Lead-cooled fast reactor development

Techniques to counter the heavy metal coolant disadvantages are being developed, but in spite of this work and the apparent disadvantages of sodium, the consensus in favour of sodium remains strong. This is demonstrated by fact that before lead-cooled fast reactor BREST-300 is built, MINATOM will first build a sodium-cooled LMFR BN-800 (E. Adamov, NW, 23 September 1999). Moreover, in the last few years sodium has been chosen in both China and the Republic of Korea for the respective fast reactor development project. This is a significant endorsement for sodium as a fast reactor coolant. 

Preliminary studies on lead-bismuth and lead cooled reactors and ADS (accelerator driven systems) have been initiated in France, Japan, the United States of America, Italy, and other countries. Considerable experience has been gained in the Russian Fedaration in the course of development and operation of reactors cooled with lead-bismuth eutectic, in particular, propulsion reactors. Studies on lead cooled fast reactors are also under way in this country. 

This made it possible for Russia to initiate development (Adamov and Orlov 1992) of a fast reactor named lead-cooled fast reactor of natural safety (BREST), i.e., a pilot 300 MWe reactor (for trying out, e.g., equilibrium operation) and the first-of-the-kind 1,200 MWe power unit, with the R D effort to support it, aimed at demonstration of the new fast reactor and its closed fuel cycle within the normal period of 20 years. 

An example of an innovative lead-cooled fast reactor of natural safety is the pilot 300 MWe reactor BREST-OD-300, which is being developed in Russia (ISTC 2001 Adamov et al. 1997). 

In the future, electricity production at large NPPs is likely to remain the main application of nuclear energy. This factor and the reduction of unit costs with increase in the power and number of nuclear units were the reasons for conceptual study of a 1,200 MWe lead-cooled fast reactor as a candidate basic component of a large-scale nuclear power mix. The BREST-OD-300 being developed as a prototype of the BREST-1200 reactor, their design and engineering features are largely similar, as may be seen from the data of Table 58.6 (ISTC 2001 Adamov et al. 1997). 

In addition to HTR and SFR, the other Generation IV concepts are also supported by different government agencies. The supercritical water-cooled reactor (SCWR) was supported under the National Key Basic Research Program of China (973 project) by the China Ministry of Science. The studies of molten salt reactors (MSRs) and lead-cooled fast reactors (LFRs) are performed in the framework of the Chinese Academy of Sciences (CAS) pilot projects. In the following section, the current research and development (R D) on Generation IV reactors in China will be briefly introduced. 

Lead-cooled fast reactor research and development... 

The lead-cooled fast reactor (LFR) system is also under development in Generation rV framework. It has to be pointed out that there is no industrial experience of lead aUoy-cooled technology except that fi om the Soviet Union submarine program. But many concepts exist worldwide for example, the MYRRHA (Multipurpose hYbrid Research Reactor for High-tech Applications) reactor is developed by the Belgian Nuclear Research Center SCK-CEN in collaboration with international partners. MYRRHA is conceived as an accelerator-driven system able to operate in subcritical and critical modes. It contains a proton accelerator of600 MeV, a spallation target, and... 

In the earlier phases of breeder reactor development, especially in the 1950s and 1960s, high pressure gases, such as helium,C02 or superheated steam were studied. Between 1960 and 1970, H2 0-steam cooled and D2 0-steam cooled fast reactor concepts were studied in the USA and the former FRG. Helium cooled fast reactor concepts have been pursued as an alternative coolant concept in Europe and the USA. Some fuel development for a CO2 cooled fast breeder has been continued on a small scale in the UK. Lead-bismuth alloy as a coolant was studied in the former USSR for propulsion and land based reactors. 

There are 11 concepts of small reactors without on-site refuelling that are currently being developed at different stages in the Russian Federation. Six of them are light water cooled reactors the UNITHERM (ANNEX II), the ELENA (ANNEX III), the VBER-150 (ANNEX IV), the ABV (ANNEX V), the KLT-20 (ANNEX VI), and the VKR-MT (ANNEX X). In addition, there is one small gas cooled fast reactor concept which is the BGR-300 (ANNEX Xm) two sodium cooled reactor concepts the MBRU-12 (ANNEX XVI), and the BN GT-300 (ANNEX XVItl) and one lead-bismuth cooled small reactor design, the SVBR-75/100 (ANNEX XIX). Finally, there is one non-conventional reactor concept the MARS (ANNEX XXVIII). 

