Integral fast reactor

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Experimental Breeder Reactor II, which served as the prototype for the Integral Fast Reactor

The integral fast reactor (IFR, originally advanced liquid-metal reactor) is a design for a nuclear reactor using fast neutrons and no neutron moderator (a "fast" reactor). IFR would breed more fuel and is distinguished by a nuclear fuel cycle that uses reprocessing via electrorefining at the reactor site. 

IFR development began in 1984 and the U.S. Department of Energy built a prototype, the Experimental Breeder Reactor II. On April 3, 1986, two tests demonstrated the inherent safety of the IFR concept. These tests simulated accidents involving loss of coolant flow. Even with its normal shutdown devices disabled, the reactor shut itself down safely without overheating anywhere in the system. The IFR project was canceled by the US Congress in 1994, three years before completion.

The proposed Generation IV Sodium-Cooled Fast Reactor is its closest surviving fast breeder reactor design. Other countries have also designed and operated fast reactors. 

S-PRISM (from SuperPRISM), also called PRISM (Power Reactor Innovative Small Module), is the name of a nuclear power plant design by GE Hitachi Nuclear Energy (GEH) based on the Integral Fast Reactor.

Overview

The IFR is cooled by liquid sodium or lead and fueled by an alloy of uranium and plutonium. The fuel is contained in steel cladding with liquid sodium filling in the space between the fuel and the cladding. A void above the fuel allows helium and radioactive xenon to be collected safely without significantly increasing pressure inside the fuel element, and also allows the fuel to expand without breaching the cladding, making metal rather than oxide fuel practical. 

The advantage of lead as opposed to sodium is that it is not reactive chemically, especially with water or air. The disadvantages are that liquid lead is far more viscous than liquid sodium (increasing pumping costs), and there are numerous radioactive neutron activation products, while there are essentially none from sodium.

Basic design decisions

Metallic fuel

Metal fuel with a sodium-filled void inside the cladding to allow fuel expansion has been demonstrated in EBR-II. Metallic fuel makes pyroprocessing the reprocessing technology of choice.

Fabrication of metallic fuel is easier and cheaper than ceramic (oxide) fuel, especially under remote handling conditions.

Metallic fuel has better heat conductivity and lower heat capacity than oxide, which has safety advantages.

Sodium coolant

Use of liquid metal coolant removes the need for a pressure vessel around the reactor. Sodium has excellent nuclear characteristics, a high heat capacity and heat transfer capacity, low viscosity, a reasonably low melting point and a high boiling point, and excellent compatibility with other materials including structural materials and fuel. The high heat capacity of the coolant and the elimination of water from the core increase the inherent safety of the core.

Pool design rather than loop

Containing all of the primary coolant in a pool produces several safety and reliability advantages.

Onsite reprocessing using pyroprocessing

Reprocessing is essential to achieve most of the benefits of a fast reactor, improving fuel usage and reducing radioactive waste each by several orders of magnitude.

Onsite processing is what makes the IFR integral. This and the use of pyroprocessing both reduce proliferation risk.

Pyroprocessing (using an electrorefiner) has been demonstrated at EBR-II as practical on the scale required. Compared to the PUREX aqueous process, it is economical in capital cost, and is unsuitable for production of weapons material, again unlike PUREX which was developed for weapons programs.

Pyroprocessing makes metallic fuel the fuel of choice. The two decisions are complementary.

Summary

The four basic decisions of metallic fuel, sodium coolant, pool design, and onsite reprocessing by electrorefining, are complementary, and produce a fuel cycle that is proliferation resistant and efficient in fuel usage, and a reactor with a high level of inherent safety, while minimizing the production of high-level waste. The practicality of these decisions has been demonstrated over many years of operation of EBR-II.

Advantages

• Breeder reactors (such as the IFR) could in principle extract almost all of the energy contained in uranium or thorium, decreasing fuel requirements by nearly two orders of magnitude compared to traditional once-through reactors, which extract less than 0.65% of the energy in mined uranium, and less than 5% of the enriched uranium with which they are fueled. This could greatly dampen concern about fuel supply or energy used in mining.
• Fast reactors can "burn" long lasting nuclear transuranic waste (TRU) waste components (actinides: reactor-grade plutonium and minor actinides), turning liabilities into assets. Another major waste component, fission products (FP), would stabilize at a lower level of radioactivity than the original natural uranium ore it was attained from in two to four centuries, rather than tens of thousands of years. The fact that 4th generation reactors are being designed to use the waste from 3rd generation plants could change the nuclear story fundamentally—potentially making the combination of 3rd and 4th generation plants a more attractive energy option than 3rd generation by itself would have been, both from the perspective of waste management and energy security.
• The use of a medium-scale reprocessing facility onsite, and the use of pyroprocessing rather than aqueous reprocessing, is claimed to considerably reduce the proliferation potential of possible diversion of fissile material as the processing facility is in-situ/integral.

Safety

In traditional light water reactors (LWRs) the core must be maintained at a high pressure to keep the water liquid at high temperatures. In contrast, since the IFR is a liquid metal cooled reactor, the core could operate at close to ambient pressure, dramatically reducing the danger of a loss-of-coolant accident. The entire reactor core, heat exchangers and primary cooling pumps are immersed in a pool of liquid sodium or lead, making a loss of primary coolant extremely unlikely. The coolant loops are designed to allow for cooling through natural convection, meaning that in the case of a power loss or unexpected reactor shutdown, the heat from the reactor core would be sufficient to keep the coolant circulating even if the primary cooling pumps were to fail. 

The IFR also has passive safety advantages as compared with conventional LWRs. The fuel and cladding are designed such that when they expand due to increased temperatures, more neutrons would be able to escape the core, thus reducing the rate of the fission chain reaction. In other words, an increase in the core temperature will act as a feedback mechanism that decreases the core power. This attribute is known as a negative temperature coefficient of reactivity. Most LWRs also have negative reactivity coefficients; however, in an IFR, this effect is strong enough to stop the reactor from reaching core damage without external action from operators or safety systems. This was demonstrated in a series of safety tests on the prototype. Pete Planchon, the engineer who conducted the tests for an international audience quipped "Back in 1986, we actually gave a small [20 MWe] prototype advanced fast reactor a couple of chances to melt down. It politely refused both times."

