[en] CYRANO3 is the new EDF thermomechanical code developed to evaluate the overall fuel rod behavior under irradiation. In that context, this paper presents the phenomena to be simulated and the correlations adopted for modelling purposes. The empirical models presented are taken from the CYRANO2 code and a compilation of the relevant literature. The present revision corrects and supplements version B on the basis of its use during the software coding phase from January 1991 to May 1993. (authors). figs., tabs., 120 refs

[en] Reliable design of LWR fuel rods requires the fission gas release to be predicted as accurately as possible. Indeed that physical phenomenon governs both the fuel temperatures and the inner gas pressure. Fission gas release data have been reviewed by the NRC and it has been concluded that a fission gas release enhancement occurs at burn-up above 20 GWd/tM. To correct deficient fission gas release models which do not include burn-up dependence, the NRC developed an empirical correction method to describe burn-up enhancement effect. BELGONUCLEAIRE has developed its own fission gas release model which is utilized in licensing calculation through the COMETHE code. Fission gas release predictions at high burn-up are confronted to the experimental data as well as to the predictions of the NRC correlation. The physics of the fission gas release phenomenon is discussed

[en] BELGONUCLEAIRE's gradually increasing in-reactor experience has enabled the continuous development and assessment over the years of a coherent set of specifications and drawings for UO2-PuO2 and UO2 fuel for LWR's. On the basis of this experience, design codes have been developed, benchmarked and are thereafter applied to cover completely the whole range of fuel specifications and irradiation histories. The sensitivity of the fuel rod behavior on as fabricated characteristics and on operating conditions (steady and transient) is outlined through calculation results on the COMETHE III-J computer code

[en] In recent LWR designs, the fuel rod failures are induced by a chemically assisted mechanical process, i.e. stress corrosion cracking. The analytical approach towards the analysis of PCI-SCC failures is mainly based on the predictions of the COMETHE code. The failure criteria rely on the concept of a stress threshold together with fission product availability. In the present paper, the use of the COMETHE code to minimize PCI induced clad failure occurrences is illustrated by parametric studies to define acceptable fuel specifications and reactor operating conditions (steady and transient). (author)

[en] Fuel rod performance modelling is sometimes taken in an academical way. The experience of the COMETHE code development since 1967 has clearly shown that benchmarking was the most important part of modelling development. Unfortunately, it requires well characterized data. Although, the two examples presented here were not intended for benchmarking, as the COMETHE calculations were only performed for an interpretation of the results, they illustrate the effects of a lack of fuel characterization and of the power history uncertainties

[en] Chemically assisted mechanical process as stress corrosion cracking is recognized as the principal mode of fracture of Zircaloy in LWR. Clad failure generally occurs during unsteady reactor operation as power ramps. Fission products as iodine, bromine and cesium and their compounds have been recognized as the most aggressive agents in Zircaloy stress corrosion cracking. Some proposed mechanisms of the release of fission products in LWR modern fuel rods of the 17 x 17 type are analyzed. The results of a parametric study performed with the COMETHE III-J code to investigate the effect of power ramping on fission produce release are displayed. The goal of that parametric study was to investigate the possibility of a fission product release enhancement during a power ramp. Three mechanisms leading to such an enhancement are proposed and discussed

No abstract available

[en] This paper first describes the effect of neutron irradiation on the thermomechanical behavior of stress-relieved Zircaloy-4 fuel tubes that have been analyzed after exposure to five different fluences ranging from nonirradiated material to high burnup. In the second part, a viscoplastic model is proposed to simulate, for different isotherms, 350 C < T < 400 C, out-of-flux anisotropic mechanical behavior of the cladding tubes over the fluence range 0 < φ < 100 . 10²⁴nm^-2(E > 1 MeV). The model, identified for tests conducted at 350 C, has been validated from tests made at 380 C and 400 C. The model is capable of simulating strain hardening under internal pressure followed by a stress relaxation period, the loading producing an interaction between the pellet and cladding. Introduction of a state variable characterizing the damage caused by a bombardment with neutrons into the model has allowed us to simulate the irradiation-induced hardening and creep rate decrease, as well as the saturation noticed after two cycles of irradiation (≅45 . 10²⁴nm^-2(E > 1 MeV)) in a pressurized water reactor (PWR). Finally, the numerical simulations show the model is able to reproduce the totality of the thermomechanical experiments

[en] The mechanical static and dynamic behaviour of an LMFBR Core depends on the sub-assembly design and the adopted core restraint concept. It is important to achieve design optimization regarding stress and strain analysis and safety requirements at the very beginning of the project studies. We present and discuss, in the present paper, some of the preliminary studies performed in the frame of the new project. The presentation will cover two aspects: - the analysis of the mechanical static behaviour of the core which has led to the definition of certain essential parameters proper to the sub-assemblies, such as axial location of the interaction levels, clearances, core periphery, ... - the evaluation of the core behaviour under safe shutdown earthquake conditions to confirm the validity of the options proposed for fuel sub-assembly design (shorter sub-assembly with a lighter top structure than the one designed for SUPER-PHENIX, ...). (orig.)

[en] The validation of a code for fuel rod behaviour prediction requires a comparison of its results with corresponding experimental data. Benchmarking of the COMETHE code has been done in parallel with its development, but more tine has been spent on that work than in the development of the models themselves. Three experiments are presented; they have been selected from amongst those used by BN for the calibration as being good examples of various features: (1) The ELP2 experiment, performed in the EL3 reactor by CEA-Saclay and related to fuel restructuring. Results show that behaviour is very well modelled in COMETHE. (2) The BR3/VN post-irradiation data, which show a large sensitivity of the fission gas release to the power level and reveal that coupling between the fission gas release model and the gaseous swelling model is beneficial. (3) The BM01 low density fuel BN pin, irradiated in the FBR RAPSODIE: close agreement is found between the cracking pattern computed by the 'pivot model' and the experimental cold state results. A lot of the benchmarking results arise from the EPRI RP 397 Fuel Rod Modeling Code Evaluation Project. (orig.)