Page 32 - Thermal Hydraulics Aspects of Liquid Metal Cooled Nuclear Reactors

Introduction to liquid metal cooled reactors 7

1.5.1.1 Introduction

The French act on the sustainable management of radioactive materials and waste,
adopted in 2006, mandates CEA to perform R&D on the reprocessing and trans-
mutation of spent fuel. This includes the commissioning of a Generation IV reac-
tor. Accordingly, CEA launched in 2010, with French and international industrial
partners, the conceptual design of a Generation IV sodium-cooled fast reactor
(SFR): the Advanced Sodium Technology Reactor for Industrial Demonstration
or ASTRID. The goals of ASTRID will be to demonstrate the multirecycling
and transmutation capabilities of the uranium-plutonium cycle on an industrial
scale and to demonstrate the feasibility and operability of SFRs for commercial
power production.

1.5.1.2 Description of the primary cooling system
In the preconceptual design phase, several major design options were selected.
This included the choice of a pool-type primary circuit with a conical inner vessel
(“redan”) designed to allow for extensive in-service inspection and repair access.
In terms of the reactor block, it was decided to use three primary pumps together
with four intermediate heat exchangers. Each intermediate heat exchanger is asso-
ciated with a secondary sodium loop that includes a chemical volume control sys-
tem and modular sodium-gas heat exchangers to nitrogen, Brayton-cycle power
conversion system. The choice of this nitrogen system eliminates the possibility
of sodium-water reactions at the steam generators. Regarding the core, a low-void
effect core design was developed. This design allows longer cycles and fuel res-
idence times and complies with all the control rod withdrawal criteria while
increasing safety margins for all unprotected loss-of-flow (ULOF) transients and
improving the general design. In this core design, the reactivity effect associated
with a total loss of primary coolant is negative, so that boiling in the core leads to a
power decrease.

1.5.1.3 Description of the safety concept

The ability to remove 100% of its long-term decay heat via passive means is one of the
key requirements of the ASTRID nuclear island design. To that end, the design
includes sodium decay heat removal loops capable of removing heat from the primary
circuit to passive sodium-air heat exchangers via natural convection. Together with
natural convection in the primary circuit itself, these loops give ASTRID fully passive
decay heat removal capabilities.
To provide defense in depth against scenarios such as the melting of the core, the
ASTRID reactor will be equipped with a core catcher. As other safety-related com-
ponents, this core catcher must be inspectable. The containment will be designed
to resist the release of the mechanical energy caused by a hypothetical core accident
or large sodium fires, to make sure that no countermeasures are necessary outside the
site in the event of an accident.