Neutron moderator

From Wikipedia, the free encyclopedia

 

In nuclear engineering, a neutron moderator is a medium that reduces the speed of fast neutrons, thereby turning them into thermal neutrons capable of sustaining a nuclear chain reaction involving uranium-235 or a similar fissile nuclide.

 

Commonly used moderators include regular (light) water (roughly 75% of the world's reactors), solid graphite (20% of reactors) and heavy water (5% of reactors). Beryllium has also been used in some experimental types, and hydrocarbons have been suggested as another possibility.

Moderation

Neutrons are normally bound into an atomic nucleus, and do not exist free for long in nature. The unbound neutron has a half-life of 10 minutes and 11 seconds. The release of neutrons from the nucleus requires exceeding the binding energy of the neutron, which is typically 7-9 MeV for most isotopes. Neutron sources generate free neutrons by a variety of nuclear reactions, including nuclear fission and nuclear fusion. Whatever the source of neutrons, they are released with energies of several MeV. 

According to the equipartition theorem, the average kinetic energy, ${\displaystyle {\bar {E}}}$, can be related to temperature, ${\displaystyle T}$,via:

${\displaystyle {\bar {E}}={\frac {1}{2}}m_{n}\langle v^{2}\rangle ={\frac {3}{2}}k_{B}T}$,

where ${\displaystyle m_{n}}$ is the neutron mass, ${\displaystyle v}$ is the average neutron speed, and ${\displaystyle k_{B}}$ is the Boltzmann constant. The characteristic neutron temperature of several-MeV neutrons is several tens of billions kelvin. 

Moderation is the process of the reduction of the initial high speed (high kinetic energy) of the free neutron. Since energy is conserved, this reduction of the neutron speed takes place by transfer of energy to a material called a moderator. 

The probability of scattering of a neutron from a nucleus is given by the scattering cross section. The first couple of collisions with the moderator may be of sufficiently high energy to excite the nucleus of the moderator. Such a collision is inelastic, since some of the kinetic energy is transformed to potential energy by exciting some of the internal degrees of freedom of the nucleus to form an excited state. As the energy of the neutron is lowered, the collisions become predominantly elastic, i.e., the total kinetic energy and momentum of the system (that of the neutron and the nucleus) is conserved. 

Given the mathematics of elastic collisions, as neutrons are very light compared to most nuclei, the most efficient way of removing kinetic energy from the neutron is by choosing a moderating nucleus that has near identical mass. 

Elastic collision of equal masses

A collision of a neutron, which has mass of 1, with a ¹H nucleus (a proton) could result in the neutron losing virtually all of its energy in a single head-on collision. More generally, it is necessary to take into account both glancing and head-on collisions. The mean logarithmic reduction of neutron energy per collision, ${\displaystyle \xi }$, depends only on the atomic mass, ${\displaystyle A}$, of the nucleus and is given by: 

${\displaystyle \xi =\ln {\frac {E_{0}}{E}}=1+{\frac {(A-1)^{2}}{2A}}\ln \left({\frac {A-1}{A+1}}\right)}$.

This can be reasonably approximated to the very simple form ${\displaystyle \xi \simeq {\frac {2}{A+1}}}$. From this one can deduce ${\displaystyle n}$, the expected number of collisions of the neutron with nuclei of a given type that is required to reduce the kinetic energy of a neutron from ${\displaystyle E_{0}}$ to ${\displaystyle E_{1}}$

${\displaystyle n={\frac {1}{\xi }}(\ln E_{0}-\ln E_{1})}$.

In a system at thermal equilibrium, neutrons (red) are elastically scattered by a hypothetical moderator of free hydrogen nuclei (blue), undergoing thermally activated motion. Kinetic energy is transferred between particles. As the neutrons have essentially the same mass as protons and there is no absorption, the velocity distributions of both particles types would be well-described by a single Maxwell–Boltzmann distribution.

