Structural materials sodium-cooled fast reactor

Fast Breeder Test Reactor (FBTR) is a 40 MWt/ 13.2 MWe sodium cooled, mixed carbide fuelled, loop type reactor. It has two primary and secondary sodium loops and a common steam water circuit, which supplies high pressure, high temperature superheated steam to turbine generator (TG). Heat is rejected in cooling tower (Fig 1). A 100% capacity dump condenser is provided for reactor operation even when the TG is not in service. The mmn aim of the reactor is to generate experience in the design, construction and operation of sodium cooled fast reactors and to serve as an irradiation facility for the development of fuels and structural material for fast reactors. It achieved first criticality in Oct 85 with Mark I core... 

The first ferritic oxide dispersion-strengthened (ODS) alloys were developed by SCK CEN, in the 1960s in the frame of the sodium-cooled fast breeders [1], Compared to austenitic alloys, ferritic/martensitic materials display a higher thermal conductivity, a lower thermal expansion, and a lower tendency to He-embrittlemenL They also exhibit a lower swelling as illustrated in Fig. 10.1 [2]. Creep properties of conventional ferritic/martensitic (F-M) alloys are not sufficient to withstand the levels of mechanical loading reached in some core structures. For example, at the end of life, for sodium-cooled fast reactors (SFRs), the internal pressure in the cladding tubes could reach... 

Figure 16.3 The vertical bars display the temperature interval between the onset of interstitial migration and the onset of vacancy migration in several pure metals and ceramics [7]. The horizontal bars show the approximate operating temperatures for structural materials in the cores of current light water-cooled reactor (LWR) power plants and proposed Generation IV sodium-cooled fast reactors (SFRs).

Ferritic steels show excellent resistance to irradiation. Therefore, within the range of irradiation on the structural materials of sodium-cooled fast reactors, normally, no major issues are expected. 

Coolant technology and structural materials developed for fast sodium cooled reactors. 

The CEFR is a sodium cooled, bottom supported 65 MW(th) experimental fast reactor fuelled with mixed uranium-plutonium oxide (the first core, however, will be loaded with uranium oxide fuel). Fuel cladding and reactor block structural materials are made of Cr-Ni austenitic stainless steel. It is a pool type reactor with two main pumps, and two loops for the primary and secondary circuit, respectively. The water-steam tertiary circuit has also two loops, with the superheated steam collected into one pipe that is connected with the turbine. CEFR s has a natural circuit decay heat removal system. 

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