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| Fig. 1: Schematic diagram of a pool-type lead-cooled fast reactor (Source: Wikimedia Commons) |
Lead-cooled fast reactors (LFRs) are a class of nuclear reactor designs that utilize circulating molten lead (or Pb-Bi alloys) as the primary reactor coolant. As a liquid-metal cooled reactor, LFRs are conceptually similar to liquid Na-cooled fast reactors (SFRs) and offer many similar design advantages: including closed-fuel operation with fast neutron spectra, high power densities, and efficient high-temperature operation without coolant pressurization. [1] However, the unique physical and chemical attributes of lead coolants compared to sodium (namely its reduced chemical reactivity and much higher density), mean that LFR designs can differ from SFRs in significant ways, offering both unique advantages as well as design challenges. Historically, Pb fast reactors have been deployed for military purposes, primarily in Soviet nuclear submarines, however, in recent decades, they have seen renewed interest for use in commercial power generation. [1] Along with Na fast reactors, LFRs are one of the six advanced reactor types being pursued as Generation IV designs, with the goal of improved performance, economics, reliability, sustainability, and safety over existing water reactor designs. [2,3] This report will provide a brief overview of the operating principle, history, unique design attributes, and outlook of Pb-cooled fast reactors.
As discussed above, LFRs employ a molten lead alloy as their primary reactor coolant, which must be kept above its melting point in order to circulate as a liquid and carry heat away from the reactor core. [1] The principal coolants employed are either pure liquid lead (with a melting point of 327°C), or Pb-Bi eutectic (LBE) alloy (composed of 44.5% lead and 55.5% bismuth), which benefits from a lower melting point of 124.5°C that is more easily maintained within the reactor plumbing. [1] Because both lead and LBE have very high boiling points (1737°C and 1670°C respectively) and low vapor pressure (~10-5 Pa at 400°C), the risk of coolant boiling is essentially eliminated, meaning Pb-Bi reactors can operate at high temperatures without coolant pressurization. The reactor outlet coolant temperature of LFR designs is in the range of 500°C to 600 °C, comparable to that of sodium fast reactors and much higher than the typical output temperature of light water reactors. [3] These high outlet temperatures allow for the use of more efficient thermodynamic cycles for power generation, such as a traditional Rankine cycle using superheated steam as well as Brayton cycle using supercritical CO2. [1]
Theoretically, coolant outlet temperatures as high as 800°C from LFRs are possible, but are currently limited by the thermal integrity of the fuel elements and reactor components. [1] Thus, future proposed LFR designs utilizing nitride fuels and advanced structural materials may allow for even high operating temperatures. Some proposed advanced LFR designs (such as US's SSTAR reactor), propose the use of a direct lead-CO2 heat exchanger or a supercritical CO2 Brayton cycle to take advantage of these temperatures. [1] However, no such reactors have been built to date, with most historic and near-term designs utilizing a closed-loop of circulating lead coolant at 500°C to 600°C with a heat exchanger to drive a traditional supercritical steam Rankine cycle, albeit at higher efficiencies than conventional reactors. [1]
As with sodium fast reactors, lead fast reactor designs are classified into two major categories: loop-type designs in which the metal coolant is circulated into an external heat exchanger outside the main reactor vessel, and pool-type designs (which integrate the primary system components, including the reactor core and heat exchangers, within a single large pool of liquid metal). [1] Loop-type LFR designs were employed in the lead fast reactors deployed in Soviet nuclear submarines due to their higher achievable power density and smaller space requirements. [1] However, due to higher melting points compared to Na, Pb-based coolants are more vulnerable to solidifying in external loops (which must be externally heated), and thus were prone to leaks and solidified blockages. [4] Thus, essentially all modern proposed LFRs instead employ a pool-type design (Fig. 1). This has the benefit of greatly simplifying the coolant circulation system (improving passive safety), increased containment of radioactive daughter products, and partially addresses the issue of solidification of the lead coolant. [4]
