Fast-neutron nuclear reactor cooled by molten lead
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The concept is generally similar tosodium-cooled fast reactors, and most liquid-metal fast reactors have used sodium instead of lead. Few lead-cooled reactors have been constructed, except for theSoviet submarine K-27 and the seven SovietAlfa-class submarines (though these wereberyllium-moderated intermediate energy reactors rather than fast reactors).[1] Some proposed new nuclear reactor designs are lead-cooled.
The lead-cooled reactor design has been proposed as ageneration IV reactor. Plans for future implementation of this type of reactor include modular arrangements rated at 300 to 400 MWe, and a large monolithic plant rated at 1,200 MWe.
Lead-cooled fast reactors operate withfast neutrons and moltenlead orlead-bismuth eutecticcoolant. Molten lead or lead-bismuth eutectic can be used as the primary coolant because especially lead, and to a lesser degree bismuth, have lowneutron absorption and relatively lowmelting points.Neutrons are slowed less by interaction with these heavy nuclei (thus not beingneutron moderators), and therefore help make this type of reactor afast-neutron reactor. If a neutron hits a particle with a similar mass (such as hydrogen in apressurized water reactor (PWR), it tends to losekinetic energy. If it hits a much heavier atom such as lead, the neutron will "bounce off" without losing this energy. The coolant serves as aneutron reflector, returning some escaping neutrons to the core.
Smaller capacity lead-cooled fast reactors (such asSSTAR) can be cooled by naturalconvection, while larger designs (such as ELSY[3]) use forced circulation in normal power operation, but will employ natural circulation emergency cooling. No operator intervention is required, nor pumping of any kind to cool the residual heat of the reactor after shutdown. The reactor outlet coolant temperature is typically in the range of 500 to 600 °C, possibly ranging over 800 °C with advanced materials for later designs. Temperatures higher than 800 °C are theoretically high enough to support thermochemicalproduction of hydrogen through thesulfur-iodine cycle, although this has not been demonstrated.
The concept is generally very similar tosodium-cooled fast reactors, and most liquid-metal fast reactors have used sodium instead of lead. Few lead-cooled reactors have been constructed, except for some Sovietnuclear submarine reactors in the 1970s, but some proposed new nuclear reactor designs are lead-cooled, with one under construction.
Reactors that use lead or lead-bismuth eutectic can be designed in a large range of power ratings. The Soviet Union was able to operate theAlfa-class submarines with alead-bismuth cooled intermediate-spectrum reactor moderated withberyllium from the 1960s to 1998, which had approximately 30 MW of mechanical output for 155 MW thermal power (see below).
Other options include units featuring long-life, pre-manufactured cores, that do not require refueling for many years.
The lead-cooled fast reactor battery is a smallturnkey-type power plant using cassette cores running on a closed fuel cycle with 15 to 20 years' refuelling interval, or entirely replaceable reactor modules. It is designed forgeneration of electricity on small grids (and other resources, includinghydrogen production anddesalinisation process for the production ofpotable water).
The use of lead as a coolant has several advantages when compared to other methods for reactor cooling:
Molten lead does not significantly moderate neutrons. Moderation occurs when neutrons are slowed down by repeated collisions with a medium. When the neutron collides with atoms that are much heavier than itself, almost no energy is lost in the process. Thus, the neutrons are not slowed down by lead, which ensures that the neutrons keep their high energy. This is similar to other fast reactor concepts, including the molten liquid sodium designs.
Molten lead acts as areflector for neutrons. Neutrons escaping the core of the reactor are to some extent directed back into the core, which allows a betterneutron economy. This in turn enables more spacing between the fuel elements in the reactor, allowing betterheat removal by the lead coolant.[4]
Lead undergoes almost noactivation by neutrons. Thus, virtually no radioactive elements are created by absorption of neutrons by the lead. This is in contrast to thelead-bismuth eutectic which was used in other fast designs, including in Russian submarines. Thebismuth-209 in this mixture (which has a lowermelting point, 123.5 °C, than that of pure lead) isactivated to form210Po,polonium-210, which is a dangerousalpha emitter.
Although lead absorbs few neutrons, because of its highdensity (10.66 g/cm3, for liquid at itsmelting point), lead is very effective at absorbinggamma rays and other less penetratingionizing radiation (alpha andbeta particles). This ensures that radiation outside the reactor is kept at a low level.
In contrast to moltensodium metal, another metallic coolant used in fast reactors, lead does not have issues with flammability, and will solidify from a leak.[5]
The very wide temperature range at which lead remains liquid (up to 1400 K or 1126 °C) implies that any thermal excursions are absorbed without any pressure increase. In practice, the operational temperature will be kept at around 500 °C (932 °F) – 550 °C (1,022 °F), mainly because of the limitations of metallic alloys used in reactor structural components, such as their sensitivity tocorrosion by a liquid metal (liquid metal embrittlement) andamalgam-driven metaldissolution (continuousCr andNi extraction from thestainless steel).