At present, sodium is considered the best coolant for fast reactors due to its superior cooling ability, which can help to increase the core power density and shorten the doubling time. Short doubling time was an indispensable requirement in the early phases of development and construction of fast breeder reactors from 1960s through 1980s. It is reported that for safety reasons, the lead-bismuth eutectic (LBE) cooled fast reactor was originally considered [XXV-2]. 

Small lead-bismuth cooled fast reactors SVBR-75/100 discussed in this paper are based on actual experience in the development and operation of lead-bismuth cooled reactors for nuclear submarines [1]. In fifteen-twenty years from now it will be possible to deploy SVBR-75/100 in both industrialized and developing countries. These reactors make it possible to resolve a contradiction between economic characteristics and safety requirements that is peculiar to reactors of traditional type. Due to their improved technical and economical characteristics and higher safety level, fast reactors with lead-bismuth coolant could be considered as one of the possible candidates for step-by-step replacement of thermal reactors [2]. 

The RBEC-M is a lead-bismuth cooled fast reactor with a high level of primary coolant natural circulation and a gas lift system in the primary circuit to provide the supply of an inert gas (e.g. argon) in the coolant under the core, see Annex XXIII. This concept is developed with an insight of future multi-component nuclear energy systems, where it might be used for breeding or the adjustment of fissile material flows. Conceptual studies for the RBEC-M are performed in the Russian Research Centre Kurchatov Institute (Moscow, Russia). 

The main objective of the development of the RBEC lead-bismuth cooled fast reactor was to provide a reliable solution for nuclear fuel breeding, while using an approach alternative to sodium cooled fast reactors. It was assumed that design development of a nuclear power plant (NPP) with such reactor could be completed in a rather short period, with modest expenditures for additional testing and qualification of separate equipment units. 

Nuclear and magneto-hydrodynamic electric power generation systems have been produced on a scale which could lead to industrial production, but to-date technical problems, mainly connected with corrosion of the containing materials, has hampered full-scale development. In the case of nuclear power, the proposed fast reactor, which uses fast neutron fission in a small nuclear fuel element, by comparison with fuel rods in thermal neutron reactors, requires a more rapid heat removal than is possible by water cooling, and a liquid sodium-potassium alloy has been used in the development of a near-industrial generator. The fuel container is a vanadium sheath with a niobium outer cladding, since this has a low fast neutron capture cross-section and a low rate of corrosion by the liquid metal coolant. The liquid metal coolant is transported from the fuel to the turbine generating the electric power in stainless steel... 

There are three fast-flux reactors proposed for development the sodium cooled, the gas cooled, and the lead cooled. The fission cross sections for fast neutrons (high-energy spectrum neutrons) for all of the fissile actinides are nearly the same so the fast-flux reactors use all of the fissile actinides as fuel. The fast-flux isotopic fission cross sections are smaller than for thermal neutrons so the fraction of fissile isotopes (e.g., 235u 239pu, range of... 

GIDROPRESS, in which a great deal of experience has been accumulated in the course of the development and operation of submarine reactors cooled with lead-bismuth eutectic. However, bismuth is expensive and the resources are limited. It is possible that its use must be confined to special applications, such as small reactors or to a limited number of fast reactors. For this reason lead cooling is also being studied in the IPPE, Kurchatov Institute, and other organization [2.19-2.26]. 

It should be emphasized, that although designs and parameters of the early experimental fast reactors showed a wide variability, those of the commercial-sized plants are rather similar. Even with the initiation of a wholly new line of development, such as Pb and Pb-Bi cooled reactor designs, it is interesting to observe that their parameters are close to those of traditional reactors being advocated elsewhere. It is a further proof that the laws of physics and the principles of good engineering inevitably lead to similar optimal solution. 

XXII-2] SIENICKI, J.J., MOISSEYTSEV, A.V., SSTAR lead-cooled, small modular fast reactor for deployment at remote sites - system thermal hydraulic development, ICAPP 2005, Paper 5426 (Int. Conf on Advances in Nuclear Power Plants Seoul, May 15-19, 2005). 

---