The pre-conceptual design of the ASTRID project was launched in 2010 by CEA. The objectives of this first phase are to consider innovative options to improve the safety level with progress made in SFR-specific fields. A few examples of these innovations are: a core with an overall negative sodium void effect, specific features to prevent and mitigate severe accidents, power conversion system decreasing drastically the sodium-water reaction risk, improvements in In-Service Inspection and Repair, etc. ASTRID will also be designed to pursue the R&D on sodium fast reactors and demonstrate the feasibility of transmutation of minor actinides. The paper describes the current status of the project, the mains results obtained during the pre-conceptual design and address also the main R&D needs and results, focused on sodium technology. Main R&D tracks and dedicated technological platforms have been identified, particularly thanks to the European project ADRIANA, and some more recent up-dates are described in this paper.

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 Commissariat à L’Énergie Atomique et aux Énergies Alternatives (CEA) and the Japan Atomic Energy Agency (JAEA) intend to develop prototype or demonstration sodium-cooled fast reactors (SFRs) within two decades. The common final goal of their respective programs is SFR commercialization. The target of a commercial SFR for CEA and JAEA is basically consistent with Generation IV goals. Based on their industrial backgrounds and feedback from past and existing reactor experiences, CEA and JAEA have selected pool and loop configurations for the Advanced Sodium Technological Reactor for Industrial Demonstration (ASTRID) and the Japanese Sodium-Cooled Fast Reactor (JSFR), respectively. CEA and JAEA have cross-analyzed both pool and loop concepts (ASTRID and demonstration JSFR, respectively). The analysis results show that both concepts are technologically feasible and meet design goals. From the viewpoint of collaboration, the analysis identifies a wide range of collaborative areas in several fields: design principles (e.g., design target and design standard); development of components and systems; development of component parts, devices, or subsystems; design methods; simulation tools; etc.

This paper presents briefly the safety approach as well as the R&D program that is underway to support the deployment of future French Sodium-Cooled fast Reactors (SFRs): A) Safety objectives and principles for future reactors. The content of the first section reflects the works of AREVA, CEA, and EDF concerning the safety orientations for the future reactors. The availability of such orientations and requirements for the SFRs has to allow introducing and managing the process that will lead to the detailed definition of the safety approach, to the selection of the corresponding safety options, and to the identification and motivation of the supporting R&D. B) Strategy and roadmap in support of the R&D for future SFRs. This section describes the R&D program led jointly by CEA, EDF, and AREVA, which has been developed with the objectives to be able to preliminarily define, by 2012, the safety orientations for the future SFRs, and to deduce from them the characteristics of the ASTRID prototype.

Main objectives of reactor dosimetry are the determination of the neutron flux and fluence. In industrial power reactors, they are also used to follow the vessel and internal structures neutron induced damages. The knowledge of the associated uncertainties represents a significant stake for nuclear industry due to the high uncertainty value, 10% 15% (one standard deviation) commonly found for the calculated neutron flux (E > I MeV). Thus, the CENSPEX is engaged in the present study in order to analyze and then to reduce uncertainties associated to the reactor dosimetry interpretation process. The first part of this paper presents the main steps of data processing. Dosimeters activities are treated using nuclear data, neutron computation results and irradiation conditions. This treatment constitutes a complex process where many entries are more or less correlated. Thus, assessment of the output uncertainties requires the implementation of rather complex methods. In order to achieve uncertainty determination, a specific software tool has been developed to automate the process and to perform Monte Carlo uncertainty propagation and sensitivity analysis. Data uncertainties identification and quantification are performed in particular with regard to covariance. Then the stochastic uncertainty propagation methodology is described and carried out on representative cases. In complement, a Monte Carlo sensitivity study based on Sobol indexes is achieved on these cases to find the most penalizing input uncertainties. A fine analysis is performed to point out whether the uncertainty is induced by the incriminate input data uncertainties or by the methodology process. The paper concludes by pointing out the needs in input data knowledge.

Probabilistic fracture mechanics is a major element to evaluate the risk of failure of the reactor pressure vessel when it is subjected to pressurized-thermal-shock transient. One of the key issues concerning this probabilistic evaluation is the validity of the statistical model for the fracture toughness of the vessel steel. The Master Curve methodology has been developed by Wallin [1] to analyse the scatter in the results of steel toughness measures. This scatter is treated using the weakest-link theory applied to a three-parameter Weibull distribution as follows: where P f if the failure probability, K IC the fracture toughness, K min the location parameter, K 0 — K min the scale parameter and α the shape parameter. Wallin has determined that and α = 4 for reactor vessel steels. Thus only K 0 has to be determined for a specific material using a maximum likelihood estimator. However due to the heterogeneity of the existing K IC databases, with different materials, different sizes of specimen and different expressions of toughness, the characterisation of the scatter in the toughness databases by a Weibull model with a unique slope α= 4 does not seem sufficient. In addition, the toughness is very dependent on material temperature and for the relative high temperatures, the assumption on the location parameter, seems too conservative. For these reasons, it was decided to develop a purely statistical model without physical considerations except the use of a Weibull distribution. In this paper we present three estimation techniques for fitting a 3-parameter Weibull distribution on a toughness database. We have also tested various sampling techniques in the field of the temperatures. For this purpose, the software WOLF3 (Weibull Or Lognormal distribution Fitting with 3-parameters) has been developed under the MATLAB data-processing system.

That paper describes thf first,works on the definition of a coherent response surface methodology for uncertainty, sensitivity and reliability applications. In case of non linear or discontinuous phenomenon, with a time consuming software as modeling, artificial neural networks have been tested as response surface. A Bootstrap method is proposed as art efficient validation tool for any kind of response surface.

The paper describes a reliability-based method for the optimization of partial safety coefficients defined and used in design codes. The purpose of the optimization problem is to determine the partial safety coefficients which minimize an objective function for sets of structures and loading situations covered by a design rule. This objective function is a sum of distances between the reliability of the structures designed by making use of safety coefficients and the target reliability. This optimization method is compared with the design point method where the choice of the safety coefficients is based an the coordinates of the most probable failure point. The example of a flawed pipe is used to illustrate this comparison.

The Reliability/Availability Assessments of the new technology components and structures have a real problem to find out relevant input data. The new technology components are very often designed to operate in such configurations which are beyond those of to-day operating experience. Failure data issued from the existing operating experience are not in many cases relevant. One way to generate lacked data may be the Monte-Carlo Simulation (MCS) approach in order to simulate the operating experience in new technology conditions. Simulation will be based art the elementary laws governing the mechanical behaviour of the components and structures under assessment. The computed failure data need to be validated.

A method to estimate the pressure drop coefficient on a nuclear power plant has been developed to help the surveillance of the hydraulics in the primary circuit. A feature of this method is to combine experimental measurements with software results to compute the pressure drop coefficient. This paper describes a method to assess this pressure drop coefficient uncertainty, that uses the likelihood function.

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: