Fast neutron reactors

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Most fast breeder reactors built to date have been based on either molten sodium or lead as coolants, although molten salts are planned by some designs.

See Fast Neutron Reactors World Nuclear Association



EBR1 Raphael Telis; facebook; 22 Dec 2018

On December 20, 1951, the Experimental Breeder Reactor I (EBR-I) in Arco, Idaho, generated usable electricity for the first time. It produced enough electricity to light four 200-W light bulbs, and a day later, the reactor produced enough power to light the whole building. EBR-I became the first power plant in the world to produce usable electricity using nuclear power.

EBR-I's primary mission, according to the US Atomic Energy Commission (AEC) – predecessor of the US Nuclear Regulatory Commission (NRC), was to develop and test the concept of the breeder reactor – a vision pursued by the Italian Enrico Fermi and his colleague, Walter Zinn, first director of the Argonne National Laboratory (ANL) in Illinois. Zinn also led the team that built the EBR-I.

The idea behind the breeder was to maximize the useful energy that could be obtained from natural uranium, only one form of which – U-235 – is a useful fuel in today's nuclear reactors. U-235, however, makes up less than one percent of all natural uranium, with the rest being U-238. Inside a nuclear reactor, U-238 atoms capture neutrons released during fission and are converted into plutonium-239 (Pu-239) – a man-made element that also can fuel reactors. Thus, breeding makes it possible to use virtually all the energy from natural uranium.

Before electricity was ever produced by a nuclear reactor, scientists and engineers from Argonne had moved to Idaho to work on the project. These professionals included EBR-I associate project engineer Leonard Koch, senior engineer Mike Novick, and physicist Newman Pettitt. Construction of EBR-I by Bechtel Corporation began in 1949 and criticality was attained in August 1951. When the bulbs were first lit, one scientist, in subdued exuberance, said simply, "This is it." There were no shows of open emotion, no hand clapping or whooping and hollering.

The day after that event, EBR-I's output was boosted to 100 kWe, enough to power all of the electrical equipment contained in the small brick building that housed the reactor. These events were recorded on a wall of the reactor building by those present.

The EBR-I operated for 12 years, and on December 30, 1963, it was officially shutdown. On August 26, 1966, President Lyndon B. Johnson presided over ceremonies that designated the retired reactor a Registered Historical National Monument. Johnson and the chemist Glenn Seaborg, who had done breeding research in the 1940s with Zinn and others, affixed a commemorative plaque to the reactor's small brick building, which still stands on the Idaho site. It remains open to the public during the summer for those inclined to visit the U.S. Department of Energy's (DOE) Idaho National Laboratory (INL).

The EBR-I spawned an international nuclear industry that now plays a significant role in meeting the world's electricity needs. Today, 98 reactors provide 20 percent of the electricity consumed in the US. More than 450 reactors provide some 11 percent of the world’s electricity.

See also: American Physical Society, "This Month in Physics History – December 20, 1951: First Electricity Generated by Atomic Power" December 2014 (Volume 23, Number 11):

This Month in Physics History: December 20, 1951: First Electricity Generated by Atomic Power APS News; Dec 2014

Should one happen to drive through the high desert of eastern Idaho, one might stumble across what is now called the Idaho National Laboratory, a federal nuclear research facility that has played a key role in the development of nuclear power. It houses the decommissioned Experimental Breeder Reactor-I (EBR-I), the first nuclear reactor to generate usable electricity via fission. The man responsible for its design and operation was a Canadian physicist named Walter Henry Zinn.


EBR-II (Experimental Breeder Reactor-II) Argonne National Laboroatory

The Experimental Breeder Reactor-II (EBR-II) was originally designed and operated with emphasis on demonstrating a complete breeder-reactor power plant with on-site reprocessing of metallic fuel. This was successfully done from 1964 to 1969. During that five years, the reactor's Fuel Cycle Facility processed 35,000 fuel elements, produced 366 subassemblies, and assembled 66 control and safety rods. The facility was then converted from a breeder to a burner reactor. The new missions emphasized testing fuels and materials for larger, liquid metal reactors.

EBR-II was the backbone of the U.S. breeder reactor effort from 1964 to 1994, when research was terminated. The EBR-II accommodated as many as 65 experimental subassemblies at one time for irradiation and operational reliability tests. EBR-II also performed over 30,000 irradiation tests. Most recently, EBR-II was the prototype for the Integral Fast Reactor (IFR).

One feature new to the EBR-II was its pool-type design. Simply put, the reactor core, its fuel handling equipment, and many other systems of the reactor were submerged under molten sodium. This type of design had many benefits, including simplified design and construction, reduction of thermal stress, elimination of some heavily shielded external facilities, and, most importantly, increased safety.