Due the high atomic number of both Pb and Bi, colliding neutrons readily scatter off these heavy nuclei with very little energy loss, meaning both pure lead and LBE have very small neutron capture cross sections of only 0.171 and 0.09 Barns, respectively, lower than sodium's 0.53-barn cross-section. [5] As a result, the neutrons produced by fission reactions in an LFR undergo very little moderation, retaining fast-spectrum kinetic energies long enough to collide with fertile isotopes in the fuel. Thus, they are capable of converting U-238 and other fertile isotopes into fissionable isotopes such as Pu-239. [1,4] Thus, Pb fast reactors, like Na-cooled reactors, are designed to burn fertile fuel by continuously producing new fissionable isotopes during operation, and thus, in principle, can used as breeder reactors to produce more fuel than they consume. [1] However, historically, Pb fast reactors were not used for Pu production/extraction (with other fast reactor designs instead being preferred for this application), and thus, most LFRs are designed with a breeding ratio of around 1 (converting roughly all fertile isotopes to fissile ones during the fuel life) in order to enable efficient fuel utilization, minimal refueling, and a closed fuel cycle. [1,4]
To date, the only lead-type fast reactors that have been successfully built and operated were constructed in the 1960s by the Soviet Union (and subsequently the Russian Federation) for military use in a few nuclear submarines. [4] While the U.S. and other countries briefly investigated lead reactor coolants for their own designs, corrosion and design issues prevented their further implementation, and most countries instead opted to instead explore Na-based fast reactors. [6] The Soviets, however, continued exploring the technology and were ultimately able to overcome some of the initial technical challenges and develop workable designs. [7]
In addition to two land prototype reactors, a total of 15 LBE reactor cores were operated in eight Alfa-class submarines (Fig. 2) from the early 1960s until decommissioning in 1995, resulting in around 80 years worth of total operational experience. [4] The reactor designs were capable of two switchable thermal outputs of 73 and 155 MW, and were pursued over water-reactor designs in order to achieve the power densities required for the Alfa-classs fast-attack combat role. [7] The project was initiated by A. B. Petrov of the Malakhit Design Bureau (SKB-143) and then completed under M. G. Rusanov, resulting in construction of the first LBE-powered submarine (K-27) in 1965. [7] No design schematics of LBE-cooled submarine reactors have ever been openly published, so detailed design information remains limited. [4]
Unlike proposed Generation-IV LFR designs, the Alfa-class reactors all utilized a lead-bismuth eutectic (LBE) coolant in a compact, loop-type design, which benefitted from a lower melting point of 124.5°C that was easier to keep liquified, but had the significant drawback of containing Bi-209, which undergoes neutron capture to form Po-210 as follows:
Po-210 is an extremely radiotoxic α-emitting isotope with a long half-life (138 days). [1] As a result, significant quantities of Po-210 can accumulate in LBE reactors during operation, which is retained in the liquid coolant as lead polonide (PbPo). [1,4] Not only does irradiated LBE coolant pose a significant radiation hazard for an extended time period, but contact with moisture can further result in the formation of polonium hydride (PoH2) which is a volatile gas at room temperature (boiling point of 35.3°C), representing an enormous airborne hazard if any activated coolant is not adequately contained. [4]
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| Fig. 2: Photograph of Soviet Alfa-class submarine in operation. (Source: Wikimedia Commons) |
Due to the hazard posed by Po-210 release, the Soviet submarine reactors were automated and designed to largely run unmanned, but any reactor malfunction or leak that required maintenance still posed a significant hazard to the crew. [7] All Soviet LBE reactors built utilized highly enriched U-235 as fuel and featured an slightly slowed, epithermal neutron energy spectrum as a result of their compact size and low capacity factor. [1] Due to their design, it was impossible to refuel these reactors without the metal coolant solidifying, and were instead replaced once the core was depleted. [7]
In response to military urgency, the LBE-cooled submarine reactors were put into service as operational prototypes with only partial land testing. Two main models of LBE reactor (BM-40A and OK-550) were produced, which had either two or three redundant coolant loops, but both designs were very susceptible to coolant leaks, which could solidify and damage the reactor or release hazardous Po-210. [7] Furthermore, if the reactor temperature fell below 123°C, the LBE coolant would solidify in the core and the reactor would have to be scrapped as it became permanently inoperable. Thus, when at port, the LBE reactors had to be externally heated through a complex on-shore superheated steam system. [7]
Limited information has been publicly released about safety incidents that occurred with the Alfa-class LBE submarine reactors. However, two major incidents of note occurred in the submarines K-27 and K-377 (formerly K-47). In 1968, one of the reactors in K-27 suffered a major pipe leak accident due to corrosion, releasing radioactive gas that resulted in the deaths of nine crew members. [7] In 1972, the reactor in K-377 suffered a coolant leak that rapidly froze, damaging the fuel access system, and permanently disabling the reactor, which then had to be sealed with bitumen and disposed of at sea. Thus, while they exhibited exceptional power density and performance for their size, the LBE reactors ultimately faced too many technical problems, leading all of the Russian LBE fleet to either be retired or retrofitted with newer-generation PWRs by 1995. [7]