As with all fast reactor designs, because of the high temperature and high thermal inertia,passive cooling is possible in emergency situations. Thus, no electrical pumping assistance should be required, natural convection being sufficient to remove residual heat after shutdown. To achieve this, reactor designs include dedicated passive heat removal systems that require no electrical power and no operator action.
All fast reactor designs operate at substantial higher temperatures in the core than water cooled (andmoderated) reactors. This allows a significantly higherthermodynamic efficiency in thesteam generators. Thus, a larger portion of thenuclear fission energy can be converted into electricity. More than 40% efficiency is achievable in real conditions, compared to around 30% in water-cooled reactors.
Similarly, as with all fast spectrum reactors, the coolant is not pressurized. This means that nopressure vessel is required, and the piping and ducts can be constructed with less pressure-resistant steel and alloys.[6] Any leak in the primary coolant circuit will not be ejected at very high pressure.
Lead has a highthermal conductivity (35 W/m・K) compared to that of water (0.58 W/m・K), which means thatheat transport from the fuel elements to the coolant is efficient.
Instead of regularrefueling, the whole core can be replaced after many years of operation. Such a reactor is suitable for countries that do not plan to build their own nuclear infrastructure.
Lead and lead-bismuth are verydense, increasing the weight of the system and therefore requiring morestructural support andseismic protection, which increases building cost.
While lead is cheap and abundant,bismuth is expensive and quite rare. Depending on its size, a grid-connected lead-bismuth reactor requires hundreds or thousands of tonnes of lead-bismuth.
Solidification of lead-bismuth solution blocks the coolant circulation and immobilizes parts of the reactor control systems, rendering the reactor inoperable. However,lead-bismuth eutectic (LBE) has a comparatively lower melting temperature of 123.5 °C (254.3 °F), making melting a less difficult task. Lead has a higher melting point of 327.5 °С, but is often used as apool type reactor where the large bulk of lead does not quickly freeze.
By leaking and solidifying, the coolant may damage equipment (seeSoviet submarine K-64) if no measures are taken to contain such leaks.
Lead-bismuth producespolonium-210 fromneutron activation ofbismuth-209. This radioactive nuclide will dissolve in the lead-bismuth and is analpha emitter with ahalf-life of 138.38 days. This can seriously complicate maintenance and pose a severe plant alpha-contamination risk. Thealpha particle emitted by210 Po has a high energy (~ 5.4 MeV). It is, therefore, highly radiotoxic in case of internal contamination of the body (inhalation andingestion risks) because of its high ionization density severely damaging thecells affected in the contaminatedtissues.
The most challenging problems of liquid lead and LBE are the possible damage caused byerosion andcorrosion to the fuel elements and reactor internals.[7][8][9] Surface erosion is aggravated by the highdensity and associatedkinetic energy of the liquid metal circulating at elevated speed in the reactor, especially if it would become contaminated byabrasive solid particles (oxides detached from metallic surfaces) or metallic debris. Corrosion is fuelled by thedissolution of metals present inalloys (e.g.,Ni,Cr, released fromstainless steel) in the liquid metal coolant (formation of liquidamalgam with precipitation of the dissolved metals at cold points) and theliquid metal embrittlement (LME) of thefuel claddings and structural materials of the reactor internals. To mitigate the corrosion problem, it is necessary to form a very thin and as dense as possible oxide filmpassivating the metallic surface. This could be achieved by accurately controlling the dissolvedoxygen/metallic oxides in the metallic coolant. Insufficient oxygen level would expose the bare metallic surface to severe corrosion problems. At the same time, anO2 excess would generate thickporous oxide films prone to be detached from the metal surface and aggravating erosion and blocking problems. The corrosion rate also increases with temperature. Newly developed alloys, such asalumina formingaustenitic steels (containing Al added aspassivating agent), which maintain a protective oxide layer onto the surface of the metallic reactor components, are also being studied as candidate materials to attempt to mitigate corrosion problems.
The high density of lead and LBE means that the fuel elements, control rods and mobile reactor components are all floating in the metallic coolant, complicating the engineering systems needed to handle them and avoid floating debris.
Metallic coolants (Pb, LBE, Na) areopaque to visible light, seriously complicating the refueling and maintenance operations of the reactor, therefore requiring special systems to handle the fuel elements and the control rods safely. However, the design and operational experience already existing forsodium-cooled fast breeders could also be applied to lead-cooled fast reactors.
