Within the framework of the ASTRID project, core design studies are being conducted by the CEA with support from AREVA and EDF. The pre-conceptual design studies are being conducted in accordance with the GEN IV reactor objectives, particularly in terms of improving safety. This involves limiting the consequences of 1) a hypothetical control rod withdrawal accident (by minimizing the core reactivity loss during the irradiation cycle), and 2) an hypothetical loss-of-flow accident (by reducing the sodium void worth).Two types of cores are being studied for the ASTRID project. The first is based on a 'large pin/small spacing wire' concept derived from the SFR V2b, while the other is based on an innovative CFV design. A distinctive feature of the CFV core is its negative sodium void worth.In 2011, the evaluation of a preliminary version (v1) of this CFV core for ASTRID underlined its potential capacity to improve the prevention of severe accidents.An improved version of the ASTRID CFV core (v2) was proposed in 2012 to comply with all the control rod withdrawal criteria, while increasing safety margins for all unprotected-loss-of-flow (ULOF) transients and improving the general design.This paper describes the CFV v2 design options and reports on the progress of the studies at the end of pre-conceptual design phase 1 concerning:- Core performance,- Intrinsic behavior during unprotected transients,- Simulation of severe accident scenarios,- Qualification requirements.The paper also specifies the open options for the materials, sub-assemblies, absorbers, and core monitoring that will continue to be studied during the conceptual design phase.

The CEA/DEN modeling and computation results with the CATHARE, CRONOS2, and FLICA4 codes of the Organisation for Economic Co-operation and Development boiling water reactor turbine trip benchmark are presented. The first exercise of the benchmark to model the whole reactor thermal hydraulics with specified power has been performed with the CATHARE system code. Exercise 2, devoted to core thermal-hydraulic neutronic analysis with provided boundary conditions and neutronic cross sections, has been carried out with the CRONOS2 and FLICA4 codes. Finally, exercise 3, combining system thermal hydraulics and core three-dimensional thermal-hydraulics-neutronics, was computed with the three coupled codes: CATHARE, CRONOS2, and FLICA4.Our one-dimensional thermal-hydraulic reactor computation agrees well with the benchmark reference data and demonstrates the capacities of CATHARE to model a turbine trip transient. Coupled three-dimensional thermal-hydraulic and neutronic analysis displays a high sensitivity of the power peak to the core thermal-hydraulic model. The use of at least 100 channels is recommended to achieve reasonable results for integral and local parameters. Deviations between experimental data and exercise 3 results are discussed: timing of events, core pressure drop, and neutronic model. Finally, analysis of extreme scenarios as sensitivity studies on the transient to assess the effect of the scram, the bypass relief valve, and the steam relief valves is presented.

Since several years, a R&D action, based on numerical simulation and experiments of flow into PWR vessels was initiated at Framatome and CEA. The aim is to reach a better understanding of the thermal hydraulic phenomena and their consequence into the whole nuclear vessel. This paper presents a comparison between numerical simulation and experimental data obtained using PIV technique. Measurements have been performed into the hot legs of a 1300 MW PWR mock-up. The experiment was carried out on a scaled down PWR model. The facility represents the upper plenum of such reactor with its internal components (control rods guides, support columns, ...) and its hot legs. The 450 l/s nominal flow rate is symmetrically distributed on 4 hot legs through regulation gates. One TSI PIV system including 200 mJ laser and high-resolution cross correlation camera (1K x 1K pixels) was used to describe the flow behavior. For practical reasons and to avoid image distortion due to circular optical access, the camera is located at the end of the hot leg, which is equipped with a flat glass window. Measurements were carried out in one hot leg and we will present in this paper data obtained in three different cross section planes (perpendicular to the main flow direction). The numerical simulation has been done using a standard CFD code that allows computing thermohydraulics phenomena into complex geometrical situation. It is solving 3D Navier Stokes equations using finite volume method and allows computing steady and unsteady conditions. In order to make a realistic comparison between experimental results and numerical simulation, we averaged the PIV data over 100 instantaneous velocities fields. The results show a good agreement between experimental data and CFD regarding the velocity spatial distribution and general flow behavior. However it is necessary to investigate with more detail in order to have a good agreement not only when comparing the general flow behavior but also comparing numerical data (velocity magnitude).

