BREST-OD-300 reactor

BREST-OD-300 reactor 1 - reactor core 2 - steam generator 3 - pump 4 - reloading machine 5 - reactor vault 6 - cooling system... 

Nuclear power complex with BREST-OD-300 reactor 1 - reactor 2 - turbine unit 3 -waste storage and treatment complex 4 - on-site nuclear fuel cycle complex... 

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... 

The BREST-OD-300 reactor (and its companion larger system design, the BREST-1200) is a system developed by the Russian organization NIKIET in association with a number of other organizations with the goal of realizing a naturally safe LFR concept. 

Figure 6.3 Sketch of the BREST-OD-300 reactor system. Dr. Andrei Moiseev, NIKIET.

Dragunov, Y., Lemekhov, V., Smirnov, V., Chemetsov, N., November 2012. Technical solutions and development stages for the BREST-OD-300 reactor unit. Atomic Energy 113 (1), 70. 

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). 

BREST-OD-300 (Russian Federation) Bystriy Reactor Estestveimoy Bezopasnosti (Fast Reactor Natural Safety)... 

Activities in the fast reactor area in Russian Federation include design studies of fast reactors with alternative coolants, including lead (BREST-OD-300 and BREST-1200). 

Two prototype land-based reactors utilizing LBE coolant were operated in Russia as prototypes for the LBE cooled submarine reactors. In the interest of reducing deployment times, full-size land-based prototype reactors were constructed. A progression of design, construction, and operation of LBE-cooled reactors of successively larger size was not followed. As a result, Russian submarines and their reactors became LBE reactor tests in the course of naval service. In light of that experience, the need for a demonstration test reactor of about 30 MW(e) power to support the development of the BREST-OD-300 300 MW(e) Pb cooled reactor has been identified in Russia [XXIII-14]. 

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. 

BREST-OD-300 is Fast Reactor with NATural safety test demonstration in Russian abbreviations (BPECT-OD-300 Biicrpiiift PeaKTOp c EdTecxBeHHOH 6e3onacHOCTLio — Oiibitho —J eMOHCTpaujHOHHtiH). 

The Russian BREST-OD-300/1200 concepts (discussed further in Section 6.6) feature a multizone concrete reactor vessel with the reactor core in the central zone, and reactor coolant pumps and SGs in separate zones to which the lead coolant flows through interconnecting channels. 

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 BREST-OD-300 is a pilot technology demonstration reactor being developed as a prototype of future commercial reactors of the BREST family, such as the larger BREST-1200. Table 6.6 provides a summary of key parameters for both the BREST-OD-300 and BREST-1200 concepts. 

The BREST-OD-300 is a reactor of pool-type design. It incorporates, within the pool, the reactor core with reflectors and control rods, the lead coolant circulation circuit with SGs and pumps, equipment for fuel reloading and management, and safety and auxiliary systems. These reactor systems and items of equipment are included in a steel-lined, thermally insulated concrete vault. 

Mononitride UN and UPuN fuel was initially tested in the BR-10 and BOR-60 reactors at 350—1045 W/cm and 4—9% bumup. Both showed good resistance to irradiation and low reaction rates with liquid metal coolants. They are also compatible with ferritic—martensitic steels, eg, EP-823 and EP-450 up to 800°C for 2000 h and 1200-1300°C for 5 h (Filin, 2000). In 2014-2015, hot cell examinations of UPuN fuel rods irradiated in BOR-60 and BN-600. Irradiation experiments on UPuN FAs for BREST and BN have started (Nikitin, 2015). The main design characteristics of the BREST-OD-300 are given in Table 12.7. 

Lessons learned from testing the BREST-OD-300 will be used to design the BREST-1200, expected to become a 1200 MWgi commercial Pb-cooled fast reactor. Conceptually, the BREST-OD-300 and BREST-1200 are similar. The main differences are... 

Bulkin, S.Yu, Lemekhov, V.V., Sila-Novitsky, A.G., et al., 2011. Research and development for demonstration of fuel performance in the BREST-OD-300 core. In IAEA Technical Meeting Design, Manufacture and In-pile Behaviour of Fast Reactor Fuel. Institute of Physics and Power Engineering, Obninsk, Russia. 

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