The pool-type design, combined with its metal alloy fuel, made the EBR-II passively safe. That is, the reactor could safely shut down, without operator assistance, even if safety systems had failed. This safety feature was not dependent on control rods or computer monitoring, but on the laws of physics. This reliance on natural physical properties is the ultimate backup safety system for a nuclear power plant. This would make nuclear incidents, such as those that occured at Three Mile Island and Chernobyl, nearly impossible to duplicate. This was demonstrated in 1986, when EBR-II underwent a series of IFR safety tests. These tests simulated accidents involving loss of coolant flow. Even with the normal shutdown devices disabled, the reactor safely shut down without reaching excessive temperatures anywhere in the system.

EBR-II stopped operations in 1994 when it lost federal funding. The tests and experiments that have been conducted in EBR-II have contributed heavily to national and international reactor technology, especially FBR (Fast Breeder Reactor) technology.

Sad-ending story of EBR-II told by three of its pioneers Rod Adams; Atomic Insights; 24 Aug 2015

During the period between 1961 and 1994, an extraordinary machine called the Experimental Breeder Reactor 2 (EBR-II) was created and operated in the high desert of Idaho by a team of dedicated, determined, and distinguished people.

In 1986, that machine demonstrated that it could protect itself in the event of a complete loss of flow without scram and a complete loss of heat sink, also without a scram. Those tests were conducted carefully, with an expanded supervisory and operating staff while being witnessed by dozens of internationally respected scientists and engineers.


Integral Fast Reactor Argonne National Labs

The Integral Fast Reactor (IFR) is a fast reactor system developed at Argonne National Laboratory in the decade 1984 to 1994. The IFR project developed the technology for a complete system; the reactor, the entire fuel cycle, and the waste management technologies were all included in the development program. The reactor concept had important features and characteristics that were completely new and fuel cycle and waste management technologies that were entirely new developments. The reactor is a “fast” reactor – that is, the chain reaction is maintained by “fast” neutrons with high energy – which produces its own fuel. The IFR reactor and associated fuel cycle is a closed system. Electrical power is generated, new fissile fuel is produced to replace the fuel burned, its used fuel is processed for recycling by pyroprocessing – a new development – and waste is put in its final form for disposal. All this is done on one self-sufficient site.

GE and Hitachi want to use nuclear waste as a fuel Lin Edwards;; 18 Feb 2010


Sustainable Nuclear Barry Brook; Brave New Climate;

Various links to articles about the IFR

The Integral Fast Reactor (IFR) project Steve Kirsch; (personal website); 10 Aug 2008

Long discussion of IFRs and nuclear issues generally


Westinghouse announces Lead Cooled Fast Reactor initiative Will Davis; atomic power review; 9 Oct 2015

On Friday, October 9, Westinghouse announced that it had launched a program to work with the US Department of Energy in the development of a new, lead-cooled fast reactor (commonly, "LFR") which would combine the advantages of lead cooling (high temperatures, primarily, as well as lower pressures) with advanced accident tolerant fuel to push the Gen-IV envelope to what it perceives as Gen-V -- a term that seems to imply the "state of the art" in perhaps 30 or 50 years down the road.

Westinghouse proposes LFR project World Nuclear News ; 14 Oct 2015

Westinghouse is seeking to collaborate with the US Department of Energy (DOE) on the development of a lead-cooled fast reactor (LFR). It will be the company's first foray into fourth generation reactor designs.

The company announced on 8 October that it had submitted a project proposal for the LFR under the DOE's Advanced Reactor Industry Competition for Concept Development funding opportunity.

Westinghouse said that its project team includes members of the national laboratory system, universities and the private sector "with expertise in areas essential to the design and commercialization of an advanced LFR plant".

The Westinghouse LFR would be "designed to achieve new levels of energy affordability, safety and flexibility", the company said. In addition to featuring accident-tolerant fuel, the reactor's use of lead as a coolant "will further enhance reactor safety, and optimize the plant's economic value through lower construction costs and higher operating efficiency than other technologies", it said.

In addition to electricity generation, the Westinghouse LFR could be used for hydrogen production and water desalination, the company noted. It also said the reactor's load-following capabilities "would further support the increased use of renewable energy sources".


