Lead-cooled fast reactors designs

Lead cooled fast reactor design assumes the subassembly with fuel pins arranged in square lattice with large pitch-to-diameter ratio (s/d 1.4-1.5). In LMFR a more tight arrangement of fuel pins is adopted (s/d 1.1-1.18) ... 

Lead-cooled fast reactor NPP (Russian design BREST-OD-300 reactor coolant — liquid lead P 0.1 MPa and Pin/Pout = 420/540°C primary power cycle — indirect subcritical pressure Rankine steam cycle Pi 17 MPa (P = 22.064 MPa) and Pin/Pout — 340/505°C (Por = 374°C) high-temperature steam superheat (in one of the previous designs of BREST-300 NPP primary power cycle was indirect supercritical-pressure Rankine steam cycle Pjn 24.5 MPa (P r = 22.064 MPa) and Pin/Pout = 340/520°C (Per = 374°C) also, note that power-conversion cycle in different lead-cooled fast reactor designs from other countries is based on a supercritical pressure carbon-dioxide Brayton gas turbine cycle. 41-43... 

Overview of Generation IV lead-cooled fast reactor designs... 

Preliminary design study of lead cooled fast reactor with nitride fuel assemblies has been performed by the Japanese specialists to improve uranium resource utilization and transmutation of HLW nuclides. Plant size limitations caused by seismic resistance... 

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). 

The lead-cooled fast reactor uses either lead or lead-bismuth eutectic in the primary coolant loop. This gives similar advantages as with the SFR in terms of operational safety. Several of these reactor designs were built and operated on Russian submarines. 

An agreement to submit design descriptions for this report was not reached with the designers of BREST-300 lead cooled fast reactor from RDIPE (NIKIET) of the Russian Federation and the designers of CANDU X NC reactor from AECL of Canada (the latter is a Generation IV system with supercritical light water coolant). A description of the BREST-300 can be found in reference [21]. 

Lead and LBE are relatively inert liquids with very good thermodynamic properties. The LFR would have multiple applications including production of electricity, hydrogen, and process heat. System concepts represented in plans of the GIF System Research Plan are based on the European Lead-cooled Fast Reactor, Russia s BREST-OD-300 (fast reactor with lead coolant BbicxpbiH PeaKTop co CBHmtoBbiM TeiiJiOHOCHTeJieM in Russian abbreviations) and the Small Secure Transportable Autonomous Reactor concept designed in the US. 

Lead-Cooled Fast Reactor (ELFR) and added a mid-size LFR (ie, the BREST-OD-300) as a new thrust and reference reactor system, while the SSTAR legacy system was retained as the reference small LFR. The typical design parameters of these GIF— LFR reference systems were previously summarized in Table 6.2 and are described further in the following subsections. 

The objective of the Lead-Cooled Fast Reactor—Amphora Shaped-200 MWg (LFR-AS-200) project is to design and build a simple and compact reactor intended to demonstrate the competitiveness of LFR reactors for apphcations in open, long-lived fuel cycles based on enriched U fuel, and applications in a closed fuel cycle with equilibrium composition of fuels. AS stands for amphora shaped, as the reactor is characterized by an amphora-shaped inner vessel and 200 is the rated electrical power of the reactor in MW. 

BELLA a computer code written specifically for the purpose of safety-informed design of lead-cooled fast reactors... 

De Bruyn, D., Alemberti, A., Mansani, L., Grasso, G., Bandini, G., Artioli, C., BubeUs, E., Mueller, G., Wallenius, J., Orden, A., April 14—18, 2013. Main achievements of theFP7-LEADER collaborative project of the Eruopean Commission regarding the design of a lead-cooled fast reactor. In Proceedings of ICAPP 2013 Jeju Island, Korea. 

Tucek, K., Carlsson, J., Wider, H., 2006. Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues. Nuclear Engineering and Design 236, 1589—1598. 