Choice of moderator materials

Some nuclei have larger absorption cross sections than others, which removes free neutrons from the flux. Therefore, a further criterion for an efficient moderator is one for which this parameter is small. The moderating efficiency gives the ratio of the macroscopic cross sections of scattering, ${\displaystyle \Sigma _{s}}$, weighted by ${\displaystyle \xi }$ divided by that of absorption, ${\displaystyle \Sigma _{a}}$: i.e., ${\displaystyle {\frac {\xi \Sigma _{s}}{\Sigma _{a}}}}$. For a compound moderator composed of more than one element, such as light or heavy water, it is necessary to take into account the moderating and absorbing effect of both the hydrogen isotope and oxygen atom to calculate ${\displaystyle \xi }$. To bring a neutron from the fission energy of ${\displaystyle E_{0}}$ 2 MeV to an ${\displaystyle E}$ of 1 eV takes an expected ${\displaystyle n}$ of 16 and 29 collisions for H₂O and D₂O, respectively. Therefore, neutrons are more rapidly moderated by light water, as H has a far higher ${\displaystyle \Sigma _{s}}$. However, it also has a far higher ${\displaystyle \Sigma _{a}}$, so that the moderating efficiency is nearly 80 times higher for heavy water than for light water.

The ideal moderator is of low mass, high scattering cross section, and low absorption cross section.

	Hydrogen	Deuterium	Beryllium	Carbon	Oxygen	Uranium
Mass of kernels	1	2	9	12	16	238
Energy decrement ${\displaystyle \xi }$	1	0.7261	0.2078	0.1589	0.1209	0.0084
Number of Collisions	18	25	86	114	150	2172

Distribution of neutron velocities once moderated

After sufficient impacts, the speed of the neutron will be comparable to the speed of the nuclei given by thermal motion; this neutron is then called a thermal neutron, and the process may also be termed thermalization. Once at equilibrium at a given temperature the distribution of speeds (energies) expected of rigid spheres scattering elastically is given by the Maxwell–Boltzmann distribution. This is only slightly modified in a real moderator due to the speed (energy) dependence of the absorption cross-section of most materials, so that low-speed neutrons are preferentially absorbed, so that the true neutron velocity distribution in the core would be slightly hotter than predicted.

Reactor moderators

In a thermal-neutron reactor, the nucleus of a heavy fuel element such as uranium absorbs a slow-moving free neutron, becomes unstable, and then splits ("fissions") into two smaller atoms ("fission products"). The fission process for ²³⁵U nuclei yields two fission products, two to three fast-moving free neutrons, plus an amount of energy primarily manifested in the kinetic energy of the recoiling fission products. The free neutrons are emitted with a kinetic energy of ~2 MeV each. Because more free neutrons are released from a uranium fission event than thermal neutrons are required to initiate the event, the reaction can become self-sustaining — a chain reaction — under controlled conditions, thus liberating a tremendous amount of energy.

Fission cross section, measured in barns (a unit equal to 10⁻²⁸ m²), is a function of the energy (so-called excitation function) of the neutron colliding with a ²³⁵U nucleus. Fission probability decreases as neutron energy (and speed) increases. This explains why most reactors fueled with ²³⁵U need a moderator to sustain a chain reaction and why removing a moderator can shut down a reactor.

 

The probability of further fission events is determined by the fission cross section, which is dependent upon the speed (energy) of the incident neutrons. For thermal reactors, high-energy neutrons in the MeV-range are much less likely to cause further fission. (Note: It is not impossible for fast neutrons to cause fission, just much less likely.) The newly released fast neutrons, moving at roughly 10% of the speed of light, must be slowed down or "moderated," typically to speeds of a few kilometres per second, if they are to be likely to cause further fission in neighbouring ²³⁵U nuclei and hence continue the chain reaction. This speed happens to be equivalent to temperatures in the few hundred Celsius range. 

In all moderated reactors, some neutrons of all energy levels will produce fission, including fast neutrons. Some reactors are more fully thermalised than others; for example, in a CANDU reactor nearly all fission reactions are produced by thermal neutrons, while in a pressurized water reactor (PWR) a considerable portion of the fissions are produced by higher-energy neutrons. In the proposed water-cooled supercritical water reactor (SCWR), the proportion of fast fissions may exceed 50%, making it technically a fast neutron reactor. 

A fast reactor uses no moderator, but relies on fission produced by unmoderated fast neutrons to sustain the chain reaction. In some fast reactor designs, up to 20% of fissions can come from direct fast neutron fission of uranium-238, an isotope which is not fissile at all with thermal neutrons. 

Moderators are also used in non-reactor neutron sources, such as plutonium-beryllium and spallation sources.