Liquid sodium presents safety problems because it ignites spontaneously on contact with air and can cause explosions on contact with water. This was the case at the Monju Nuclear Power Plant in a 1995 accident and fire. To reduce the risk of explosions following a leak of water from the steam turbines, the IFR design (as with other sodium-cooled fast reactors) includes an intermediate liquid-metal coolant loop between the reactor and the steam turbines. The purpose of this loop is to ensure that any explosion following accidental mixing of sodium and turbine water would be limited to the secondary heat exchanger and not pose a risk to the reactor itself. Alternative designs use lead instead of sodium as the primary coolant. The disadvantages of lead are its higher density and viscosity, which increases pumping costs, and radioactive activation products resulting from neutron absorption. A lead-bismuth eutectate, as used in some Russian submarine reactors, has lower viscosity and density, but the same activation product problems can occur.

Efficiency and fuel cycle

Medium-lived
fission products

Prop: Unit:	t½ (a)	Yield (%)	Q * (keV)	βγ *
155Eu	4.76	0.0803	252	βγ
85Kr	10.76	0.2180	687	βγ
113mCd	14.1	0.0008	316	β
90Sr	28.9	4.505	2826	β
137Cs	30.23	6.337	1176	βγ
121mSn	43.9	0.00005	390	βγ
151Sm	88.8	0.5314	77	β

The goals of the IFR project were to increase the efficiency of uranium usage by breeding plutonium and eliminating the need for transuranic isotopes ever to leave the site. The reactor was an unmoderated design running on fast neutrons, designed to allow any transuranic isotope to be consumed (and in some cases used as fuel). 

Compared to current light-water reactors with a once-through fuel cycle that induces fission (and derives energy) from less than 1% of the uranium found in nature, a breeder reactor like the IFR has a very efficient (99.5% of uranium undergoes fission) fuel cycle. The basic scheme used pyroelectric separation, a common method in other metallurgical processes, to remove transuranics and actinides from the wastes and concentrate them. These concentrated fuels were then reformed, on site, into new fuel elements. 

The available fuel metals were never separated from the plutonium isotopes nor from all the fission products, and therefore relatively difficult to use in nuclear weapons. Also, plutonium never had to leave the site, and thus was far less open to unauthorized diversion.

Another important benefit of removing the long half-life transuranics from the waste cycle is that the remaining waste becomes a much shorter-term hazard. After the actinides (reprocessed uranium, plutonium, and minor actinides) are recycled, the remaining radioactive waste isotopes are fission products, with half-life of 90 years (Sm-151) or less or 211,100 years (Tc-99) and more; plus any activation products from the non-fuel reactor components.

Comparisons to light-water reactors

Buildup of heavy actinides in current thermal-neutron fission reactors, which cannot fission actinide nuclides that have an even number of neutrons, and thus these build up and are generally treated as Transuranic waste after conventional reprocessing. An argument for fast reactors is that they can fission all actinides.

Nuclear waste

IFR-style reactors produce much less waste than LWR-style reactors, and can even utilize other waste as fuel. 

The primary argument for pursuing IFR-style technology today is that it provides the best solution to the existing nuclear waste problem because fast reactors can be fueled from the waste products of existing reactors as well as from the plutonium used in weapons, as is the case in the operating, as of 2014, BN-800 reactor. Depleted uranium (DU) waste can also be used as fuel in fast reactors. 

The waste products of IFR reactors either have a short half-life, which means that they decay quickly and become relatively safe, or a long halflife, which means that they are only slightly radioactive. Due to pyroprocessing the total volume of true waste/fission products is 1/20th the volume of spent fuel produced by a light water plant of the same power output, and is often all considered to be waste. 70% of fission products are either stable or have half lives under one year. Technetium-99 and iodine-129, which constitute 6% of fission products, have very long half lives but can be transmuted to isotopes with very short half lives (15.46 seconds and 12.36 hours) by neutron absorption within a reactor, effectively destroying them (see more Long-lived fission products). Zirconium-93, another 5% of fission products, could in principle be recycled into fuel-pin cladding, where it does not matter that it is radioactive. Excluding the contribution from Transuranic waste (TRU) - which are isotopes produced when U-238 captures a slow thermal neutron in a LWR but does not fission, all the remaining high level waste/fission products("FP") left over from reprocessing out the TRU fuel, is less radiotoxic (in Sieverts) than natural uranium (in a gram to gram comparison) within 400 years, and it continues its decline following this.

Edwin Sayre has estimated that a ton of fission products(which also include the very weakly radioactive Palladium-107 etc.) reduced to metal, has a market value of $16 million.

The two forms of IFR waste produced, contain no plutonium or other actinides. The radioactivity of the waste decays to levels similar to the original ore in about 300–400 years.

The on-site reprocessing of fuel means that the volume of high level nuclear waste leaving the plant is tiny compared to LWR spent fuel. In fact, in the U.S. most spent LWR fuel has remained in storage at the reactor site instead of being transported for reprocessing or placement in a geological repository. The smaller volumes of high level waste from reprocessing could stay at reactor sites for some time, but are intensely radioactive from medium-lived fission products (MLFPs) and need to be stored securely, like in the present Dry cask storage vessels. In its first few decades of use, before the MLFP's decay to lower heat producing levels, geological repository capacity is constrained not by volume but by heat generation, and decay heat generation from medium-lived fission products is about the same per unit power from any kind of fission reactor, limiting early repository emplacement. 