Lead has a positivevoid coefficient, or void reactivity, meaning that as voids occur in the circulating coolant an increase in fertile fission and a decrease in the capture rate of all heavy nuclides results in reactivity increases as the void content inside the reactor increases, potentially leading to a positive feedback loop unless controlled by automatic mechanisms.[10]
TheMYRRHA project (forMulti-purpose hYbrid Research Reactor for High-tech Applications) is aimed to contribute to design a future nuclear reactor coupled to aproton accelerator (so-calledAccelerator-driven system, ADS). This could be a 'lead-bismuth-cooled,[11] or a lead-cooled, fast reactor' with two possible configurations: sub-critical or critical. It could be apool-, or a loop-type, reactor.
The project is managed bySCK CEN, the Belgian research center for nuclear energy. It is based on a first small prototype research demonstrator, the Guinevere system, derived from the zero-power reactor Venus existing at SCK CEN since the beginning of the 1960s and modified to host a bath of moltenlead-bismuth eutectic (LBE) coupled to a smallproton accelerator.[12][13] In December 2010, MYRRHA was listed by theEuropean Commission[14] as one of 50 projects for maintaining European leadership in nuclear research in the next 20 years. In 2013, the project entered a further development phase when a contract for the front-end engineering design was awarded to a consortium led byAreva.[15][16]
Aiming at a compact core with high power density (i.e. with a highneutron flux) to be able to operate as amaterials testing reactor, the fuel to be used in the ADS MYRRHA must be highly enriched in afissile isotope. A highly enrichedMOx fuel with30 – 35 wt. % of239 Pu was first selected to obtain the desired neutronic performances.[17][18][19] However, according to Abderrahimet al. (2005)[18] "this choice should still be checked against the non-proliferation requirements imposed to newtest reactors by the RERTR (Reduced Enrichment of fuel for Research Testing Reactors) program launched byUS DOE in 1996". So, the fuel to be selected for MYRRHA also needs to respect the criteria of non-proliferation while keeping its neutronic performance. Moreover, such a highly enriched MOx fuel has never been industrially produced and poses severe technical and safety challenges in order to prevent anycriticality accident during handling in the factory.
In 2009, under the auspices of theNuclear Energy Agency (NEA,OECD), an international team of experts (MYRRHA International Review Team, MIRT) examined the MYRRHA project and delivered prudent recommendations to theBelgian government.[20] Beside the technical challenges identified, they were also financial and economical risks related to the construction and exploitation costs expected to strongly increase when the project should enter a more detailed design stage. Long construction delays related to design complications, underestimated technical difficulties and insufficient budget are not uncommon for such a project. The limited participation of theBelgian State (40% of all the costs) and the uncertain benefits for the external project owners were also pointed out.[20]
Because of recurrent financial shortcomings and also important uncertainties still subsisting in the reactor design (pool-, or loop-type, reactor?) and the choice still to be made for the liquid metal coolant (inLBE,209 Bi isneutron activated producing the highlyradiotoxic ⍺-emitting210 Po)[21] thefront-end engineering design (FEED) activities[22] had to be suspended and have not progressed beyond the preliminary stage.[23] Quite surprisingly, the preliminary results of the FEED activities were published in a journal absolutely not related to the field of ADS or fast neutron reactor: theInternational Journal of Hydrogen Energy (IJHE) while there was never any question of producinghydrogen with MYRRHA.[citation needed] The choice of this journal to present the preliminary results of the FEED activities is disconcerting. The journal where the FEED activities were announced,Physics Procedia, is also discontinued.[citation needed] Beside continuously increasing costs and financial uncertainties, the project still has to address many technical challenges: severe corrosion issues[7][8][9] (liquid metal embrittlement,amalgam-drivendissolution in the molten metal ofCr andNi from thestainless steel used for thefuel claddings and reactor structure materials), operating temperature (metal solidification risks versus increased corrosion rate),nuclear criticality safety issues...