Leadcold SEALER

SEALER website

SEALER (Swedish Advanced Lead Reactor) is a lead-cooled reactor designed with the smallest possible core that can achieve criticality in a fast spectrum using 19.9% enriched uranium oxide (UOX) fuel. The rate of electricity production may vary between 3 to 10 MW, leading to a core-life between 10 and 30 full power years (at 90% availability). The reactor is designed to maintain a maximum temperature of the lead coolant below 450°C, making corrosion of fuel cladding and structural materials a manageable phenomenon, even over a life-span of several decades.

The safety features of lead mean that the core can manage a complete loss of off-site power for weeks before integrity of the fuel rods is challenged. Should any volatile fission products be released into the coolant, 99.99% will be chemically retained by the lead. The eventual release of noble gases and residual volatiles results in a radiological exposure at the site boundary which is smaller than the natural back-ground dose received during a few months. Hence, no accident scenario can lead to a situation where evacuation becomes necessary.


LeadCold entered Phase 1 of the Canadian Nuclear Safety Commission’s pre-license review in December 2016. The eventual objective is to receive a license for construction in Canada by end of 2021, aiming at having our first SEALER-unit ready for operation in 2025.


The future cost for purchasing a SEALER reactor is estimated at 100 million Canadian dollars.

Waste management

After 10 - 30 years of operation, the first SEALER units will be transported back to a centralised recycling facility. The plutonium and minor actinides present in the spent fuel may then be separated and converted into an inert matrix nitride fuel for indefinite recycle in SEALER reactors. The residual high level waste (mainly short lived fission products) will be vitrified and isolated from the biosphere in a geological repository for a period of less than 1000 years.

Canada set for first lead-cooled reactor by 2025 after $200mn funding boost Nuclear Energy Insider; 8 Mar 2017

LeadCold plans to use the large investment by Indian conglomerate Essel Group to fund pre-licensing, detailed engineering design, and development costs for a 3 MW demonstration reactor ahead of deployment on remote sites, Janne Wallenius, CEO of LeadCold, told Nuclear Energy Insider in an interview.

A number of advanced nuclear reactor developers are targeting the Canadian market, where the risk-informed regulatory framework is considered more supportive for licensing new designs than in the U.S. and where numerous remote communities and industrial facilities represent captive electricity consumers.

In late December LeadCold filed its fast neutron Swedish Advanced Lead Reactor (SEALER) design with the Canadian Nuclear Safety Commission (CNSC) for phase 1 of the pre-licensing review.

LeadCold aims to deploy its reactors within the remote Arctic regions in the Northwest Territories and Nunavut, where power users are off-grid and depend on high-cost diesel-fired generators.

The company, a spin off from the Royal Institute of Technology in Stockholm (KTH), is developing a reactor that can provide a capacity of between 3 MW and 10 MW, to meet the different power needs of remote communities and mining customers.

The Levelized Cost of Energy (LCOE) is estimated at C$450/MWh ($337/MWh) for 3 MW capacity and C$220/MWh for 10 MW, Wallenius told Nuclear Energy Insider in an interview.

Electricity costs can be as high as C$2,000/MWh in the most remote regions of Northern Canada, Roger Humphries, director of SMR Development for Amec Foster Wheeler, said in an interview in January 2016.

Some 200,000 people live in over 200 remote communities and 80% of their power comes from diesel-fired generators.



Dounreay Wikipedia


Sodium cooled: BN-800 etc

The BN-800 Fast Reactor – a Milestone on a Long Road Syndroma; Energy Matters; 4 November 2016

Guest post by Russian commenter Syndroma who was trained in IT and now works in his family business. The BN-800 was commissioned this week.

Lead cooled: BREST-OD-300

Construction Contracts Awarded For Fast Neutron Reactor Facility Nuclear Street; 11 Dec 2019

Siberian Chemical Plant (a subsidiary of TVEL Fuel Company of ROSATOM in Seversk, Tomsk region, West Siberia) has signed the contract with Concern Titan-2 engineering company for construction and installation works within the project of BREST-OD-300 lead-cooled fast neutron reactor facility.

Rosatom said December 5 that one of its subsidiaries had signed a $420.9 million contract for construction and installation within the project of BREST-OD-300 lead-cooled fast neutron reactor facility.

The contract was signed by the Siberian Chemical Plant, which is a subsidiary of Rosatom through the TVEL Fuel Company, and Concern Titan-2 engineering company.

“The contractor will accomplish the construction of the reactor hall, turbine hall and related infrastructure facilities of the power plant. The completion is scheduled till the end of 2026,” the company announced.