The potential complete removal of plutonium from the waste stream of the reactor reduces the concern that presently exists with spent nuclear fuel from most other reactors that arises with burying or storing their spent fuel in a geological repository, as they could possibly be used as a plutonium mine at some future date. "Despite the million-fold reduction in radiotoxicity offered by this scheme, some believe that actinide removal would offer few if any significant advantages for disposal in a geologic repository because some of the fission product nuclides of greatest concern in scenarios such as groundwater leaching actually have longer half-lives than the radioactive actinides. These concerns do not consider the plan to store such materials in insoluble Synroc, and do not measure hazards in proportion to those from natural sources such as medical x-rays, cosmic rays, or natural radioactive rocks (such as granite). These persons are concerned with radioactive fission products such as technetium-99, iodine-129, and cesium-135 with half-lives between 213,000 and 15.7 million years" Some of which are being targeted for transmutation to belay even these comparatively low concerns, for example the IFR's positive void coefficient could be reduced to an acceptable level by adding technetium to the core, helping destroy the long-lived fission product technetium-99 by nuclear transmutation in the process.

Efficiency

IFRs use virtually all of the energy content in the uranium fuel whereas a traditional light water reactor uses less than 0.65% of the energy in mined uranium, and less than 5% of the energy in enriched uranium.

Carbon dioxide

Both IFRs and LWRs do not emit CO₂ during operation, although construction and fuel processing result in CO₂ emissions, if energy sources which are not carbon neutral (such as fossil fuels), or CO₂ emitting cements are used during the construction process. 

A 2012 Yale University review published in the Journal of Industrial Ecology analyzing CO₂ life cycle assessment emissions from nuclear power determined that:

The collective LCA literature indicates that life cycle GHG [ greenhouse gas ] emissions from nuclear power are only a fraction of traditional fossil sources and comparable to renewable technologies.

Although the paper primarily dealt with data from Generation II reactors, and did not analyze the CO₂ emissions by 2050 of the presently under construction Generation III reactors, it did summarize the Life Cycle Assessment findings of in development reactor technologies.

Theoretical FBRs [ Fast Breeder Reactors ] have been evaluated in the LCA literature. The limited literature that evaluates this potential future technology reports median life cycle GHG emissions... similar to or lower than LWRs[ light water reactors ] and purports to consume little or no uranium ore.

Actinides and fission products by half-life
Actinides by decay chain				Half-life range (y)	Fission products of 235U by yield
4n	4n+1	4n+2	4n+3
						4.5–7%	0.04–1.25%	<0 .001="" span="">
228Ra№				4–6	†		155Euþ
244Cmƒ	241Puƒ	250Cf	227Ac№	10–29		90Sr	85Kr	113mCdþ
232Uƒ		238Puƒ	243Cmƒ	29–97		137Cs	151Smþ	121mSn
248Bk[20]	249Cfƒ	242mAmƒ		141–351	No fission products have a half-life in the range of 100–210 k years ...
	241Amƒ		251Cfƒ	430–900
		226Ra№	247Bk	1.3 k – 1.6 k
240Pu	229Th	246Cmƒ	243Amƒ	4.7 k – 7.4 k
	245Cmƒ	250Cm		8.3 k – 8.5 k
			239Puƒ	24.1 k
		230Th№	231Pa№	32 k – 76 k
236Npƒ	233Uƒ	234U№		150 k – 250 k	‡	99Tc₡	126Sn
248Cm		242Pu		327 k – 375 k			79Se₡
				1.53 M		93Zr
	237Npƒ			2.1 M – 6.5 M		135Cs₡	107Pd
236U			247Cmƒ	15 M – 24 M			129I₡
244Pu№				80 M	... nor beyond 15.7 M years
232Th№		238U№	235Uƒ№	0.7 G – 14.1 G
Legend for superscript symbols ₡  has thermal neutron capture cross section in the range of 8–50 barns ƒ  fissile m  metastable isomer №  primarily a naturally occurring radioactive material (NORM) þ  neutron poison (thermal neutron capture cross section greater than 3k barns) †  range 4–97 y: Medium-lived fission product ‡  over 200,000 y: Long-lived fission product

Fuel cycle

Fast reactor fuel must be at least 20% fissile, greater than the low enriched uranium used in LWRs. The fissile material could initially include highly enriched uranium or plutonium, from LWR spent fuel, decommissioned nuclear weapons, or other sources. During operation the reactor breeds more fissile material from fertile material, at most about 5% more from uranium, and 1% more from thorium. 

The fertile material in fast reactor fuel can be depleted uranium (mostly U-238), natural uranium, thorium, or reprocessed uranium from spent fuel from traditional light water reactors, and even include nonfissile isotopes of plutonium and minor actinide isotopes. Assuming no leakage of actinides to the waste stream during reprocessing, a 1GWe IFR-style reactor would consume about 1 ton of fertile material per year and produce about 1 ton of fission products. 

The IFR fuel cycle's reprocessing by pyroprocessing (in this case, electrorefining) does not need to produce pure plutonium free of fission product radioactivity as the PUREX process is designed to do. The purpose of reprocessing in the IFR fuel cycle is simply to reduce the level of those fission products that are neutron poisons; even those need not be completely removed. The electrorefined spent fuel is highly radioactive, but because new fuel need not be precisely fabricated like LWR fuel pellets but can simply be cast, remote fabrication can be used, reducing exposure to workers. 

Like any fast reactor, by changing the material used in the blankets, the IFR can be operated over a spectrum from breeder to self-sufficient to burner. In breeder mode (using U-238 blankets) it will produce more fissile material than it consumes. This is useful for providing fissile material for starting up other plants. Using steel reflectors instead of U-238 blankets, the reactor operates in pure burner mode and is not a net creator of fissile material; on balance it will consume fissile and fertile material and, assuming loss-free reprocessing, output no actinides but only fission products and activation products. Amount of fissile material needed could be a limiting factor to very widespread deployment of fast reactors, if stocks of surplus weapons plutonium and LWR spent fuel plutonium are not sufficient. To maximize the rate at which fast reactors can be deployed, they can be operated in maximum breeding mode. 

