12.6.2.1 BREST-300. Lead cooled reactor systems are under study at the Institute of Physics and Power Engineering (IPPE) and the Kurchatov Institute (IAEA-TEC- DOC-1289, 2002).

In the BREST-300 designs, developed by RDIPE and Kurchatov there is a two circuit design, there are four parallel loops including pumped lead flow removing heat from the reactor core. Lead inlet and outlet temperatures are 420 and 540°C, respectively. The design is integral with a supercritical pressure (24.5 MPa) steam water cycle. The uranium and plutonium nitride fuel implies low moderation and absorption of neutrons hence it is possible to achieve a core breeding ratio equal to one.

The BREST-300 reactor has various safety features such as negative void temperature coefficient; it operates with a breeding ratio of near unity with consequently minimal excess reactivity and there are no soluble poisons in the reactor coolant (IEA/OECD (NEA)/IAEA, 2002). Regarding the coolant, there is decay heat removal by passive systems, increased reactor coolant inertia, and the system pressure is low.

It has good thermodynamic efficiency due to high core outlet temperature, reduced number of components in the nuclear steam plant and reduced containment design requirements. The relatively small size implies reduced capital cost and this together with increased core outlet temperature means that the plant is also applicable to process heat applications.

There are, however, some penalties in using lead. There is a much greater pressure drop (about 7 times greater than sodium) across the core for otherwise similar conditions of reactor power, coolant flow cross section area in the core and fuel element length. This is caused by the lower thermal capacity of lead compared with sodium. The higher density of lead compared with sodium does not compensate. Lead cooled reactors therefore need to have a reduction in fuel fraction and increase in core diameter to reduce the hydraulic resistance. This implies that the core dimensions of the BREST reactors are large.

12.6.2.2 BREST-600. The plant has been scaled up to 600 MWe by RDIPE in co-operation with RRC Kurchatov (IAEA-TECDOC-1289, 2002). The characteristics of BREST-300 and BREST-600 are similar.

There is also an active programme on lead cooled reactors in Japan.

12.6.2.3 LCFR. Design studies of lead cooled fast reactors (LFRs) with nitride have been performed by the Japanese (IAEA-TECDOC-1289, 2002) as part of their programme to improve uranium resource utilisation and for the transmutation of high-level waste nuclides. The Japanese studied the impact of plant size on seismic issues and ways of developing more compacted and integrated plant designs.

The LCFR has negative void reactivity but a high breeding ratio of 1.26. The design is integral with the core, support structure and primary heat exchange systems situated within the reactor vessel. On the secondary side, the once through steam generator and its helical tubes are situated around the core and core diagrid. Regarding safety characteristics, the design is to reduce the propensity for lead-steam interaction.

12.6.2.4 General. Lead cooled systems have both advantages and disadvantages compared with sodium systems. Lead is much less reactive with air and water compared with sodium. In terms of other advantages in regard to minimum reactivity excess, transmutation of old actinides and fission products, proliferation safeguards consider­ations, safety in accident situations and economic competitiveness, lead and sodium systems have comparable properties.

There are however some negative aspects of lead associated with its corrosiveness; it may freeze in the steam generator in the case of feed-water heater failure. Repair and maintenance and remote re-fuelling operations are carried out at high lead temperatures of over 400°C. There is a potential for fuel subassembly blockage caused by lead/water/steam interactions.