“Contracting for BREST-OD-300 reactor facility and power unit construction is the major long-anticipated milestone of “the Breakthrough” project implementation in 2019. After commencing the works for the fabrication/refabrication facility, we are moving further to the construction of the key facility of PDEC which should become the prototype of the nuclear power of the future. Besides, construction and operation of PDEC facilities would create over 800 new jobs in Seversk,” said Vitaly Khadeev, Vice President for Development of Closed Nuclear Fuel Cycle Technologies and Industrial Facilities at TVEL JSC.

“Proryv” or “the Breakthrough” project targets the creation of the new technology platform for the industry with a closed nuclear fuel cycle, as well as tackling the issues of spent nuclear fuel and radioactive waste. One of the project implementations is the construction of the Pilot Demonstration Energy Complex with a lead-cooled BREST-OD-300 fast neutron reactor facility with an on-site closed nuclear fuel cycle.

A 300 MW plant powered by the innovative fast reactor is the key facility of the Pilot Demonstration Energy Complex (PDEC), which is in process of construction at the Siberian Chemical Plant, as a part of the strategic Russian nuclear project “Proryv” or “the Breakthrough”. In addition to the power unit, PDEC will include an on-site closed nuclear fuel cycle with a facility for fabrication/re-fabrication of mixed nitride uranium-plutonium nuclear fuel, as well as a spent fuel reprocessing facility.

TVEL Fuel Company of ROSATOM provides nuclear fuel for 76 power reactors in 15 countries worldwide, research reactors in eight countries, as well as transport reactors of the Russian nuclear fleet. As such, every sixth power reactor in the world operates on fuel manufactured by TVEL.


India’s First Prototype Fast Breeder Reactor Has a New Deadline. Should We Trust It? R. Ramachandran; The Wire; 20 Aug 2020

The PFBR is a nuclear power reactor currently under construction at the Madras Atomic Power Station in Kalpakkam, Tamil Nadu. Fast breeder reactors, or FBRs, in general produce more fissile material than they consume. The PFBR in Kalpakkam will use a mixed oxide of plutonium-239 – derived from reprocessed spent fuel from the thermal pressurised heavy water reactors – and uranium-238 as fuel to generate energy in a nuclear reaction. This reaction will also produce – or ‘breed’ – more plutonium-239. This is possible because the reaction converts both uranium-238 in the fuel mix as well as a blanket of depleted uranium surrounding the core into plutonium.

This plutonium will then be processed and used as nuclear fuel in a chain of commercial FBRs that constitutes stage II of the nuclear programme. The stage will also include FBRs that will use thorium-232, mined in India, as a blanket. Thorium will get converted to uranium-233, which will serve as the fuel for advanced reactors in stage III. Ultimately, these reactors will burn uranium-233 and convert thorium-232 to more uranium-233, creating a self-sustaining cycle of nuclear power generation.


Chinese fast reactor starts supplying electricity World Nuclear News; 21 Jul 2011

Exactly one year after achieving first criticality, China's experimental fast neutron reactor has been connected to the electricity grid.

The sodium-cooled, pool-type fast reactor has been constructed with some Russian assistance at the China Institute of Atomic Energy (CIEA), near Beijing, which undertakes fundamental research on nuclear science and technology. The reactor has a thermal capacity of 65 MW and can produce 20 MW in electrical power. The CEFR was built by Russia's OKBM Afrikantov in collaboration with OKB Gidropress, NIKIET and Kurchatov Institute.

Xu Mi, chief engineer at the CEFR program at CIEA, told Bloomberg that the unit was connected to the grid at 40% capacity. "The next step for us is to increase the generating capacity of the reactor to 100% while connected to the grid," he said. "After that, we can use the technology to build our own commercial fast reactors."

Beyond the pilot plant, China once planned a 600 MWe commercial scale version by 2020 and a 1500 MWe version in 2030 but these ambitious ideas have been overtaken by the import of ready-developed Russian designs. In October 2009, an agreement was signed by CIAE and China Nuclear Energy Industry Corporation (CNEIC) with AtomStroyExport to start pre-project and design works for a commercial nuclear power plant with two BN-800 reactors with construction to start in August 2011, probably at a coastal site. The project is expected to lead to bilateral cooperation of fuel cycles for fast reactors, which promise to vastly extend the fuel value of uranium as well as reduce radioactive wastes.

Travelling Wave Reactor

A TWR converting U-238 to Pu-239.
Red: U-238, light green: Pu-239, black: fission products.
Intensity of blue colour between squares indicates neutron density

A travelling-wave reactor (TWR) is a proposed (i.e. paper) reactor designed to use fissile U-235 (as in normal enriched uranium fuel) to also convert fertile material such as depleted uranium (U-238), natural (unenriched) uranium, thorium, and/or spent nuclear fuel, into further usable fuel through nuclear transmutation.