Because the current cost of enriched uranium is low compared to the expected cost of large-scale pyroprocessing and electrorefining equipment and the cost of building a secondary coolant loop, the higher fuel costs of a thermal reactor over the expected operating lifetime of the plant are offset by increased capital cost. (Currently in the United States, utilities pay a flat rate of 1/10 of a cent per kilowatt hour to the Government for disposal of high level radioactive waste by law under the Nuclear Waste Policy Act. If this charge were based on the longevity of the waste, closed fuel cycles might become more financially competitive. As the planned geological repository in the form of Yucca Mountain is not going ahead, this fund has collected over the years and presently $25 billion has piled up on the Government's doorstep for something they have not delivered, that is, reducing the hazard posed by the waste.

Reprocessing nuclear fuel using pyroprocessing and electrorefining has not yet been demonstrated on a commercial scale, so investing in a large IFR-style plant may be a higher financial risk than a conventional light water reactor.

IFR concept (color), an animation of the pyroprocessing cycle is also available.

 

IFR concept (Black and White with clearer text)

Passive safety

The IFR uses metal alloy fuel (uranium/plutonium/zirconium) which is a good conductor of heat, unlike the LWR's (and even some fast breeder reactors') uranium oxide which is a poor conductor of heat and reaches high temperatures at the center of fuel pellets. The IFR also has a smaller volume of fuel, since the fissile material is diluted with fertile material by a ratio of 5 or less, compared to about 30 for LWR fuel. The IFR core requires more heat removal per core volume during operation than the LWR core; but on the other hand, after a shutdown, there is far less trapped heat that is still diffusing out and needs to be removed. However, decay heat generation from short-lived fission products and actinides is comparable in both cases, starting at a high level and decreasing with time elapsed after shutdown. The high volume of liquid sodium primary coolant in the pool configuration is designed to absorb decay heat without reaching fuel melting temperature. The primary sodium pumps are designed with flywheels so they will coast down slowly (90 seconds) if power is removed. This coast-down further aids core cooling upon shutdown. If the primary cooling loop were to be somehow suddenly stopped, or if the control rods were suddenly removed, the metal fuel can melt as accidentally demonstrated in EBR-I, however the melting fuel is then extruded up the steel fuel cladding tubes and out of the active core region leading to permanent reactor shutdown and no further fission heat generation or fuel melting. With metal fuel, the cladding is not breached and no radioactivity is released even in extreme overpower transients.

Self-regulation of the IFR's power level depends mainly on thermal expansion of the fuel which allows more neutrons to escape, damping the chain reaction. LWRs have less effect from thermal expansion of fuel (since much of the core is the neutron moderator) but have strong negative feedback from Doppler broadening (which acts on thermal and epithermal neutrons, not fast neutrons) and negative void coefficient from boiling of the water moderator/coolant; the less dense steam returns fewer and less-thermalized neutrons to the fuel, which are more likely to be captured by U-238 than induce fissions. However, the IFR's positive void coefficient could be reduced to an acceptable level by adding technetium to the core, helping destroy the long-lived fission product technetium-99 by nuclear transmutation in the process.

IFRs are able to withstand both a loss of flow without SCRAM and loss of heat sink without SCRAM. In addition to passive shutdown of the reactor, the convection current generated in the primary coolant system will prevent fuel damage (core meltdown). These capabilities were demonstrated in the EBR-II. The ultimate goal is that no radioactivity will be released under any circumstance.

The flammability of sodium is a risk to operators. Sodium burns easily in air, and will ignite spontaneously on contact with water. The use of an intermediate coolant loop between the reactor and the turbines minimizes the risk of a sodium fire in the reactor core.

Under neutron bombardment, sodium-24 is produced. This is highly radioactive, emitting an energetic gamma ray of 2.7 MeV followed by a beta decay to form magnesium-24. Half-life is only 15 hours, so this isotope is not a long-term hazard. Nevertheless, the presence of sodium-24 further necessitates the use of the intermediate coolant loop between the reactor and the turbines.

Proliferation

IFRs and Light water reactors (LWRs) both produce reactor grade plutonium, and even at high burnups remains weapons usable, but the IFR fuel cycle has some design features that would make proliferation more difficult than the current PUREX recycling of spent LWR fuel. For one thing, it may operate at higher burnups and therefore increase the relative abundance of the non-fissile, but fertile, isotopes Plutonium-238, Plutonium-240 and Plutonium-242.

Unlike PUREX reprocessing, the IFR's electrolytic reprocessing of spent fuel did not separate out pure plutonium, and left it mixed with minor actinides and some rare earth fission products which make the theoretical ability to make a bomb directly out of it considerably dubious. Rather than being transported from a large centralized reprocessing plant to reactors at other locations, as is common now in France, from La Hague to its dispersed nuclear fleet of LWRs, the IFR pyroprocessed fuel would be much more resistant to unauthorized diversion. The material with the mix of plutonium isotopes in an IFR would stay at the reactor site and then be burnt up practically in-situ, alternatively, if operated as a breeder reactor, some of the pyroprocessed fuel could be consumed by the same or other reactors located elsewhere. However, as is the case with conventional aqueous reprocessing, it would remain possible to chemically extract all the plutonium isotopes from the pyroprocessed/recycled fuel and would be much easier to do so from the recycled product than from the original spent fuel, although compared to other conventional recycled nuclear fuel, MOX, it would be more difficult, as the IFR recycled fuel contains more fission products than MOX and due to its higher burnup, more proliferation resistant Pu-240 than MOX. 

An advantage of the IFRs actinides removal and burn up (actinides include plutonium) from its spent fuel, is to eliminate concerns about leaving the IFRs spent fuel or indeed conventional, and therefore comparatively lower burnup, spent fuel - which can contain weapons usable plutonium isotope concentrations in a geological repository(or the more common dry cask storage) which then might be mined sometime in the future for the purpose of making weapons."

Because reactor-grade plutonium contains isotopes of plutonium with high spontaneous fission rates, and the ratios of these troublesome isotopes-from a weapons manufacturing point of view, only increases as the fuel is burnt up for longer and longer, it is considerably more difficult to produce fission nuclear weapons which will achieve a substantial yield from higher-burnup spent fuel than from conventional, moderately burnt up, LWR spent fuel. 