Form and location

The form and location of the moderator can greatly influence the cost and safety of a reactor. Classically, moderators were precision-machined blocks of high purity graphite with embedded ducting to carry away heat. They were in the hottest part of the reactor, and therefore subject to corrosion and ablation. In some materials, including graphite, the impact of the neutrons with the moderator can cause the moderator to accumulate dangerous amounts of Wigner energy. This problem led to the infamous Windscale fire at the Windscale Piles, a nuclear reactor complex in the United Kingdom, in 1957. 

Some pebble-bed reactors' moderators are not only simple, but also inexpensive: the nuclear fuel is embedded in spheres of reactor-grade pyrolytic carbon, roughly of the size of tennis balls. The spaces between the balls serve as ducting. The reactor is operated above the Wigner annealing temperature so that the graphite does not accumulate dangerous amounts of Wigner energy. 

In CANDU and PWR reactors, the moderator is liquid water (heavy water for CANDU, light water for PWR). In the event of a loss-of-coolant accident in a PWR, the moderator is also lost and the reaction will stop. This negative void coefficient is an important safety feature of these reactors. In CANDU the moderator is located in a separate heavy-water circuit, surrounding the pressurized heavy-water coolant channels. This design gives CANDU reactors a positive void coefficient, although the slower neutron kinetics of heavy-water moderated systems compensates for this, leading to comparable safety with PWRs."

Moderator impurities

Good moderators are also free of neutron-absorbing impurities such as boron. In commercial nuclear power plants the moderator typically contains dissolved boron. The boron concentration of the reactor coolant can be changed by the operators by adding boric acid or by diluting with water to manipulate reactor power. The Nazi Nuclear Program suffered a substantial setback when its inexpensive graphite moderators failed to work. At that time, most graphites were deposited on boron electrodes, and the German commercial graphite contained too much boron. Since the war-time German program never discovered this problem, they were forced to use far more expensive heavy water moderators. In the U.S., Leó Szilárd, a former chemical engineer, discovered the problem.

Non-graphite moderators

Some moderators are quite expensive, for example beryllium, and reactor-grade heavy water. Reactor-grade heavy water must be 99.75% pure to enable reactions with unenriched uranium. This is difficult to prepare because heavy water and regular water form the same chemical bonds in almost the same ways, at only slightly different speeds.

The much cheaper light water moderator (essentially very pure regular water) absorbs too many neutrons to be used with unenriched natural uranium, and therefore uranium enrichment or nuclear reprocessing becomes necessary to operate such reactors, increasing overall costs. Both enrichment and reprocessing are expensive and technologically challenging processes, and additionally both enrichment and several types of reprocessing can be used to create weapons-usable material, causing proliferation concerns. Reprocessing schemes that are more resistant to proliferation are currently under development.

The CANDU reactor's moderator doubles as a safety feature. A large tank of low-temperature, low-pressure heavy water moderates the neutrons and also acts as a heat sink in extreme loss-of-coolant accident conditions. It is separated from the fuel rods that actually generate the heat. Heavy water is very effective at slowing down (moderating) neutrons, giving CANDU reactors their important and defining characteristic of high "neutron economy".

Nuclear weapon design

Early speculation about nuclear weapons assumed that an "atom bomb" would be a large amount of fissile material, moderated by a neutron moderator, similar in structure to a nuclear reactor or "pile". Only the Manhattan project embraced the idea of a chain reaction of fast neutrons in pure metallic uranium or plutonium. Other moderated designs were also considered by the Americans; proposals included using uranium deuteride as the fissile material. In 1943 Robert Oppenheimer and Niels Bohr considered the possibility of using a "pile" as a weapon. The motivation was that with a graphite moderator it would be possible to achieve the chain reaction without the use of any isotope separation. In August 1945, when information of the atomic bombing of Hiroshima was relayed to the scientists of the German nuclear program, interred at Farm Hall in England, chief scientist Werner Heisenberg hypothesized that the device must have been "something like a nuclear reactor, with the neutrons slowed by many collisions with a moderator".