Overall, despite facing frequent technical challenges and some major accidents, the Russian LBE submarine reactors provided a valuable experience base for the development of lead-cooled reactors, leading to renewed interest for LFRs starting in the 1980s in Russia as well as in modern Generation IV designs. [1,2]
Despite being a relatively nascent technology compared to some other Generation IV designs, Pb fast reactors possess a number of attractive design attributes. Perhaps the largest inherent advantage of lead-cooled fast reactors over their Na-cooled counterparts is that molten Pb, unlike Na, does not violently react in either water or air, limiting the hazards posed by a coolant leak (such as fires). Pb, unlike Na metal, is also cheap and readily refined, decreasing the cost of coolant. [1,8] Some other design advantages are outlined below:
As discussed above, Pb-cooled fast reactors benefit from similar advantages to Na-cooled reactors, including lack of coolant boiling, high coolant conductivity, high thermal mass, lack of pressurization, a fuel-efficient fast neutron spectrum, energy-efficient high-temperature operation, and some passive heat dissipation. [1,2]
Pb is very effective at shielding from γ radiation and can readily retain fission products such as I and Cs, meaning radiation levels near the reactor are very low. [1,4]
Because of its small capture cross-section, Pb undergoes very slow activation by neutrons. Thus, a pure Pb-cooled reactor will have minimal radioactivity during operation, potentially facilitating reactor servicing. This is in contrast to Na-based coolants, which undergo neutron activation to form radioactive Na-24 with a half-life of 15 hours. [2,8] Due to the presence of Bi-209 impurities and transmutation reactions (208Pb + n → 209Pb → 209Bi + e-) in lead, minor amounts of radioactive Po-210 are produced even in pure Pb coolants. However, this is generally manageable in pure-Pb reactors, and only a significant issue with Pb-Bi eutectic coolants (discussed below).
In addition to allowing LFRs to operate with a harder (higher energy) neutron spectrum compared to other fast-reactor types, leads low neutron moderation permits core designs with larger spacing between fuel rods compared to sodium fast reactors, which decreases the flow resistance of coolant through the core, reduces the risk of coolant flow blockage, and allows some designs to be partially (or entirely) cooled passively, enhancing safety in the event of sudden reactor shutdown. [1]
One of the most significant design challenges of lead-cooled reactors is that molten lead is highly erosive and corrosive to structural steels, fuel rod casings, and internal reactor components, readily dissolving Fe, Cr, and Ni from unprotected metals at rates that increase with temperature. [4] Corrosion rates can be somewhat limited by actively managing dissolved oxygen, limiting circulating coolant velocities, careful monitoring, and specialized surface treatments, but LFRs will likely require novel material and reactor construction techniques to address these issues is commercial deployment. [1,2] Some other design disadvantages include:
Similar to Na-based reactors, the fast neutron spectrum of Pb fast reactors may present a proliferation hazard, due to their ability to breed nuclear fuel which in principle can be refined to produce weapons-grade Pu. To mitigate this risk, all proposed Generation IV designs of LFRs feature a closed-fuel cycle that makes extraction and purification of Pu more difficult, but nevertheless, some security and policy concerns remain about widespread deployment of fast reactors. [3,4]
Pb fast reactors have only been operated in submarines and have never been deployed for on-site civilian power generation. Thus, LFR designs for power generation are far less mature and draw from a much smaller experience base compared to Na fast reactors. [1,4]
Since both Pb and Pb-Bi are very dense, lead coolant adds a significant amount of weight to a reactor core, requiring additional structural design and seismic protection. [2,4]
While most modern grid-scale LFR proposals opt for its inherent safety and design advantages, larger fuel pin spacing in LFRs comes at the cost of reduced power density, so most designs achieve core power densities in the range of 80-160 MW/m3, lower than the achievable power densities of Na fast reactors (around 300 MW/m3), but comparable to Pressurized Water Reactors (PWR) that achieve 100 MW/m3. [8]
As discussed in the prior historical section, Pb-Bi eutectic (LBE) have the significant drawback of containing 55.5% Bismuth-209 by mass, which undergoes neutron capture to form Po-210
which is an extremely radiotoxic, α-emitting isotope with a relatively long half-life (138 days). [1] As a result of the significant health hazards posed by neutron activation of LBE coolant, the majority of proposed Generation IV LFR designs utilize bismuth-free lead coolants in order to minimize Po-210 production.