The mass inventory of the lead-bismuth eutectic (LBE) for the proposed pool-type design of MYRRHA considered in the preliminary FEED analyses of 2013-2015 represents 4500 tons metallic Pb-Bi.[22] This would lead to the production of more than 4 kg of210 Po during the reactor operations. After the first operating cycle, 350 g of210 Po would already be formed in the LBE exposed to a highneutron flux of the order of 1015 neutrons・cm–2・s–1, typical for amaterials testing reactor (MTR).[24] This would correspond to anactivity of 5.5 × 1016becquerels,[24] or 1.49 × 106curies of210 Po, just for the first operation cycle. The presence of such a large ponderable quantity of highly radiotoxic210 Po represents a considerable radiological safety challenge for the maintenance operations and the storage of the MYRRHA nuclear fuel. Because of the high volatility of210 Po, the plenum space above the reactor could also become alpha-contaminated. As pointed out by Fioritoet al. (2018): "Some polonium will migrate to the cover gas in the reactor plenum and will diffuse outside the primary system when the reactor is opened for refueling or maintenance". All operations in210 Po contaminated areas will require appropriate radiological protection measures much more severe than for the239 Pu handling, or to be completely performed by remotely-operated robots. An envisaged mitigation strategy[24] could consist into a continuous removal of polonium from LBE, but the considerable heat generated by210 Po represents a major obstacle.[24]
In 2023, based on interviews with key SCK CEN players and documents publicly available, Hein Brookhuis explored the interactions between the MYRRHA promoters and the Belgian media and political spheres to show how MYRRHA was developed in a narrative that made the project seems essential to the future of SCK CEN, the Belgian nuclear research center.[25]
Paris based companynewcleo is developing 30 MWe and 200 MWe lead-cooledsmall modular reactors for naval and land use. The first operational reactor is planned to be deployed in 2031 in France. By 2026 an electrical simulator of a liquid-lead cooled reactor named PRECURSOR is to be built byENEA and newcleo at the ENEA Research Center in Brasimone (Bologna, Italy).[26][27][28]
Thedual fluid reactor (DFR) project was initially developed by a German research institute, the Institute for Solid-State Nuclear Physics, in Berlin. In February 2021, the project was transferred to a newly founded Canadian company, Dual Fluid Energy Inc., to industrialize the concept. The DFR project attempts to combine the advantages of themolten salt reactor with these of theliquid metal cooled reactor.[29] As a fast breeder reactor, the proposed DFR reactor is designed to burn both naturaluranium orthorium, as well astransmutating andfissioning minoractinides. Due to the highthermal conductivity of the molten metal, the residual decay heat of a DFR reactor could be passively removed.
ALFRED (Advanced Lead Fast Reactor European Demonstrator) is a lead cooled fast reactor demonstrator designed byAnsaldo Energia from Italy, planned to be built inMioveni, Romania. ATHENA, a molten lead pool used for research purposes, is going to be built in the same site as well.[30]
Two types of lead-cooled reactor were used inSovietAlfa-class submarines of the 1970s. TheOK-550 andBM-40A designs were both capable of producing 155MWt. They were significantly lighter than typical water-cooled reactors and had an advantage of being capable to quickly switch between maximum power and minimum noise operation modes.[citation needed]. These included a beryllium moderator and were therefore not fast-neutron reactors, but rather intermediate-neutron reactors.[1]
Construction ofBREST-OD-300 has started on 8 June 2021.[31][32] It is expected to start operation in 2026.[33]
The companyBlykalla is in collaboration withKTH Royal Institute of Technology andUniper[34] developing theSEALER-55 (Swedish Advanced Lead Reactor) reactor, a 55 MW lead-cooled mass-produced reactor using uranium nitride as fuel.[35][36] TheGovernment of Sweden committed 720 millionSwedish krona and started building a test facility in early 2025 for a lead-cooled prototype reactor.[37] The reactor, called SEALER-E, is planned to be built by 2026 in collaboration withABB.[38] The first commercial nuclear reactor (SEALER-One) is planned to be built in Oskarshamn in with the hope of reaching criticality in 2029.[35][39] Serial production of the SEALER-55 is planned to start in the early 2030s.[35]
The initial design of theHyperion Power Module was to be of this type, usinguranium nitride fuel encased in HT-9 tubes, using a quartz reflector, and lead-bismuth eutectic as coolant. The firm went out of business in 2018.
^abAbderrahim, H. A.; Sobolev, V.; Malambu, E. (October 2005).Fuel design for the experimental ADS MYRRHA. Technical Meeting on Use of LEU in ADS. 10–12 October 2005. Vienna, Austria: IAEA. pp. 1–13.
^Eckerman, K.; Harrison, J.; Menzel, H-G.; Clement, C.H.; Clement, CH (January 2012). "ICRP Publication 119: Compendium of Dose Coefficients based on ICRP Publication 60".Annals of the ICRP.41:1–130.doi:10.1016/j.icrp.2012.06.038 (inactive 1 July 2025).PMID23025851.S2CID41299926.{{cite journal}}: CS1 maint: DOI inactive as of July 2025 (link)
Tuček, Kamil; Carlsson, Johan; Wider, Hartmut (2006). "Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues".Nuclear Engineering and Design.236 (14–16):1589–1598.Bibcode:2006NuEnD.236.1589T.doi:10.1016/j.nucengdes.2006.04.019.
