Lead-cooled fast reactors coolant

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. 

It should be pointed out that presently only liquid metal coolant-sodium is widely adopted for fast reactors. Mercury was used for a short period ("Clementine" reactor in the USA and BR-2 in the USSR having, respectively, thermal output powers of 30 and 100 kW). A number of lead cooled fast reactors are being studied presently. 

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 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... 

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. 

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. 

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]. 

To achieve a long-life safe simple small portable proliferation-resistant reactor, a lead-bismuth-eutectic (LBE) coolant was selected as the best candidate. The original concept of a long-life small LBE cooled fast reactor was proposed more than 10 years ago [XXV-3], which was the world s first trial of this kind. The name of this reactor, the LBE cooled long-life safe simple small portable proliferation-resistant reactor (LSPR) distinguishes it from similar reactors proposed by other institutes. 

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 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 ensure a supply of inert gas (argon) in the coolant under the core. 

In Russia, two initiatives are currently being pursued. One of these is known as the SVBR (Svintsovo-Vismutovyi Bystryi Reaktor or Lead—Bismuth Fast Reactor ) (Zrodnikov et al., 2009). The SVBR-100 is generally considered a foUow-on technology to the prior submarine propulsion technology and is a small reactor cooled by LBE. The second major initiative, known as the BREST Bystry Reaktor so Svintsovym Teplonositelem or Fast Reactor with Lead Coolant ) (Dragunov et al., 2012), is a medium-sized reactor cooled by pure lead and detailed further in this chapter as one of the reference LFR reactor systems in the Generalion IV program (GIF-LFR-pSSC, 2014). 

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. 

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... 

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 ... 

XXni-5] SIENICKI, J.J., et al.. The STAR-LM lead-cooled closed fuel cycle fast reactor with a supercritical carbon dioxide Brayton cycle advanced power converter, Russian Forum for Science and Technology FAST NEUTRON REACTORS, HLMC-2003 (Paper presented at the 2 Int. Conf. on Heavy Liquid Metal Coolants in Nuclear Technologies, Obninsk, Russia, December 8-12, 2003), paper 2105. 

Low specific mass of lead bismuth coolant in small power fast reactors with the specific core power density several times lower than that in sodium cooled reactors is achieved through the elimination of in-vessel repository of spent nuclear fuel and in-vessel refuelling mechanisms (rotating plugs, etc). 

Among liquid metal candidates, mercury (Hg), sodium-potassium (NaK) alloy, sodium (Na), lead (Pb), and lead-bismuth eutectic (Pb-Bi) have been considered and used to build and operate liquid metal nuclear systems. However, liquid Na became the most smdied and used option mainly because it allowed, together with the selection of an appropriate fuel type (e.g., metal or oxide fuel), for a lower doubling time. On the other hand, hquid Hg was abandoned due to its toxicity, high vapor pressure and low boiling temperature as well as poor nuclear and heat transfer properties. More recently, in the framework of Generation IV, the development of fast reactors cooled with liquid metals considers hquid Na but also liquid Pb and liquid Pb-Bi as coolant... 

Liquid metal coolants for fast reactors cooled by sodium, lead and lead-bismuth eutectic, IAEA Nuclear Energy Series, No. NP-T-1.6, Vienna 2012. 

The possibility and expediency of developing the NP based on unified small power reactor modules SVBR-75/100 with fast neutron reactors cooled by lead-bismuth eutectic coolant (LBC) is substantiated for the nearest decades in the paper. 

Despite the fact that the proposed reactor technology is backed by a long design and operating experience of the lead-bismuth cooled reactors for nuclear submarines and the fast spectrum reactors with sodium coolant for NPPs it is still innovative for civil nuclear power. Therefore, additional validation and testing, as well as licensing would be required. 

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