a new component for assessing fission product chemical recombination. (authors)

[en] In this paper we study the brittle behavior of UO₂ nuclear fuel. Firstly, we present the interpretation of bending tests with three different approaches to assess rupture parameters (critical stress and surface energy). Secondly, we present Vickers' indentation tests on fresh UO₂ fuel. The comparison between bending and indentation tests on fresh fuel allows us to assess the parameters necessary to derive the critical stress and the surface energy from indentation tests. Vickers' indentation is then used to evaluate rupture parameters of irradiated fuels. At the end, we present some applications to fuel rod modeling taking into account the different rupture mechanisms. (authors)

[en] A “multi design” new generation software environment called PLEIADES has been developed by the CEA in the framework of a research cooperative program with EDF and AREVA. In this general software environment, ALCYONE is the PWR fuel performance simulation code. It is a multi-dimensional simulation software (1D, 2D and 3D), with applications for normal, transient and accidental conditions. It also has several levels of modelling, from industrial models to mechanistic ones depending on the amount of multi-scale details expected in the results of the simulation. The different dimensional schemes share the same thermomechanical Finite Element Method code CAST3M. The 1D scheme describes the behaviour of the whole rod and gives access to integral values such as rod fission gas release, clad profilometry and elongation. The 3D scheme allows a local study of Pellet Clad Mechanical Interaction (PCMI) by modelling the thermo-mechanical behaviour of one or several pellet fragments and overlying cladding. The 2D scheme is a compromise between calculation time and the accuracy of the local fuel description. Recently the 3D approach has been extended to a short fuel rod model in order to simulate the ballooning phenomenon during accidental transients. In this paper, we will present the general description of the ALCYONE simulation code in the PLEIADES environment (general computation algorithm, advanced fission gas model for UO_2 and MOX fuels, 3D computation scheme). A focus will be presented on specific developments which have already been done to simulate accidental conditions such as LOCA and fast transients for different dimensional models. (author)