[en] Phenomena limiting the lifetime of the RCCAs in PWRs have already been highlighted in the past. Wear of the rodlet cladding and swelling in the lower tip end of the cladding arising from the structural changes in the Ag-In-Cd absorber material are the two damage mechanisms encountered: wear, potentially leading first to cladding perforation secondly further later rod mechanical failure; and swelling, causing clad cracking and rod swelling possibly culminating in RCCA jamming in the fuel assembly guide thimble. The upgrades applied to the new generation HARMONI^TM RCCAs have provided a solution to the RCCA wear problems and delayed contact between absorber and cladding. Better knowledge and prediction of Acceding swelling in the reactor is necessary, however, as it conditions the lifetime of the RCCAs. An extensive analysis programme of this Ag-In-Cd absorber was undertaken by FRAMATOME in conjunction with EDF, in order to determine the causes of its out-of-pile and in-reactor changes (creep and metallurgical changes in the material under irradiation), predict its behaviour and make structural improvements limiting its in-reactor changes. The outcome of this analysis was the HARMONI^TM 2G design featuring an improvement in the Ag-In-Cd material. Turning to the 1300 MWe RCCAs containing boron carbide (B₄C), an experiment on the behaviour of this absorber has been conducted in an EDF reactor. B₄C was inserted into claddings pre-perforated at the fabrication stage and subjected to corrosion by the reactor coolant. The first lessons on the conditions of boron carbide dissolution have been drawn. The improved knowledge in the behaviour of these absorbers in the reactor and the design enhancements made are allowing prediction of a lifetime with respect to swelling which is further extended for new generation HARMONI^TM RCCA relative to the first HARMONI generation. A development programme on an RCCA design using hafnium absorber is also in progress. The use of hafnium would completely rule out swelling in the lower tip end of the RCCAs. (author)

[en] Since 1988 EdF has decided to inspect the RCCA of all the nuclear plants. This decision has reinforced by two incidents which occurred in 1988 and 1989 where one rod broke by excessive wear. The inspection allowed to identify two problems: Clad wear due to fretting; clad swelling which could induce cracks. To address these problems essentially three actions are initiated: The first concerns the immediate requirements and consists in defining a RCCA management strategy based on mechanical calculations. This strategy includes controls, axial repositioning and specific rejection criteria. Second concerns some design improvements of RCCA. Regarding the wear problem, various hardening methods of the clad have been investigated (nitride, chrome carbide, chrome). For the swelling, an increase of the gap between the absorber and the clad should delay their interaction. Third, is a long term work and more ambitious approach. It consists in a large R and D program concerning the study of the phenomena involved in three wear process. The aim is to build a full set of models and to gather it in a computer code, for predicting RCCA wear. The state of these actions is the following: The first action has allowed Utilities to operate their power plants in safety conditions, even with standard RCCA. In the second, out of pile tests have demonstrated the potential efficiency of various remedies. These remedies are progressively introduced, now about 180 new design RCCA are in reactors. The first controls confirm the good behaviour of these solutions. The third action is still in progress. EdF has planned to qualify the code in 1995. Then it will be possible to perform analytical studies to investigate modifications on the design of RCCA or internal guide tubes. (author). 2 refs, 6 figs, 2 tabs

[en] Today the use of enhanced nuclear fuels that can sustain higher burnups has allowed a better optimization of the fuel management in nuclear power plants. The optimization for the near future is based on 3 aims: -) a better competitiveness of nuclear energy, longer campaigns mean a higher availability and less refueling so it has a direct impact on costs, -) a better flexibility to meet energy demand: a modulation of cycle lengths by more or less 2 months is possible by introducing or withdrawing 8 assemblies in the refueling load, this modulation will allow an optimization of the scheduling of the refueling shutdowns with respect to the seasonal energy demand peaks, -) a reduced volume of spent fuels (but with a higher level of radioactivity). (A.C.)

[en] Plutonium recycling in PWR's started in France in 1987 with the first reload containing MOX fuel in the Saint-Laurent B1 reactor. By the end of 1994, more than 400 MOX fuel assemblies had been delivered by Fragema to 7 different EDF 900 MW power plants. As the number of PWR units recycling Plutonium is increasing, MOX fuel is required to assume the same operational flexibility as UO2 fuel. So, MOX fuel must follow UO2 fuel developments (1/4 core fuel management, load follow) in terms of discharge burnup and maneuverability. The acquisition of new irradiation data of MOX fuel representative of the current product and operation mode allows to optimize the fuel design leading to improved performance. The development programme set up by French partners CEA, EDF and FRAMATOME, analyzes MOX fuel behaviour through: analytical experiments in experimental reactors conducted under normal and transient operation and, surveillance programmes consisting of the irradiation in commercial reactors and the subsequent examination of MOX fuel assemblies and fuel rods. The paper describes our ongoing programme and discusses the main results obtained so far with a particular focus on the surveillance programme that has been set up on the first MOX reload. 9 refs, 7 figs, 1 tab

