Facing the challenge of predicting the standard formation enthalpies of n-butyl-phosphate species with ab initio methods

Tributyl-phosphate (TBP), a ligand used in the PUREX liquid-liquid separation process of spent nuclear fuel, can form explosive mixture in contact with nitric acid, that might lead to violent explosive thermal runaway. In the context of safety of a nuclear reprocessing plant facility, it is crucial to predict the stability of TBP at elevated temperatures. So far, only the enthalpies of formation of TBP is available in the literature with a rather large uncertainties, while those of its degradation products, di-(HDBP) and mono-(H$_2$MBP}) are unknown. In this goal, we have used state-of-the art quantum chemical methods to compute the formation enthalpies and entropies of TBP and its degradation products di-(HDBP), mono-(H$_2$MBP) in gas and liquid phases. Comparisons of levels of quantum chemical theory revealed that there are significant effects of correlation on their electronic structures, pushing for the need of not only high level of electronic correlation treatment, namely local coupled cluster with single and double excitation operators and perturbative treatment of triple excitations [LCCSD(T)], but also extrapolations to the complete basis to produce reliable and accurate thermodynamics data. Solvation enthalpies were computed with the conductor like screening model for real solvents [COSMO-RS], for which we observe errors not exceeding 22 kJ mol$^{-1}$. We thus propose with final uncertainty of about 20 kJ mol$^{-1}$ standard enthalpies of formation of TBP, HDBP, and H$_2$MBP which amounts to -1281.7$\pm$24.4, -1229.4$\pm$19.6 and -1176.7$\pm$14.8 kJ mol$^{-1}$, respectively, in the gas phase. In the liquid phase, the predicted values are -1367.3$\pm$24.4, -1348.7$\pm$19.6 and -1323.8$\pm$14.8 kJ mol$^{-1}$, to which we may add about -22 kJ mol$^{-1}$ error from the COSMO-RS solvent model. From these data, we predict the complete hydrolysis of TBP to be nearly thermoneutral

Contribution à l'étude du rejet à l'environnement de l'iode radioactif lors d'une séquence accidentelle de type RTGV

Dans une séquence accidentelle de rupture de tube(s) de générateur de vapeur d un réacteur à eau pressurisée (séquence RTGV), une fraction des espèces radioactives présentes dans le circuit primaire est susceptible d être transférée à l environnement. Parmi ces espèces, on porte une attention particulière à l iode qui est le plus dangereux à court terme pour les populations et susceptible de former des espèces volatiles. En fonctionnement normal, le circuit primaire est contaminé par des produits de fission radioactifs à cause de micro fissures qui se développent dans les gaines des crayons combustible.Pour mieux estimer les rejets en cas de RTGV, il est primordial de déterminer la répartition des espèces iodées entre la phase gazeuse et la phase liquide en aval de la brèche ainsi que la granulométrie des gouttes générées (fraction transférée au secondaire) lors du flashing. La première partie de l étude concerne la modélisation du jet diphasique généré à la brèche. Ainsi, un modèle physique a été développé dans le but de calculer la fraction vaporisée en champ proche ainsi que la distribution des gouttes (granulométrie) générée en sortie de brèche. Ce modèle a ensuite été appliqué et validé sur des expériences disponibles dans la littérature (essais conduits à l US/ NRC et à l INERIS). Une seconde partie est consacrée à la modélisation de la spéciation chimique de l iode dans le circuit primaire et à la détermination des coefficients de partage des espèces de l iode (calculs de dynamique moléculaire). Enfin, ces modèles ont été intégrés dans le logiciel de simulation des accidents ASTEC pour calculer le rejet gazeux et liquide lors d une séquence accidentelle type RTGV.In a Steam Generator Tube Rupture (SGTR) accident occurring to a pressurised nuclear water reactor, a fraction of the radioactive species present in the primary circuit is likely to be transferred to the environment. Particular attention is paid to iodine for two reasons; the first one it is well known that iodine is a high contributor to the dose at short term and in second, due to possible formation of volatile species, which could be largely sprayed in the environment. In normal operating conditions, the primary circuit is contaminated with some radioactive products flowing through micro-cracks existing in the fuel rod claddings. To better estimate the releases for SGTR sequence, it is crucial to determine the iodine partition between the gas and the liquid phase downstream the tube break as well as the droplet size distribution generated during the flashing. The first part of the PhD presents a heat and mass transfer model developed to predict the two-phase jet behaviour at the break. The steam fraction is calculated as well as the droplet size distribution upstream the break. Experiments available in the literature (tests conducted at the U.S/NRC and INERIS) are used to validate the model. The second part concerns the modelling of the iodine chemical speciation in the primary conditions (irradiation, low concentration and presence of impurities). For each iodine species, the partition coefficient has been determined either in using literature data or with the help of molecular dynamics computations. Last, this global release modelling has been implemented in ASTEC, the IRSN accident simulation software and the releases have been calculated for one SGTR scenario.LILLE1-Bib. Electronique (590099901) / SudocSudocFranceF

Thermochemistry of CsxByOz type of compounds and microsolvation of CsBO2

International audienc

Lifetime extension of French 900 MWe NPPs: French TSO main conclusions regarding long term sump performance during a severe accident