Therefore, proliferation risks are considerably reduced with the IFR system by many metrics, but not entirely eliminated. The plutonium from ALMR recycled fuel would have an isotopic composition similar to that obtained from other high burnt up spent nuclear fuel sources. Although this makes the material less attractive for weapons production, it could be used in weapons at varying degrees of sophistication/with fusion boosting. 

The U.S. government detonated a nuclear device in 1962 using then defined "reactor-grade plutonium", although in more recent categorizations it would instead be considered as fuel-grade plutonium, typical of that produced by low burn up magnox reactors.

Plutonium produced in the fuel of a breeder reactor generally has a higher fraction of the isotope plutonium-240, than that produced in other reactors, making it less attractive for weapons use, particularly in first generation nuclear weapon designs similar to Fat Man. This offers an intrinsic degree of proliferation resistance, but the plutonium made in the blanket of uranium surrounding the core, if such a blanket is used, is usually of a high Pu-239 quality, containing very little Pu-240, making it highly attractive for weapons use.

"Although some recent proposals for the future of the ALMR/IFR concept have focused more on its ability to transform and irreversibly use up plutonium, such as the conceptual PRISM (reactor) and the in operation(2014) BN-800 reactor in Russia, the developers of the IFR acknowledge that it is 'uncontested that the IFR can be configured as a net producer of plutonium'."

As mentioned above, if operated not as a burner, but as a breeder, the IFR has a clear proliferation potential "if instead of processing spent fuel, the ALMR system were used to reprocess irradiated fertile (breeding) material (that is if a blanket of breeding U-238 was used), the resulting plutonium would be a superior material, with a nearly ideal isotope composition for nuclear weapons manufacture."

Reactor design and construction

A commercial version of the IFR, S-PRISM, can be built in a factory and transported to the site. This small modular design (311 MWe modules) reduces costs and allows nuclear plants of various sizes (311 MWe and any integer multiple) to be economically constructed.

Cost assessments taking account of the complete life cycle show that fast reactors could be no more expensive than the most widely used reactors in the world – water-moderated water-cooled reactors.

Liquid metal Na coolant

Unlike reactors that use relatively slow low energy (thermal) neutrons, fast-neutron reactors need nuclear reactor coolant that does not moderate or block neutrons (like water does in an LWR) so that they have sufficient energy to fission actinide isotopes that are fissionable but not fissile. The core must also be compact and contain as small amount of material that might act as neutron moderators as possible. Metal sodium (Na) coolant in many ways has the most attractive combination of properties for this purpose. In addition to not being a neutron moderator, desirable physical characteristics include:

• Low melting temperature
• Low vapor pressure
• High boiling temperature
• Excellent thermal conductivity
• Low viscosity
• Light weight
• Thermal and radiation stability

Other benefits:  Abundant and low cost material. Cleaning with chlorine produces non-toxic table salt. Compatible with other materials used in the core (does not react or dissolve stainless steel) so no special corrosion protection measures needed. Low pumping power (from light weight and low viscosity). Maintains an oxygen (and water) free environment by reacting with trace amounts to make sodium oxide or sodium hydroxide and hydrogen, thereby protecting other components from corrosion. Light weight (low density) improves resistance to seismic inertia events (earthquakes.) 

Drawbacks include: Extreme fire hazard with any significant amounts of air (oxygen) and spontaneous combustion with water, rendering sodium leaks and flooding dangerous. This was the case at the Monju Nuclear Power Plant in a 1995 accident and fire. Reactions with water produce hydrogen which can be explosive. Sodium activation product (isotope) ²⁴Na releases dangerous energetic photons when it decays (however it has a very short half-life of 15 hours). Reactor design keeps ²⁴Na in the reactor pool and carries away heat for power production using a secondary sodium loop, adding costs to construction and maintenance.

History

Research on the reactor began in 1984 at Argonne National Laboratory in Argonne, Illinois. Argonne is a part of the U.S. Department of Energy's national laboratory system, and is operated on a contract by the University of Chicago. 

Argonne previously had a branch campus named "Argonne West" in Idaho Falls, Idaho that is now part of the Idaho National Laboratory. In the past, at the branch campus, physicists from Argonne had built what was known as the Experimental Breeder Reactor II (EBR II). In the mean time, physicists at Argonne had designed the IFR concept, and it was decided that the EBR II would be converted to an IFR. Charles Till, a Canadian physicist from Argonne, was the head of the IFR project, and Yoon Chang was the deputy head. Till was positioned in Idaho, while Chang was in Illinois.

With the election of President Bill Clinton in 1992, and the appointment of Hazel O'Leary as the Secretary of Energy, there was pressure from the top to cancel the IFR. Sen. John Kerry (D-MA) and O'Leary led the opposition to the reactor, arguing that it would be a threat to non-proliferation efforts, and that it was a continuation of the Clinch River Breeder Reactor Project that had been canceled by Congress.

Simultaneously, in 1994 Energy Secretary O'Leary awarded the lead IFR scientist with $10,000 and a gold medal, with the citation stating his work to develop IFR technology provided "improved safety, more efficient use of fuel and less radioactive waste."

IFR opponents also presented a report by the DOE's Office of Nuclear Safety regarding a former Argonne employee's allegations that Argonne had retaliated against him for raising concerns about safety, as well as about the quality of research done on the IFR program. The report received international attention, with a notable difference in the coverage it received from major scientific publications. The British journal Nature entitled its article "Report backs whistleblower", and also noted conflicts of interest on the part of a DOE panel that assessed IFR research. In contrast, the article that appeared in Science was entitled "Was Argonne Whistleblower Really Blowing Smoke?". Remarkably, that article did not disclose that the Director of Argonne National Laboratories, Alan Schriesheim, was a member of the Board of Directors of Science's parent organization, the American Association for the Advancement of Science.