After the success of the Manhattan project, all major nuclear weapons programs have relied on fast neutrons in their weapons designs. The notable exception is the Ruth and Ray test explosions of Operation Upshot–Knothole. The aim of the University of California Radiation Laboratory designs was the exploration of deuterated polyethylene charge containing uranium as a candidate thermonuclear fuel, hoping that deuterium would fuse (becoming an active medium) if compressed appropriately. If successful, the devices could also lead to a compact primary containing minimal amount of fissile material, and powerful enough to ignite RAMROD a thermonuclear weapon designed by UCRL at the time. For a "hydride" primary, the degree of compression would not make deuterium to fuse, but the design could be subjected to boosting, raising the yield considerably. The cores consisted of a mix of uranium deuteride (UD₃), and deuterated polyethylene. The core tested in Ray used uranium low enriched in U²³⁵, and in both shots deuterium acted as the neutron moderator. The predicted yield was 1.5 to 3 kt for Ruth (with a maximum potential yield of 20 kt) and 0.5-1 kt for Ray. The tests produced yields of 200 tons of TNT each; both tests were considered to be fizzles.

The main benefit of using a moderator in a nuclear explosive is that the amount of fissile material needed to reach criticality may be greatly reduced. Slowing of fast neutrons will increase the cross section for neutron absorption, reducing the critical mass. A side effect is however that as the chain reaction progresses, the moderator will be heated, thus losing its ability to cool the neutrons. 

Another effect of moderation is that the time between subsequent neutron generations is increased, slowing down the reaction. This makes the containment of the explosion a problem; the inertia that is used to confine implosion type bombs will not be able to confine the reaction. The end result may be a fizzle instead of a bang. 

The explosive power of a fully moderated explosion is thus limited, at worst it may be equal to a chemical explosive of similar mass. Again quoting Heisenberg: "One can never make an explosive with slow neutrons, not even with the heavy water machine, as then the neutrons only go with thermal speed, with the result that the reaction is so slow that the thing explodes sooner, before the reaction is complete."

While a nuclear bomb working on thermal neutrons may be impractical, modern weapons designs may still benefit from some level of moderation. A beryllium tamper used as a neutron reflector will also act as a moderator.

Materials used

• Hydrogen, as in ordinary "light water". Because protium also has a significant cross section for neutron capture only limited moderation is possible without losing too many neutrons. The less-moderated neutrons are relatively more likely to be captured by uranium-238 and less likely to fission uranium-235, so light water reactors require enriched uranium to operate.
  • There are also proposals to use the compound formed by the chemical reaction of metallic uranium and hydrogen (uranium hydride—UH₃) as a combination fuel and moderator in a new type of reactor.
  • Hydrogen is also used in the form of cryogenic liquid methane and sometimes liquid hydrogen as a cold neutron source in some research reactors: yielding a Maxwell–Boltzmann distribution for the neutrons whose maximum is shifted to much lower energies.
  • Hydrogen combined with carbon as in paraffin wax was used in some early German experiments.
• Deuterium, in the form of heavy water, in heavy water reactors, e.g. CANDU. Reactors moderated with heavy water can use unenriched natural uranium.
• Carbon, in the form of reactor-grade graphite or pyrolytic carbon, used in e.g. RBMK and pebble-bed reactors, or in compounds, e.g. carbon dioxide . Lower-temperature reactors are susceptible to buildup of Wigner energy in the material. Like deuterium-moderated reactors, some of these reactors can use unenriched natural uranium.
  • Graphite is also deliberately allowed to be heated to around 2000 K or higher in some research reactors to produce a hot neutron source: giving a Maxwell–Boltzmann distribution whose maximum is spread out to generate higher energy neutrons.
• Beryllium, in the form of metal. Beryllium is expensive and toxic, so its use is limited.
• Lithium-7, in the form of a lithium fluoride salt, typically in conjunction with beryllium fluoride salt (FLiBe). This is the most common type of moderator in a Molten Salt Reactor.

Other light-nuclei materials are unsuitable for various reasons. Helium is a gas and it requires special design to achieve sufficient density; lithium-6 and boron-10 absorb neutrons.

Currently operating nuclear power reactors by moderator

Moderator	Reactors	Design	Country
none (fast)	1	BN-600	Russia (1)
graphite	25	AGR, Magnox, RBMK	United Kingdom (14), Russia (11)
heavy water	29	CANDU PHWR	Canada (17), South Korea (4), Romania (2), China (2), India (18), Argentina, Pakistan
light water	359	PWR, BWR	27 countries

Integral fast reactor

From Wikipedia, the free encyclopedia

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.