International audienceIn the framework of the lifetime extension of the French 900 MWe NPPs beyond 40 operating years, EDF has included, in case of severe accident, a strategy to remove the decay heat from the containment without opening the emergency containment filtered venting system. The new dedicated circuit uses an injection line with a heat exchanger connected to the cold leg of the primary coolant circuit and another feeding the sump of the reactor building and a pump qualified to extreme external hazards conditions and to severe accident situations. After the drainage of the RWST (Refueling Water Storage Tank), water is taken from the containment sump in the lower part of the nuclear reactor building using one of the filtering system implemented at the bottom of the containment. Consequently this filtering system was previously used during the LOCA phase, before reaching the severe accident phase. A cooling mobile device (ultimate heat sink) is lined on the heat exchanger by the EDF rescue team “FARN” (Nuclear Rapid Response Force). One of the major issues, common to the LOCA issue, is the clogging of the filtering system due to physical and chemical conditions which can lead to an inadequate net positive suction head (NPSH) margin for the pump during the severe accident and can affect the mechanical integrity of the strainers. Furthermore, despite the filtering system, a part of the debris bypassing the strainers is transported through the dedicated circuit: a second major issue is to ensure that these debris do not damage the pump, degrade the heat exchanger performance or clog other equipment of the circuit like valves and diaphragms.Compared to the debris released in the sump in case of a LOCA, additional debris created during a severe accident come from the core degradation, the erosion of the concrete basemat by the corium and paints damaged in the containment due to high irradiation process.EDF presented its safety demonstration based on studies and experimental program on strainer clogging. IRSN conducted the safety review of EDF safety case and performed simultaneously one severe accident test on the Viktoria loop in Slovakia. The paper will present the French background and IRSN conclusions (dated the end of March 2019) on sumps performance issue in case of severe accident for the 4th periodic safety review of 900MWe NPPs. Some open issues will be discussed

Le rôle du bore sur la spéciation de l'iode dans le circuit primaire

Dans le cadre d un accident nucléaire majeur, les produits résultant de la dégradation des barres de contrôle sont susceptibles d'influencer le transport de l'iode dans le Circuit Primaire (CP) d'un réacteur à eau pressurisée comme l atteste les résultats du programme PHEBUS-PF. Trois essais expérimentaux PHEBUS-PF (FPT0, FPT1 et FPT2) ont été réalisés en présence de barres de contrôle en Argent-Indium-Cadmium (AIC) et un essai PHEBUS-PF (FPT3) avec des barres de contrôle constituées de carbure de bore (B4C). Lors de l essai FPT3 une fraction beaucoup plus importante d iode gazeux à la brèche du CP a été observée. Il est suspecté que la formation de CsI soit limitée au profit de composés de type CsxByOz (en particulier le métaborate de césium CsBO2). Les résultats de cette thèse permettent de consolider des données de type thermochimique concernant les borates de césium qui sont mal connues dans la littérature et d acquérir des premières données cinétiques concernant les réactions conduisant à la formation du CsBO2 (CsI + H3BO3 CsBO2 + HI + H2O et CsOH + H3BO3 CsBO2 + 2 H2O). Afin d accéder à ces grandeurs thermocinétiques, des outils de chimie théorique, la thermodynamique statistique et les théories cinétiques appropriées ont été mis en œuvre en prenant le soin de valider les méthodes employées. Ces données ont été prises en compte dans le code de simulation des accidents graves ASTEC (Accident Source Term Evaluation Code) et permettent de réconcilier les résultats de la simulation avec les données expérimentales concernant l iode gazeux à la brèche pour l essai PHEBUS FPT3.As part of a major nuclear accident, the products resulting from the degradation of the control rods are likely to influence the transport of iodine in the Reactor Coolant System (RCS) of a pressurized water reactor as evidenced by the results of the Phebus-FP program. Three experimental Phebus-FP tests (FPT0, FPT1, and FPT2) were performed with Silver-Indium-Cadmium (AIC) control rods whereas in FPT3 test, the control rod is boron carbide (B4C). For FPT3, a much larger fraction of gaseous iodine was observed at the RCS break. It is suspected that the CsI (caesium iodide) formation has been restricted due to CsxByOz (especially caesium metaborate CsBO2) formation. The PhD results allow us to consolidate thermochemical data on cesium borates, which are poorly known in the literature, and to get first kinetic data for reactions leading to the formation of CsBO2 (CsI + H3BO3 CsBO2 + HI + H2O and CsOH + H3BO3 CsBO2 + 2 H2O). In order to estimate these thermokinetic parameters, theoretical chemistry tools were used, with the help of statistical thermodynamics and appropriate kinetic theories ; a special care was dedicated to the validation of the applied methodologies. All data have been implemented in the severe accident simulation software ASTEC (Accident Source Term Evaluation Code) and allow us reconciling the simulation results with experimental data concerning gaseous iodine at the break for FPT3 test.LILLE1-Bib. Electronique (590099901) / SudocSudocFranceF

Étude théorique du rôle du bore sur le comportement de l’iode dans le circuit primaire d’un réacteur

International audienc

Étude théorique du rôle du bore sur le comportement de l’iode dans le circuit primaire d’un réacteur

International audienc

Étude théorique du rôle du bore sur le comportement de l’iode dans le circuit primaire d’un réacteur

International audienc

A theoretical study of the kinetics of gas-phase elementary reactions containing caesium species of nuclear safety interest

International audienc

A theoretical study of cesium borates compounds

International audienc