Despite support for the reactor by then-Rep. Richard Durbin (D-IL) and U.S. Senators Carol Moseley Braun (D-IL) and Paul Simon (D-IL), funding for the reactor was slashed, and it was ultimately canceled in 1994 by S.Amdt. 2127 to H.R. 4506, at greater cost than finishing it. When this was brought to President Clinton's attention, he said "I know; it's a symbol." By this time Senator Kerry and the majority of democrats had switched to supporting the continuation of the program. The final count was to 52 to 46 to terminate the program, with 36 republicans and 16 democrats voting for its termination, while just 8 republicans and 38 democrats voted for its continuation.

In 2001, as part of the Generation IV roadmap, the DOE tasked a 242-person team of scientists from DOE, UC Berkeley, MIT, Stanford, ANL, LLNL, Toshiba, Westinghouse, Duke, EPRI, and other institutions to evaluate 19 of the best reactor designs on 27 different criteria. The IFR ranked #1 in their study which was released April 9, 2002.

At present there are no Integral Fast Reactors in commercial operation, however a very similar fast reactor, operated as a burner of plutonium stockpiles, the BN-800 reactor, became commercially operational in 2014.

Nuclear fusion–fission hybrid

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Hybrid nuclear fusion–fission (hybrid nuclear power) is a proposed means of generating power by use of a combination of nuclear fusion and fission processes. The basic idea is to use high-energy fast neutrons from a fusion reactor to trigger fission in otherwise nonfissile fuels like U-238 or Th-232. Each neutron can trigger several fission events, multiplying the energy released by each fusion reaction hundreds of times. This would not only make fusion designs more economical in power terms, but also be able to burn fuels that were not suitable for use in conventional fission plants, even their nuclear waste. 

 

The concept dates to the 1950s, and was strongly advocated by Hans Bethe during the 1970s. At that time the first powerful fusion experiments were being built, but it would still be many years before they could be economically competitive. Hybrids were proposed as a way of greatly accelerating their market introduction, producing energy even before the fusion systems reached break-even. However, detailed studies of the economics of the systems suggested they could not compete with existing fission reactors.

The idea was abandoned and lay dormant until the 2000s, when the continued delays in reaching break-even led to a brief revival around 2009, notably as the basis of the LIFE program. This program was cancelled when the underlying technology, from the National Ignition Facility, failed to reach its design performance goals. Apollo Fusion, a company founded by Google executive Mike Cassidy in 2017, was also reported to be focused on using the subcritical nuclear fusion-fission hybrid method.

In general terms, the hybrid is similar in concept to the fast breeder reactor, which uses a compact high-energy fission core in place of the hybrid's fusion core. Another similar concept is the accelerator-driven subcritical reactor, which uses a particle accelerator to provide the neutrons instead of nuclear reactions.

Fission basics

Conventional fission power plants rely on the chain reaction caused when nuclear fission events release neutrons that cause further fission events. Each fission event in uranium releases two or three neutrons, so by careful arrangement and the use of various absorber materials, you can balance the system so one of those neutrons causes another fission event while the other one or two are lost. This careful balance is known as criticality. 

Natural uranium is a mix of several isotopes, mainly a trace amount of U-235 and over 99% U-238. When they undergo fission, both of these elements release fast neutrons with an energy distribution peaking around 1 to 2 MeV. This energy is too low to cause fission in U-238, which means it cannot sustain a chain reaction. U-235 will undergo fission when struck by neutrons of this energy, so it is possible for U-235 to sustain a chain reaction, as is the case in a nuclear bomb. However, the probability of one neutron causing fission in another U-235 atom before it escapes the fuel is too low to maintain criticality in a mass of natural uranium, so the chain reaction can only occur in fuels with increased amounts of U-235. This is accomplished by concentrating, or enriching, the fuel, increasing the amount of U-235 to produce enriched uranium, while the leftover, now mostly U-238, is a waste product known as depleted uranium.

U-235 will undergo fission more easily if the neutrons are of lower energy, the so-called thermal neutrons. Neutrons can be slowed to thermal energies through collisions with a neutron moderator material, the easiest to use being the hydrogen atoms found in water. By placing the fission fuel in water, the probability that the neutrons will cause fission in another U-235 is greatly increased, which means the level of enrichment needed to reach criticality is greatly reduced. This leads to the concept of reactor-grade enriched uranium, with the amount of U-235 increased from just less than 1% to between 3 and 5% depending on the reactor design. This is in contrast to weapons-grade enrichment, which increases to the U-235 to at least 20%, and more commonly, over 90%.

In order to maintain criticality, the fuel has to retain that extra concentration of U-235. However, a typical fission reactor burns off enough of the U-235 to cause the reaction to stop over a period on the order of a few months. A combination of burnup of the U-235 along with the creation of neutron absorbers, or poisons, as part of the fission process eventually results in the fuel mass not being able to maintain criticality. This burned up fuel has to be removed and replaced with fresh fuel. The result is nuclear waste that is highly radioactive and filled with long lived radionuclides that present a safety concern. 

The waste contains most of the U-235 it started with, only 1% or so of the energy in the fuel is extracted by the time it reaches the point where it is no longer fissile. One solution to this problem is to reprocess the fuel, which uses chemical processes to separate the U-235 (and other non-poison elements) from the waste, and then uses that U-235 in fresh fuel loads. This reduces the amount of new fuel that needs to be mined, and also concentrates the unwanted portions of the waste into a smaller load. Reprocessing is expensive, however, and has generally been more expensive than simply buying fresh fuel from the mine. 

Another possibility is to breed Pu-239 from the U-238 through neutron capture, or various other means. In order to do this, higher energy neutrons are required, which means they cannot be moderated as in a conventional reactor. The simplest way to achieve this is to further enrich the original fuel well beyond what is needed for use in a moderated reactor, to the point where the U-235 maintains criticality even with the fast neutrons. The extra fast neutrons escaping the fuel load can then be used to breed fuel in a U-238 assembly surrounding the reactor core, most commonly taken from the stocks of depleted uranium. The Pu-239 is then chemically separated and mixed into fresh fuel for conventional reactors, in the same fashion as normal reprocessing, but the total volume of fuel created in this process is much greater. In spite of this, like reprocessing, the economics of breeder reactors has proven unattractive, and commercial breeder plants have ceased operation.