[en] To improve the knowledge of mixed oxide fuel rods behaviour, two prototype assemblies have been irradiated in the fourth core of the CAP reactor (Chaudiere Avancee Prototype) in Cadarache. The irradiation campaign started on the 29th of September 1985 and ended on the 2nd of July 1987, reaching 419 effective full power days, without any rod failure. The average burn-up of the MOX rods is evaluated at 20 GWd/tM, which is roughly equivalent to two commercial reactor cycles. Moreover, the reactor was operated under load following and frequency control mode. As a consequence, the MOX rods have undergone power cycles similar to those in an EDF commercial reactor. After irradiation, both assemblies were dismantled in the reactor handling and storage pool, and rods were shipped to CEN Saclay for post-irradiation examinations. This paper presents some preliminary results obtained on six MOX fuel rods examined in the hot cells of the LECI (Laboratoire d'Etude du Combustible Irradie). (author). 6 refs, 5 figs

[en] The nuclear microprobe technique and the mass spectrometry technique have been used to measure respectively the radial and the axial ¹⁰B(n,α)⁷Li profiles after neutrons irradiation nf B₄C absorber (1 cycle in PALUEL 2 reactor). In spite of very low burn-ups (from 0.3 to 5 x 10²⁰ capt./cm³), both techniques show very good sensitivities. The results obtained are in good agreement. This permits to have a better knowledge of the neutronic environment of the PWR control rod. (authors)

[en] Some elements of the study of fibrous structure are presented in this paper. X-ray synchrotron microtomography is used to provide information about the three-dimensional structure. In this context a segmentation method is proposed to separate the different components of the porous material considered (paper)

[en] The experimental procedure is described: Short rods, equipped with centreline thermocouples and gas sweeping lines (Grimox), were irradiated at various power levels in a pool reactor, in a pressurized water loop. For better calibration, the rods were separated in two parts, the lower one made of standard UO₂ pellets and the upper one of MOX fuel. Centreline temperature and fission gas release kinetics were measured. First comparative results between UO₂ and MOX centreline temperatures are presented, for a given linear heat rate under the particular irradiation conditions of the experiment. To be meaningful, the temperature difference observed in the MOX fuel needs to be analysed through a detailed calculation of all the components of the thermal behaviour. These terms are analysed, especially the radial profile of specific power inside the MOX fuel pellets, which was derived from the analysis of gamma scanning along a diameter of a cross section. An expression for the integrated value of the MOX thermal conductivity is planned to be published soon on the basis of this analysis; as are the results of fission gas release measurements. (author). 2 figs

[en] Up to now, more than 1000 MOX fuel assemblies have been loaded in the French 900 MWe PWR since the first reload in the Saint-Laurent B1 reactor in 1987. The recycled fuel irradiation experience feedback built up in this way, including NPPs operated in load-following mode, demonstrates the outstanding reliability of this product. Nevertheless, the current hybrid fuel management of the reactors with MOX fuel restrains the average MOX fuel discharge burn-up. The goal of the 'MOX parity' project is to achieve parity between UO₂ and MOX fuel performance with a significant increase of the batch average discharge burn-up. The preliminary phase was completed end 1998. It has demonstrated the technical feasibility of an annual quarter core reload type management, by means of the use of the new AFA 3G MOX fuel product and of the implementation of NSSS system adaptation in order to meet the safety criteria. March 1999, EDF decided to start the detail design phase for a gradual introduction of this new UO₂-MOX fuel management in its PWR 900 MWe NPPs. These studies will benefit from the use of advanced design and licensing methodologies, for the safety analysis as well as for the product performance standpoint. Surveillance programs and R and D programs specific to the MOX product will support them. The aim of this paper is to present the main results of the preparatory phase. (authors)

[en] One of the criteria for rejecting COntrol Rod Assemblies is the clad cracking linked to the diameter increase at the bottom of the rods. Examinations in Hot Cells have been performed to understand the rod behaviour under operating conditions. Swelling of absorber rod is due to transmutations which lead to modification in the chemical composition of the alloy and, further, to a new phase formation. The cold worked structure of the clad is replaced by many small dislocation loops which lead to grain hardening and low ductility. Under the combined effects of swelling under irradiation and absorber slumping because of creep and repeated shocks, the absorber comes into contact with the cladding. When the gap is closed, the absorber applies a strain to the cladding which cracks. However, the intergranular cracking cannot be only explained by the drastic decrease of cladding ductility under irradiation. As a short-term remedial action, FRAMATOME combines a reduction of diameter in the lower part of the rod with helium rod filling and, in some cases, a decrease of the spring hold down force. (authors). 6 figs., 6 tabs., 11 refs