Johan Carlsson

ELSY—European LFR Activities

The European Lead Fast Reactor has been developed in the frame of the European lead system (ELSY) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involved a wide consortium of European organizations. The ELSY reference design is a 600MWe pool-type reactor cooled by pure lead. The project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goals. The paper focuses on the main aspects of the proposed design for the European lead fast reactor highlighting the innovation of this reactor concept and overall objectives. Special attention has been dedicated to safety starting from the first step of the design development taking into account other important aspects, such as the investment protection, the compactness of the primary system as well as sustainability. The main safety features of the proposed innovative decay heat removal (DHR) systems are presented. From the beginning of 2010, and for a duration of three years, the European Commission (EC) is financing the new project Lead European Advanced Demonstration Reactor (LEADER) as part of the 7th Framework Program. This paper highlights the main objectives of the LEADER project.

Emergency decay heat removal by reactor vessel auxiliary cooling system from an accelerator-driven system

Abstract The passive emergency decay heat removal during severe cooling accidents in Pb/Bi-cooled 80- and 250-MW(thermal) accelerator-driven system (ADS) designs was investigated with the computational fluid dynamics code STAR-CD. For the 80-MW(thermal) design, the calculations show that no structural problems occur as long as the accelerator proton beam is switched off immediately after accident initiation. A highly unlikely delay of beam stop by 30 min after a combined loss-of-heat-sink and loss-of-flow accident would lead to increased reactor vessel temperatures, which do not cause creep failure. By using a melt-rupture disk on the vacuum pipe of the accelerator proton beam to interrupt the beam at elevated temperatures in a passive manner, the grace time before beam stop is necessary is increased from 30 min to 6 h. An emergency decay heat removal design, which would prevent radioactive release to the atmosphere even more reliably than the Power Reactor Inherently Safe Module (PRISM) design, was also investigated. For an ADS of 250-MW(thermal) power with the same vessel as the 80-MW(thermal) ADS examined, the maximum wall temperature reaches 745 K after an immediate beam stop. This does not cause any structural problems either. The grace time until a beam stop becomes necessary for the 250-MW(thermal) system was found to be ~12 min. To reduce elevated vessel temperatures more rapidly after a beam stop, alternative cooling methods were investigated, for example, filling the gap between the reactor and the guard vessel with liquid metal and the simultaneous use of water spray cooling on the outside of the guard vessel. This decreases the coolant temperatures already within minutes after switching off the proton beam. The use of chimneys on the reactor vessel auxiliary cooling system, which increase the airflow rate lowers the maximum reactor vessel wall temperature only by ~20 K. It can be concluded that the critical parameter for the emergency cooling of an ADS is the time delay in switching off the accelerator after accident initiation.

Investigations of Alternative Steam Generator Location and Flatter Core Geometry for Lead-Cooled Fast Reactors

This paper concerns two independent safety investigations on critical and sub-critical heavy liquid metal cooled fast reactors using simple flow paths. The first investigation applies to locating the steam generators in the risers instead of the down-comers of a simple flow path designed sub-critical reactor of 600 MWth power. This was compared to a similar design, but with the steam generators located in the downcomers. The transients investigated were Total-Loss-of-Power and unprotected Loss-Of-Flow. It is shown that this reactor peaks at 1041 K after 29 hours during a Total-Loss-Of-Power accident. The difference between locating the steam generators in the risers and the downcomers is insignificant for this accident type. During an unprotected Loss-Of-Flow accident at full power, the core outlet temperature stabilizes at 1010 K, which is 337 K above nominal outlet temperature. The second investigation concerns a 1426 MWth critical reactor where the influence of the core height versus the core outlet temperature is studied during an unprotected Loss-Of-Flow and Total-Loss-Of-Power accident. A pancake type core geometry of 1.0 m height and 5.8 m diameter, is compared to a compact core of 2 m height and 4.5 m diameter. Moderators, like BeO and hydrides, and their influence on safety coefficients and burnup swings are also presented. Both cores incinerate transuranics from spent LWR fuel with minor actinde fraction of 5%. We show that LFRs can be designed both to breed and burn transuranics from LWRs. It is shown that the hydrides lead to the most favorable reactivity feedbacks, but the poorest reactivity swing. The computational fluid dynamics code STAR-CD was used for all thermal hydraulic calculations, and the MCNP and MCB for neutronics, and burn-up calculations.Copyright