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| Fig. 3: 3D-Rendered Model of the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED), a pool-type Generation IV lead-fast reactor design. [9] (Courtesy of Springer Nature) |
Despite their drawbacks, the efficiency and potential design advantages of lead fast reactors have led to renewed interest in the technology in recent decades. Various Generation IV designs for lead- cooled fast reactors (LFRs) have been proposed. In line with Generation IV goals, these designs typically feature a fast-neutron spectrum, closed fuel cycle compatibility, proliferation resistance, pool-type cooling, and passive safety systems. [2] Among the most mature designs is the Russian BREST-OD-300 reactor, which is a 300 MW pure-lead cooled pool-type reactor with partial natural circulation, a closed MOX fuel cycle, widely spaced fuel lattice. [1,3] The BREST-OD-300 is meant as a pilot scale proof of concept to the larger BREST-OD-1200 MW reactor and may be the earliest LFR to enter operation. Construction was approved February 10, 2021 and was slated to be completed sometime in the 2020s. [1] Other pool-type reactor LFR concepts include the European Lead-cooled Fast Reactor (ELFR), the Advanced Lead-cooled Fast Reactor European Demonstrator (ALFRED, Fig. 3), and the US's Small Secure Transportable Autonomous Reactor (SSTAR) concept, a naturally-circulated small-scale CO2-cooled reactor. [1,3] Efforts among private companies include the Westinghouse Lead Fast Reactor (LFR) concept as well as the LFR-AS-200 pool-type reactor from UK startup Newcleo. [1] Overall, LFRs represent an interesting nuclear technology and show promise as a potential Generation IV design, although ongoing design challenges, such as coolant-induced corrosion, require further research and development.
© Alex Fontani Herreros. The author warrants that the work is the author's own and that Stanford University provided no input other than typesetting and referencing guidelines. The author grants permission to copy, distribute and display this work in unaltered form, with attribution to the author, for noncommercial purposes only. All other rights, including commercial rights, are reserved to the author.
[1] C. F. Smith and L. Cinotti, "Lead-cooled Fast Reactors (LFRs)," in Handbook of Generation IV Nuclear Reactors, 2nd Ed., ed. by I. L. Pioro (Woodhead Publishing, 2023).
[2] "Technology Roadmap Update for Generation IV Nuclear Energy Systems," Gen IV International Forum, January 2014.
[3] "Annual Report 2022," GenIV International Forum, 2022.
[4] A. E. Waltar, D. R. Todd, and P. V. Tsvetkov, Eds., Fast Spectrum Reactors (Springer, 2012).
[5] K. Tuček, J. Carlsson and H. Wider, "Comparison of Sodium and Lead-Cooled Fast Reactors Regarding Severe Safety and Economical Issues," Nucl. Eng. Des. 236, 1589 (2006).
[6] J. R. Weeks, "Lead, Bismuth, Tin and Their Alloys as Nuclear Coolants," Nucl. Eng. Des. 15, 363 (1971).
[7] M. Rawool-Sullivan, P. D. Moskowitz, and L. N. Shelenkova, "Technical and Proliferation-Related Aspects of the Dismantlement of Russian Alfa-Class Nuclear Submarines," Nonproliferation Rev. 9, 161 (2002).
[8] A. Rojas, "Sodium-Cooled Fast Reactors as a Generation IV Nuclear Reactor," Physics 241, Stanford University, Winter 2018.
[9] J. Ongena et al., "Advances in the Physics and Thermodynamics of Nuclear Reactors," Eur. Phys. J. Plus 136, 845 (2021).
