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[en] This report presents the mechanisms of fission product release out of the fuel, and more particularly depending on temperature (750 to 20000C). The prediction model currently used at the IPSN, developed from experiments, is finally recalled. The knowledge of the fission product release history out of the fuel during an accidental transient involves this one of the fission product distribution between the UO2 pellets and the fuel-can space in the pre-accidental phase (fuel operating in normal conditions). 5 figs
[fr]Ce rapport fait le point sur les mecanismes de relachement des P.F. (Produits de Fission) hors du combustible tels qu'ils sont percus actuellement, en fonction notamment de sa temperature (750 a 20000C). Le modele de prevision couramment utilise a l'IPSN, d'origine experimentale, sera enfin rappele. La connaissance de la chronologie du degagement des P.F. hors du combustible lors d'un transitoire accidentel implique notamment celle de la distribution de ces P.F. entre les pastilles d'UO2 et l'espace gaine-combustible dans la phase preaccidentelle (combustible en conditions de fonctionnement normal). 5 figs.
[en] The aim of this program (four experiments FLASH 01 to 04) is the determination of the quantity of fission products released by a fuel rod during an accidental transient of the LOCA type, as well at the moment of the cladding burst as during the reflooding phase. 4 figs
[fr]Ce programme (4 essais reperes FLASH 01 a 04) a pour principal objective la determination de la quantite de produits de fission relachee par un crayon combustible lors d'un transitoire accidentel de type Accident de Perte de Refrigerant Primaire (APRP), tant au moment de la rupture de la gaine qu'au cours de la phase de renoyage. 4 figs.
[en] The aim of researches developed now in France on water reactor safety is to obtain means and knowledge allowing to control accidental situations, including severe situations beyond design basis accidents. The main studies and researches concerning water reactors and described in this report are the following ones: core cooling accident and prevention of severe accidents, fuel behavior in accidental situation, behavior of the containment building, fission product transfer and releases in case of accident, problems related to equipment aging, and, methodology of risk analysis and ''human factor'' studies. Most of these studies follow an analytic approach of phenomena
[fr]Les recherches effectuees en France en surete des reacteurs a eau ont actuellement pour motivation principale l'acquisition des moyens et connaissances permettant de maitriser les situations accidentelles, y compris les situations severes ou degradees au-dela des accidents de dimensionnement. Les principaux themes d'etudes et de recherches concernant les reacteurs a eau et decrits dans le present rapport sont les suivants: accidents de refroidissement du coeur et prevention des accidents severes, comportement du combustible en situation accidentelle, comportement des enceintes de confinement, transfert et rejets de produits de fission en cas d'accident, problemes lies au vieillissement du materiel, et, methodologie des etudes de risque et etudes du facteur humain. L'essentiel des etudes engagees dans ces differents domaines sont relatives au deroulement des accidents et procedent, pour la plupart, d'une approche analytique des phenomenes
[en] A large experimental in-pile program has been set up at the PHEBUS facility to investigate the actual behavior of .8 m active height, 25-rod PWR-type pressurized fresh fuel bundles under typical accident conditions. The program consists of four stages. Stage 1 was devoted to the adjustment of the operational procedure for stage 2. Stage 2 refers to the simulation of conservatively calculated L.B. LOCA 2 - peak transients. Stages 3/4 refer to four PWR severe accident scenarios retained for in-pile simulation at PHEBUS: a) a large break LOCA with injection failure; b) a small break LOCA associated with an injection failure; c) a prolonged total loss of the steam generator feedwater; and, d) a prolonged core uncovery a few days after reactor shutdown. The main PHEBUS stage 2 results are presented and finally interpreted
[en] French safety philosophy as concerns severe PWR accidents has already been outlined by the Director of CEA/IPSN in an article published in ''Nuclear Safety''. Therefore the present paper will focus on: a) the French reference source terms, as used for elaborating ultimate emergency procedures on PWRs and for emergency planning; b) the methods currently developed for more realistic assessments of the release of fission products during a severe accident