TWRs differ from other kinds of fast-neutron and breeder reactors in their claimed ability to use fuel efficiently without further enrichment or reprocessing.

The name refers to the fact that fission occurs in a boundary zone in the reactor core which advances over time; a wave of fission travelling through the core. TWRs could theoretically run self-sustained for decades without refuelling or removing spent fuel.

See the Wikipedia article "Traveling Wave Reactor" for the history and physics of this sort of reactor.


Terrapower, a company co-founded by Bill Gates, is working on a design of Travelling Wave Reactor in which the reactor's fuel rods are intended to be shuffled around, with fissile material being moved into the centre of the core replacing burned up material, so that the region of fission remains static rather than expanding through the core like a ripple in a pond (as in the animated graphic shown here).

TerraPower's Traveling Wave Reactor Technology page has a graphic showing the physical design of their reactor.

In 2015 TerraPower contracted with China National Nuclear Corp. to build a prototype of the reactor, followed by a commercial version but conflicted relationships between the US under Trump and China appear to have paused or halted the project.


Natrium is a joint project between GE-Hitachi and TerraPower. It product combines a 345-MWe sodium fast reactor (SFR) with a molten salt energy storage system. It seems to be a combination of TerraPower’s Traveling Wave Reactor (TWR) and GEH’s PRISM technology. See GE Hitachi, TerraPower Team on Nuclear-Storage Hybrid SMR by Sonal Patel in Power magazine (3 Sep 2020). Reuters reports that the reactor would have a power output of 345MWe and cost $1bn.

Sodium safety

Many Fast Breeder Reactors use molten sodium as coolant. Sodium is highly reactive, reacting vigorously with air and even more with water.

Metal Fires in Fast Reactors: Part I Tara J. Olivier, Ross F. Radel, Steven P. Nowlen, Thomas K. Blanchat, John C. Hewson; Nuclear Green Revolution; 23 Feb 2015

Refers to report by Sandia Labs

Criticism of Sodium-cooled reactor designs by Robert Steinhaus

The Short of Why I do not like the Terrapower TWR pool style SFR

(Safety Concern) - Sodium Cooled Fast Reactors like the Terrapower TWR contain a lot of reactive sodium coolant.

The French Superphenix used 5500 metric tons of sodium (3300 tons in the primary reactor vessel) for a 1.2 GWe reactor - exact figures have not so far been provided for other mature SFR designs like GE PRISM, BN-800, or the Terrapower TWR so scaling the Superphenix numbers is about the best analysts can do given the reticence of current SFR designers to reveal to the public and decision makers the numbers that would allow fair and accurate analysis of the potential hazard of their SFR designs.

Sodium Safety - Sodium reacts exothermically with liquid water or steam to generate sodium hydroxide and hydrogen:

Na + H2O -> NaOH + 1/2H2 + heat Heat of reaction: ~ 162kJ/mole-Na (around 7.05MJ/kg-Na)

Hydrogen tends to accumulate in the roof area of a reactor containment building and if the conditions are right, hydrogen air mixtures can detonate.

Reference – G. Manzini and F. Parozzi “Sodium Safety” (Intrinsically safer nuclear technology exits)-

A 1 GWe Pu-239 fueled Sodium Cooled Reactor, like the Terrapower TWR, fissions about 1 ton of Pu-239 fuel while producing about 1 ton (1000 kilograms) of fission products a year. The Terrapower TWR is designed to retain spent fuel inside the reactor and "shuffle fuel" while operating. One effect of this arrangement is that over the designed lifetime of the reactor, more and more fission products accumulate.After 60 years of operation, some of the early produced fission products inside the TWR will have decayed to lower levels of activity - becoming less radioactive. On the other hand, depending on the specific arrangements of the fuel shuffling system used in the TWR, old fission products which are exposed to neutrons will continue to absorb neutrons and become further neutron activated, in some instances making them more radioactive. A more complex modeling of 60 year isotopic inventory of a Terrapower TWR would need to be made with a code like ORIGEN-S to give a reliable and accurate assessment of radioactivity of the fission product isotopic inventory after 60 years of operation. Still, I think it can be fairly stated that the radiological inventory of a sodium cooled TWR would be very large indeed and the chemical stored energy in Sodium coolant also very large which means that the potential is there for a really large accident (INES-7) if any of the SFR engineered safety systems (like the Argon cover gas over the main reactor sodium pool) is ever even momentarily lost as the result of an air leak..

Safer reactor coolants than hot reactive sodium exist (I like Molten Salts or helium).