In this paper, we analyse all the sentences handed out by the Assize Court of Rennes in France, in the last decade. A recent increase in the number of sexual offences has been observed. An increase in the number of cases of incest has also been noted. This paper analyses the sex ratio, the age of the victim, the association with other violence and the relationship between the victim and his or her aggressor. The outcomes of the trials are discussed.

L’augmentation des admissions aux Urgences dépassant quotidiennement les capacités de soins a conduit les services à s’organiser afin de déterminer quels patients devaient être pris en charge rapidement et lesquels pouvaient attendre. Ce processus, dénommé « triage », consiste à déterminer à la phase initiale de la prise en charge d’un patient, la filière adaptée à son état en termes de délai et de type de soins. Cette mission est confiée en France, le plus souvent, aux infirmières d’accueil et d’orientation avec, si possible, un appel à un médecin référant dans les situations complexes et/ou les périodes de débordement. Alors qu’un tri en « très-urgent, urgent ou non urgent » s’est avéré décevant, la fiabilité de plusieurs échelles de triage en cinq priorités de prise en charge médicale a été confirmée, par une concordance entre infirmières et médecins, par une corrélation avec le recours à l’hospitalisation et par une bonne reproductibilité. Ces échelles s’appuient sur une classification des motifs de recours et, éventuellement, l’estimation des paramètres vitaux et de la douleur. L’utilité réelle du triage dans un service d’Urgences dépend des capacités des équipes à s’approprier l’échelle, à intégrer la priorité dans l’organisation des filières de soins et à mettre en place des procédures d’évaluation.As a result of increasing numbers of patients presenting to emergency departments (EDs), EDs have attempted to identify patients who need to be seen on a priority basis and patients who can wait safely. The aim of this process, named “triage”, is to assign at each patient the best channel in terms of timing and place. In France, triage is performed most-of-time by frontline nurses who can be helped by a physician for the more difficult situations or when ED is overcrowded. A simple classification “emergent, urgent, non urgent ” is unreliable but several 5-level triage scales are validated with high agreement between physicians and nurses, ability to detect admission and inter-rater & intra-rater agreements. These scales used classification of the presenting complaints, vital signs and pain. Triage usefulness depends on the team’s abilities to adopt scale, to adapt ED organization and to evaluate practice.

After a brief summary of the work performed within the frame of the NET Shielding Blanket segment predesign and manufacturing feasibility study, which first phase has been recently completed, this paper gives the outlines of a first development phase to come. This work was contracted by the NET Team to the CEA with Framatome and NFM as subcontractors.

ABSTRACT

This paper presents a preliminary analysis aiming to guide the choice of the power capacity of the future French Sodium Fast Reactor (SFR) prototype. Given the French background on SFRs, the reactor prototype, called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration), will be a demonstration reactor to validate at an industrial scale, the future commercial reactor technologies (1) mainly in terms of architecture and components design, core features, safety approach, operating specificities and performances. Thus, the ASTRID prototype is currently considered as the last step before a first of kind SFR power plant and it is admitted that this prototype power will be significant, in a range between 250 and 600 eMW. The ASTRID prototype should be operational by 2020 and will be designed to allow irradiation experiments and to demonstrate the SFR capability for minor actinides transmutation, as requested in the framework of the French law of June 2006 on nuclear waste management. The first part of this analysis aims at highlighting the constraints related to the prototype operation and estimating their impact on the choice of the power level. Different environmental constraints are considered such as for example: regulatory requirements, fuel cycle plants capacity, SFR prototypes international context, constraints related to the industrial facilities availability and the nuclear plant site. In a second part, a list of functions is defined, and for each of them an analysis of the power level impact is proposed, based on quantitative evaluations when possible. The main functions to be ensured by the SFR prototype concern the industrial demonstration of commercial reactor technologies for the primary circuit (components, fuel-handling system), secondary circuit, core design, ISI&R strategy, in-situ instrumentation, safety approach, irradiation tests and Minor Actinides transmutation capabilities. Functional analysis is carried out assuming that the SFR prototype is aimed to be representative to a 1500 eMW SFR commercial pool type reactor, and its core must reflect the main design options of the current oxide SFR reference core design (2). In the purpose of industrial demonstration regards to a large size commercial reactor (1500 MWe), a first analysis shows that, the higher is the SFR prototype power capacity, the more representative is the prototype according to the commercial SFR features. This trend is to be confirmed concerning in particular industrial demonstration of core features, handling systems, steam generators technology, and also economic aspects. Finally, recommendations will be given about the power capacity of the ASTRID SFR prototype. References: