Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the “Introduction” of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes

Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs

The Consequences of a Sharp Temperature Change in the Fuel Pins of an Accelerator-Driven Subcritical System

International audienceThe effect of temperature changes and in particularthose that are accompanied by strong gradients was extensivelyinvestigated for fast reactors. Subcritical systemsdesigned for their transmutation ability are to someextent similar to critical power reactors in their subassemblystructure. However, they differ in two mainaspects. First, the coolant in a subcritical system is leador lead-bismuth eutectic (LBE) and not sodium, and second,the main cause for steep temperature gradients in afast power reactor is sudden control rod insertion, orscram, whereas in subcritical systems shutdown of theaccelerator and its proton beam is the main cause fortemperature gradients. Furthermore, the increased probabilityof operational interruptions in an acceleratordrivensystem is largely due to the instability of theaccelerator generating the proton beam.This study uses the knowledge gained from fast reactorsas a preliminary reference and concentrates furtheron the unique features of the proposed subcriticalsystems.In particular, the effect of beam trips on the fuel pinintegrity is evaluated as a function of the temperaturegradients and the duration of the beam trips. It seems,however, that the largest hazard to the fuel pin integrityis due to the lead (or LBE) coolant. In particular, thestability of the protective oxide layer built on the cladsurface with the lead coolant appears quite sensitive tosudden temperature changes. In the second part of thisstudy, several available experimental results show thateven very moderate temperature changes are sufficient tocause crack formation in the oxide layer thereby exposingthe clad surface to enhanced LBE corrosion. In theworst case, complete exfoliation of the magnetite outerlayer is observed. As a consequence, clad failure probabilitydue to corrosion is considerably increased

Safety analysis results of the DBC transients performed for the ALFRED reactor

The LEADER project aims at the development to a conceptual level of a Lead Fast Reactor Industrial size plant and at a scaled demonstrator of the LFR technology - ALFRED. This paper presents the main safety transient analysis results using the system codes RELAP5, TRACE-FRED, SIM-LFR, CATHARE and SIMMER of the DBC (Design Basis Condition) transients for the ALFRED reactor. Apart from the traditional set of protected transients (PLOF, PTOP, LOOP), safety analysis was carried out for a number of carefully selected plant specific DBC transients, thus enveloping a wide spectrum of design basis conditions

EUROTRANS/ECATS or neutronic experiments for the validation of XT-ADS and EFIT monitoring

OECD (2011) NEA 6897 ISBN 978926411727