Fusion basics

Fusion reactors typically burn a mixture of deuterium (D) and tritium (T). When heated to millions of degrees, the kinetic energy in the fuel begins to overcome the natural electrostatic repulsion between nuclei, the so-called coulomb barrier, and the fuel begins to undergo fusion. This reaction gives off an alpha particle and a high energy neutron of 14 MeV. A key requirement to the economic operation of a fusion reactor is that the alphas deposit their energy back into the fuel mix, heating it so that additional fusion reactions take place. This leads to a condition not unlike the chain reaction in the fission case, known as ignition. 

Deuterium can be obtained by the separation of hydrogen isotopes in sea water (see heavy water production). Tritium has a short half life of just over a decade, so only trace amounts are found in nature. To fuel the reactor, the neutrons from the reaction are used to breed more tritium through a reaction in a blanket of lithium surrounding the reaction chamber. Tritium breeding is key to the success of a D-T fusion cycle, and to date this technique has not been demonstrated. Predictions based on computer modeling suggests that the breeding ratios are quite small and a fusion plant would barely be able to cover its own use. Many years would be needed to breed enough surplus to start another reactor.

Hybrid concepts

Fusion–fission designs essentially replace the lithium blanket with a blanket of fission fuel, either natural uranium ore or even nuclear waste. The fusion neutrons have more than enough energy to cause fission in the U-238, as well as many of the other elements in the fuel, including some of the transuranic waste elements. The reaction can continue even when all of the U-235 is burned off; the rate is controlled not by the neutrons from the fission events, but the neutrons being supplied by the fusion reactor. 

Fission occurs naturally because each event gives off more than one neutron capable of producing additional fission events. Fusion, at least in D-T fuel, gives off only a single neutron, and that neutron is not capable of producing more fusion events. When that neutron strikes fissile material in the blanket, one of two reactions may occur. In many cases, the kinetic energy of the neutron will cause one or two neutrons to be struck out of the nucleus without causing fission. These neutrons still have enough energy to cause other fission events. In other cases the neutron will be captured and cause fission, which will release two or three neutrons. This means that every fusion neutron in the fusion–fission design can result in anywhere between two and four neutrons in the fission fuel.

This is a key concept in the hybrid concept, known as fission multiplication. For every fusion event, several fission events may occur, each of which gives off much more energy than the original fusion, about 11 times. This greatly increases the total power output of the reactor. This has been suggested as a way to produce practical fusion reactors in spite of the fact that no fusion reactor has yet reached break-even, by multiplying the power output using cheap fuel or waste. However, a number of studies have repeatedly demonstrated that this only becomes practical when the overall reactor is very large, 2 to 3 GWt, which makes it expensive to build.

These processes also have the side-effect of breeding Pu-239 or U-233, which can be removed and used as fuel in conventional fission reactors. This leads to an alternate design where the primary purpose of the fusion–fission reactor is to reprocess waste into new fuel. Although far less economical than chemical reprocessing, this process also burns off some of the nastier elements instead of simply physically separating them out. This also has advantages for non-proliferation, as enrichment and reprocessing technologies are also associated with nuclear weapons production. However, the cost of the nuclear fuel produced is very high, and is unlikely to be able to compete with conventional sources.

Neutron economy

A key issue for the fusion–fission concept is the number and lifetime of the neutrons in the various processes, the so-called neutron economy. 

In a pure fusion design, the neutrons are used for breeding tritium in a lithium blanket. Natural lithium consists of about 92% Li-7 and the rest is mostly Li-6. Li-7 requires neutron energies even higher than those released by fission, around 5 MeV, well within the range of energies provided by fusion. This reaction produces Tritium and Helium-4, and another slow neutron. Li-6 can react with high or low energy neutrons, including those released by the Li-7 reaction. This means that a single fusion reaction can produce several tritiums, which is a requirement if the reactor is going to make up for natural decay and losses in the fusion processes. 

When the lithium blanket is replaced, or supplanted, by fission fuel in the hybrid design, neutrons that do react with the fissile material are no longer available for tritium breeding. The new neutrons released from the fission reactions can be used for this purpose, but only in Li-6. One could process the lithium to increase the amount of Li-6 in the blanket, making up for these losses, but the downside to this process is that the Li-6 reaction only produces one tritium atom. Only the high-energy reaction between the fusion neutron and Li-7 can create more than one tritium, and this is essential for keeping the reactor running. 

To address this issue, at least some of the fission neutrons must also be used for tritium breeding in Li-6. Every one that does is no longer available for fission, reducing the reactor output. This requires a very careful balance if one wants the reactor to be able to produce enough tritium to keep itself running, while also producing enough fission events to keep the fission side energy positive. If these cannot be accomplished simultaneously, there is no reason to build a hybrid. Even if this balance can be maintained, it might only occur at a level that is economically infeasible.

Overall economy

Through the early development of the hybrid concept the question of overall economics appeared difficult to handle. A series of studies starting in the late 1970s provided a much clearer picture of the hybrid in a complete fuel cycle, and allowed the economics to be better understood. These studies appeared to indicate there was no reason to build a hybrid.

One of the most detailed of these studies was published in 1980 by Los Alamos National Laboratory (LANL). Their study noted that the hybrid would produce most of its energy indirectly, both through the fission events in its own reactor, and much more by providing Pu-239 to fuel conventional fission reactors. In this overall picture, the hybrid is essentially identical to the breeder reactor, which uses fast neutrons from plutonium fission to breed more fuel in a fission blanket in largely the same fashion as the hybrid. Both require chemical processing to remove the bred Pu-239, both presented the same proliferation and safety risks as a result, and both produced about the same amount of fuel. Since that fuel is the primary source of energy in the overall cycle, the two systems were almost identical in the end.