In the field of fast reactor design and operational data, the last reference document published by the IAEA was the 1996 Fast Reactor Database (IAEA-TECDOC-866). Since its publication, quite a lot has happened the construction of two new reactors has been laimched, and conceptual/design studies were initiated for various fast reactors, e.g. the Japanese JSFR-1500 and the Russian BN-1800 (both cooled by sodium), as well as for a wholly new line of LMFR concepts — modular reactors cooled by sodium and by lead-bismuth alloy, and prototype and demonstration conunercial size fast reactors cooled by lead. 

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). 

The SVBR-75/100 (ANNEX XIX) is a modular multi-purpose lead-bismuth cooled fast reactor of 75 to lOOMW(e), offering a refuelling interval of 6 to 9 years. The design is backed by the experience of 50 years in design and operation of reactor installations with lead-bismuth coolant for nuclear submarines, available in the Russian Federation. Specifically, the marine prototypes of the SVBR-75/100 have achieved a total of 80-years of operating experience. 

XXVII-2] Buongiomo, J., Todreas, N.E., Kazimi, M.S., et al. Conceptual design of a Lead-Bismuth Cooled Fast Reactor with In-Vessel Direct-Contact Steam Generation , MIT-ANP-TR-078, Center for Advanced Nuclear Energy Systems, Massachusetts Institute of Technology, Boston, USA, 2001. 

Of the twenty-six concepts and designs addressed, 13 (50%) are water cooled SMRs, 6 (23%) are gas cooled SMRs-high temperature gas cooled reactors (HTGRs), 6 are sodium or lead-bismuth cooled fast reactors, and 1 is a non-conventional very high temperature reactor concept, a liquid salt cooled reactor with HTGR type prismatic fuel. 

Of the six liquid metal cooled SMRs, three are sodium cooled fast reactors (KALIMER, BMN-170 and MDP), and 3 are lead-bismuth cooled fast reactors (RBEC-M, PEACER-300/550, and Medium Scale Lead-bismuth Cooled Reactor). All designs implement indirect thermodynamic cycles. All sodium cooled SMRs incorporate intermediate heat transport systems (secondary sodium circuits to transport heat to a steam turbine circuit and to prevent the possibility of a contact of water with the primary sodium). All lead-bismuth cooled SMRs have no intermediate heat transport system. All designs use steam turbine power circuit. 

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. 

Pure lead and the eutectic alloy of LBE (consisting of 44.5% lead and 55.5% bismuth) are the principal potential coolants for LFR systems. Table 6.1 shows some key properties of LBE and lead with sodium also included for reference and comparison. Further details on the properties of lead coolants can be found in OECD-NEA (2015). The shared property that both LBE and lead are essentially inert in terms of interaction with air or water is the noteworthy advantage that LFRs have in comparison with the other principal liquid metal-cooled reactor, the sodium-cooled fast reactor (SFR). This basic property has significant implications for design simplification, safety performance, and the associated economic performance of such systems in comparison with SFRs and other Generation IV systems. 

Jun Lim Journal of Nuclear Materials Design of alumina forming FeCrAl steels for lead or lead—bismuth cooled fast reactors... 

Liquid metal-cooled systems include those for the SFR and the lead (or lead-bismuth)-cooled fast reactor (LFR). The SFR (see a possible design in Fig. 1.3) uses liquid... 

The natural properties of lead coolant and mononitride fuel and the neutronic characteristics of the fast reactor combined with the design of the core and cooling circuits raise the BREST reactor to a radically different level of safety and provide for its stable behavior without involving active safety features in the severe accidents unmanageable in any one of the existing reactors, such as ... 

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. 

XXV-7] ZAKI, S., SEKIMOTO, H., Accident Analysis of lead or lead-bismuth cooled small safe long-life fast reactor using metallic and nitride fuel. Nuclear Engineering and Design, 162, pp. 205-222 (1996). 

XXVI-4] SU UD, Z., SEKIMOTO, H., Design and safety aspect of lead and lead bismuth cooled long-life small-safe fast reactor for various core configuration. Journal of Nuclear Science and Technology 32/9 (1995). 

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