[en] During the course of a severe core accident, hydrogen evolving, mainly due to zircaloy cladding-steam reaction, may form early a flammable mixture in the containment. The risk of a short-term containment failure due to a hydrogen explosion, which would result in a large radioactivity release into the environment, is currently being assessed for the various types of large dry containments existing for French PWRs. In this framework, comparisons between pressure peaks, due to the most severe conceivable hydrogen deflagration, and the realistic ultimate strength of French containment buildings have led to the conclusion that containment integrity should not be questioned. However the problem of the likehood of local detonations and of their impact on structures is still examined. Another aspect of the hydrogen risk is the possible impairment of safety-related equipment in the containment during an eventual hydrogen combustion or explosion. Information coming from the EPRI research program on hydrogen combustion and control, in which the safety body (CEA/IPSN) and the utility (EDF) jointly took a participation, is to back up our own studies on hydrogen risk analysis. Up to now all the results gathered as regards the integrity of the large, dry containments of the French PWRs against hydrogen explosion tend to relegate such a risk to the level of a residual risk. Although the studies are going on a drastic change of this trend in the future is not expected. Therefore no hydrogen specific design modification is presently required by the safety authorities
[en] French studies have early considered the fact that, despite all the precautions taken, the possibility of severe accidents cannot be absolutely excluded; these accidents include core meltdown and a more or less significant loss, at an early or later stage, of the confinement of the radioactive substances in the containment. For a given scenario, one can almost always imagine a more severe scenario by postulating additional failures, but it is obvious that, as the severity of the imagined scenario increases, the probability of its occurrence tends towards zero. However, it does not appear reasonable to attempt to set a probability threshold below which the scenarios should be excluded. First of all, the higher the improbability of the scenarios, the greater the uncertainty in the calculation of their probability, with the result that the calculation is not very meaningful. Secondly, and more importantly, this approach ignores the essential problem of accident situation management. From the outset, French studies have been focused on controlling the development of these situations and mitigating their consequences by means of a series of appropriate actions involving, on the one hand, optimum use of the resources available in the installation during the course of the accident and, on the other hand, the taking of protective measures for the population. To attempt to prevent an initial event to degenerate into a severe accident leading to core meltdown if the proper actions are not taken, Electricite de France has proposed a new operating procedure based on the characterization of every possible cooling state of the core
[en] The Cupidon code has been developed to analyze the thermo-mechanical behaviour of a fuel rod during a Loss Of Coolant Experiment. Models included are drawn from out-of-pile results such Edgar and the first use is to predict and calculate the tests carried out in the Phebus facility. Although the Flash program initiated at Grenoble is devoted to the study of fission product release during a LOCA (Loss Of Coolant Accident), interesting informations have been obtained on in-pile cladding deformation during transients. Analyses of the PIE (Post Irradiated Examination) results in the two first experiments with Cupidon code have shown fairly good agreement regarding diametral strain
[en] French PWR power plant design relies basically on a deterministic approach. A probabilistic approach was introduced in France in the early seventies to define safety provisions against external impacts. In 1977 an overall safety objective was issued by the safety authority in terms of an upper probability limit for having unacceptable consequences. Additional measures were taken (the ''H'' operating procedures) to complement the automatic systems normally provided by the initial design, so as to safisfy the safety objective. The TMI-2 accident enhanced the interest in confused situations in which possible multiple equipment failure and/or unappropriate previous actions of the operators impede the implementation of any of the existing event-oriented procedures. In such situations, the objective becomes to avoid core-melt by any means available: this is the goal of the Ul symptom-oriented procedure. Whenever a core-melt occurs, the radioactive releases into the environment must be compatible with the feasibility of the off-site emergency plans; that means that for some hypothetical, but still conceivable scenarios, provisions have to be made to delay and limit the consequences of the loss of the containment: the U2, U4 and U5 ultimate procedures have been elaborated for that purpose. For the case of an emergency, a nationwide organization has been set up to provide the plant operator with a redundant technical expertise, to help him save his plant or mitigate the radiological consequences of a core-melt