What was not identical, however, was the technical maturity of the two designs. The hybrid would require considerable additional research and development before it would be known if it could even work, and even if that were demonstrated, the end result would be a system essentially identical to breeders which were already being built at that time. The report concluded:

The investment of time and money required to commercialize the hybrid cycle could only be justified by a real or perceived advantage of the hybrid over the classical FBR. Our analysis leads us to conclude that no such advantage exists. Therefore, there is not sufficient incentive to demonstrate and commercialize the fusion–fission hybrid.

Rationale

The fusion process alone currently does not achieve sufficient gain (power output over power input) to be viable as a power source. By using the excess neutrons from the fusion reaction to in turn cause a high-yield fission reaction (close to 100%) in the surrounding subcritical fissionable blanket, the net yield from the hybrid fusion–fission process can provide a targeted gain of 100 to 300 times the input energy (an increase by a factor of three or four over fusion alone). Even allowing for high inefficiencies on the input side (i.e. low laser efficiency in ICF and Bremsstrahlung losses in Tokamak designs), this can still yield sufficient heat output for economical electric power generation. This can be seen as a shortcut to viable fusion power until more efficient pure fusion technologies can be developed, or as an end in itself to generate power, and also consume existing stockpiles of nuclear fissionables and waste products. 

In the LIFE project at the Lawrence Livermore National Laboratory LLNL, using technology developed at the National Ignition Facility, the goal is to use fuel pellets of deuterium and tritium surrounded by a fissionable blanket to produce energy sufficiently greater than the input (laser) energy for electrical power generation. The principle involved is to induce inertial confinement fusion (ICF) in the fuel pellet which acts as a highly concentrated point source of neutrons which in turn converts and fissions the outer fissionable blanket. In parallel with the ICF approach, the University of Texas at Austin is developing a system based on the tokamak fusion reactor, optimising for nuclear waste disposal versus power generation. The principles behind using either ICF or tokamak reactors as a neutron source are essentially the same (the primary difference being that ICF is essentially a point-source of neutrons while Tokamaks are more diffuse toroidal sources).

Use to dispose of nuclear waste

The surrounding blanket can be a fissile material (enriched uranium or plutonium) or a fertile material (capable of conversion to a fissionable material by neutron bombardment) such as thorium, depleted uranium or spent nuclear fuel. Such subcritical reactors (which also include particle accelerator-driven neutron spallation systems) offer the only currently-known means of active disposal (versus storage) of spent nuclear fuel without reprocessing. Fission by-products produced by the operation of commercial light water nuclear reactors (LWRs) are long-lived and highly radioactive, but they can be consumed using the excess neutrons in the fusion reaction along with the fissionable components in the blanket, essentially destroying them by nuclear transmutation and producing a waste product which is far safer and less of a risk for nuclear proliferation. The waste would contain significantly reduced concentrations of long-lived, weapons-usable actinides per gigawatt-year of electric energy produced compared to the waste from a LWR. In addition, there would be about 20 times less waste per unit of electricity produced. This offers the potential to efficiently use the very large stockpiles of enriched fissile materials, depleted uranium, and spent nuclear fuel.

Safety

In contrast to current commercial fission reactors, hybrid reactors potentially demonstrate what is considered inherently safe behavior because they remain deeply subcritical under all conditions and decay heat removal is possible via passive mechanisms. The fission is driven by neutrons provided by fusion ignition events, and is consequently not self-sustaining. If the fusion process is deliberately shut off or the process is disrupted by a mechanical failure, the fission damps out and stops nearly instantly. This is in contrast to the forced damping in a conventional reactor by means of control rods which absorb neutrons to reduce the neutron flux below the critical, self-sustaining, level. The inherent danger of a conventional fission reactor is any situation leading to a positive feedback, runaway, chain reaction such as occurred during the Chernobyl disaster. In a hybrid configuration the fission and fusion reactions are decoupled, i.e. while the fusion neutron output drives the fission, the fission output has no effect whatsoever on the fusion reaction, completely eliminating any chance of a positive feedback loop.

Fuel cycle

There are three main components to the hybrid fusion fuel cycle: deuterium, tritium, and fissionable elements. Deuterium can be derived by separation of hydrogen isotopes in sea water. Tritium may be generated in the hybrid process itself by absorption of neutrons in lithium bearing compounds. This would entail an additional lithium bearing blanket and a means of collection. The third component is externally derived fissionable materials from demilitarized supplies of fissionables, or commercial nuclear fuel and waste streams. Fusion driven fission also offers the possibility of using Thorium as a fuel, which would greatly increase the potential amount of fissionables available. The extremely energetic nature of the fast neutrons emitted during the fusion events (up to 0.17 the speed of light) can allow normally non-fissioning U-238 to undergo fission directly (without conversion first to Pu-239), enabling refined natural Uranium to be used with very low enrichment, while still maintaining a deeply subcritical regime.

Engineering considerations

Practical engineering designs must first take into account safety as the primary goal. All designs should incorporate passive cooling in combination with refractory materials to prevent melting and reconfiguration of fissionables into geometries capable of un-intentional criticality. Blanket layers of Lithium bearing compounds will generally be included as part of the design to generate Tritium to allow the system to be self-supporting for one of the key fuel element components. Tritium, because of its relatively short half-life and extremely high radioactivity, is best generated on site to obviate the necessity of transportation from a remote location. D-T fuel can be manufactured on site using Deuterium derived from heavy water production and Tritium generated in the hybrid reactor itself. Nuclear spallation to generate additional neutrons can be used to enhance the fission output, with the caveat that this is a tradeoff between the number of neutrons (typically 20-30 neutrons per spallation event) against a reduction of the individual energy of each neutron. This is a consideration if the reactor is to use natural Thorium as a fuel. While high energy (0.17c) neutrons produced from fusion events are capable of directly causing fission in both Thorium and U-238, the lower energy neutrons produced by spallation generally cannot. This is a tradeoff which affects the mixture of fuels against the degree of spallation used in the design.