There has been renewed interest in recent years in liquid metal cooled reactors particularly for smaller sized designs and from a sustainable development point of view. They are significant because they can breed new fissile material and extend the potential of nuclear energy. Because of their fast neutron spectrum, which can be used as a burner or a breeder, they have also received recent attention for incinerating weapons plutonium, thorium utilization, partitioning and transmutation of actinides and burning nuclear waste. First used in Russian submarines, liquid lead and lead — bismuth have received worldwide attention in the last few years for power reactors and also for accelerator driven transmutation systems. Russia, India, and Japan have remained most active in recent years in liquid metal power reactor development14. The Republic of Korea is developing a pool-type sodium-cooled 150 MWe KALIMER plant with metal fuel and a passive safety decay heat removal system.

India: India’s sodium-cooled Fast Breeder Test Reactor (FBTR), has been operating in Kalpakkam for several years. It has a unique mixed uranium carbide — plutonium carbide fuel. It was designed for 40 MWt but has only recently reached a power level of 17.4 MWt. It has achieved a fuel burnup of 90 GWd/t. Thorium blankets have been used in the breeder reactor in Kalpakkam. A 500 MWe sodium — cooled pool type Prototype Fast Breeder Reactor (PFBR) design is under development, also for the Kalpakkam site. It will use U-Pu MOX fuel. The Preliminary Safety Analysis Report for this reactor is nearing completion.

Japan: The two sodium-cooled fast reactors, the Experimental Fast Reactor “Joyo” and the prototype fast breeder reactor “Monju” are not operating at this time. Joyo will start operation in 2003 with a new high-flux core, and Monju is waiting for governmental approval for improvement work for sodium leaks, leading to its eventual startup in 3 more years. However, several small and medium size designs are being developed in Japan, the most prominent one being the 50 — 100 MWe sodium — cooled fast reactor design known as Super Safe, Small and Simple (4S)15. In this reactor, Burnup of the core is controlled by the annular reflector surrounding the core, and a long life is achieved by the long length of the core and upward movement of the reflector. The Modular Double Pool (MDP) is another concept of 325 MWe sodium — cooled fast reactor, which has steam generator and secondary pumps in the sodium filled annular space between the primary and the secondary vessel thereby reducing the secondary piping system. Metallic fuel is used for both of these two designs. MDP has been designed to reduce the construction cost and improve reliability by factory manufacture of most components, and 4S has been designed to obtain a long life core. A concept of Multipurpose Fast Reactor (MPFR) has also been proposed which has liquid plutonium-Uranium metallic fueled core. It has 300MW thermal power and does not require fuel reloading16.

A Pb-Bi cooled Long-life, Safe, Simple, Small, Portable, proliferation-resistant reactor (LSPR)17 has also been proposed. This is a 35 MWe (150 MWt) integral type design where the steam generators are installed within the reactor vessel. Nitride fuel is used. Natural or depleted Uranium fuel assemblies are placed at the center of the core and Pu fuel assemblies at the outside. In this composition, the burnup will progress from the outer core into the inner blanket region.

Russian Federation: Russia’s experience in the construction and operation of sodium-cooled experimental and prototype fast reactors (the BR-10, BOR-60, BN — 350 in Kazakhstan and BN-600 with hybrid core) has been very good. Efforts have been directed towards further improving safety and reliability, and making the Liquid Metal Fast Reactors (LMFRs) economically competitive to other energy sources. While these efforts would take some time, LMFRs are being considered to burn weapons plutonium and minor actinides. The current main efforts in sodium cooled fast reactors in Russia have been the lifetime extension for BOR-60 and BN-600, decommissioning of BR-10 and designing BN-800. By 2010, Russia wants to complete construction of the BN-800 fast reactor at Beloyarsk. Russia has also developed three small sodium-cooled reactor designs: MBRU-1.5, MBRU-12 and BMN-170 for production of 1.5, 12 and 170 MWe of electricity18.

The design from Russia that has received the most recent attention is the BREST reactor, which uses lead coolant, uranium-plutonium mono-nitride fuel and indirect cycle for heat removal to a supercritical steam turbine. Owing to unique combination of the thermo-physical properties of the lead coolant and mono-nitride fuel, BREST can boast of a very high level of natural safety. Two conceptual designs have been developed for the 300 MWe and 1200 MWe BREST reactors. Figure 5

gives the schematic details of the 300 MWe BREST design. Russian fast reactor R&D activities are concentrating on advanced concepts with enhanced safety features and designs with alternative coolants, as well as on the development of the basic design, and experimental confirmation, of the lead cooled BREST-300 demonstration reactor with on-site closed fuel cycle19.

Studies of small fast spectrum reactor modules cooled by lead-bismuth eutectic are also being pursued. These designs, called SVBR-75/100, are based on the reactor operation experience with nuclear submarines. The designs could be used for electricity production, seawater desalination, or the utilization and transmutation of actinides. The SVBR-75 is a Pb-Bi cooled 75 MWe (268 MWt) fast reactor with two — circuits, the primary Pb-Bi circuit and the steam-water secondary loop20. Two other heat removal systems are provided for both scheduled and emergency cooling. The reactor operates for 8 years without refueling. Average fuel enrichment is 15.6%.

USA: Although the U. S. had a strong sodium cooled reactor program for many years, it has essentially halted. Recently, however, because of impetus in research for new generation of reactors, one innovative liquid metal cooled design called the Encapsulated Nuclear Heat Source (ENHS) has been proposed21. The ENHS is a Pb — Bi natural circulation cooled, 50 MWe (125 MWt), modular, fast reactor concept. It is designed that the fuel is installed sealed into the reactor module at the factory and transported to the site to be inserted into a secondary pool of Pb-Bi that contains the steam generators. Major components, such as the pool vessel and steam generators, are permanent and remain at the site while the reactor module is replaced every 15 or 20 years. The heat generated in the core is transferred through the primary coolant vessel wall to the secondary pool. The natural circulation avoids the need for active components but it requires a tall 19m primary vessel. A design with a lift pump reduces the height to 10m and reduces the coolant mass. The fuel considered is metallic Pu-U-Zr fuel with 11-12% of Pu. The peak fuel Burnup is approximately 105,000 MWD/t. The autonomous control and no fuel handling reduce the nuclear operations onsite to a minimum. Figure 6 gives a schematic description of ENHS.

Water/Steam Connections

Primary Pb-Bi

Secondary Pb-Bi

Primary Vessel of ENHS Module

Fig. 7. Schematic diagram of a molten salt reactor such as the MSRE21

Heat Removal

Fig. 8. Schematic diagram of a molten salt cooled reactor such as MARS