- 1 Oxide fuel
- 2 Metal fuel
- 3 Less common chemical forms
- 4 Common physical forms of nuclear fuel
- 5 Less common fuel forms
- 6 Spent nuclear fuel
- 7 Post Irradiation Examination (PIE) and fuel behavior
- 8 Radioisotope decay fuels
- 9 Fusion fuels
- 10 References
- See also: Nuclear fuel cycle
Nuclear fuel is any material that can be consumed to derive nuclear energy, by analogy to chemical fuel that is burned to derive energy. By far the most common type of nuclear fuel is isotopes of elements of high atomic number, that can be made to undergo nuclear fission chain reactions in a nuclear fission reactor; nuclear fuel can refer to the material or to physical objects (for example fuel bundles composed of fuel rods) composed of the fuel material, perhaps mixed with structural, neutron moderating, or neutron reflecting materials. The most common fissile nuclear fuels are uranium (U) and plutonium (Pu), and the actions of mining, refining, purifying, using, and ultimately disposing of these elements together make up the nuclear fuel cycle, which is important for its relevance to nuclear power generation and nuclear weapons.
Not all nuclear fuels are used in fission chain reactions. For example, 238Pu, an isotope not useful in and some other elements directly small amounts of heat fromradioactive decay in radiothermal generators, and other atomic batteries. Light isotopes such as 3H (tritium) are used as fuel for nuclear fusion. If one looks at binding energy of specific isotopes, there can be an energy gain from fusing most elements with a lower atomic number than iron, and fissioning isotopes with a higher atomic number than iron.
Planning for producing nuclear fuel is a necessary prerequisite to make a nuclear weapon, unless weapons-grade fuel is made available by a third party. Worldwide management and security of nuclear fuel production, as well as extremely attention to fuel types that are not dual-use and only have weapons application, is a key part of nonproliferation policy. Under international agreements, it should be monitored by the International Atomic Energy Agency.
Some specialized components of nuclear weapons, such as the initiator are radioactive, but not usually considered fuel. Nevertheless, these are even more important in a nonproliferation regime, as they often have only military applications.
The thermal conductivity of uranium dioxide is low, it is affected by porosity and burn-up. The burn-up results in fission products being dissolved in the lattice (such as lanthanides), the precipitation of fission products such as palladium, the formation of fission gas bubbles due to fission products such as xenon and krypton and radiation damage of the lattice. The low thermal conductivity can lead to overheating of the centre part of the pellets during use. The porosity results in a decrease in both the thermal conductivity of the fuel and the swelling which occurs during use.
The bulk density of the fuel can be related to the thermal conductivity
- p = (ρtd-ρ)/ρ
Where ρ is the bulk density of the fuel and ρtd is the theoretical density of the uranium dioxide.
Then the thermal conductivity of the porous phase (Kf)is related to the conductivity of the perfect phase (Ko, no porosity) by the following equation. Note that s is a term for the shape factor of the holes.
- Kf = Ko.(1-p/1+(s-1)p)
Rather than measuring the thermal conductivity using the traditional methods in physics such as Lee's disk, the Forbes' method or Searle's bar it is common to use a laser flash method where a small disc of fuel is placed in a furnace. After being heated to the required temperature one side of the disc is illuminated with a laser pulse, the time required for the heat wave to flow through the disc, the density of the disc, and the thickness of the disk can then be used to calculated to give the thermal conductivity.
- λ = ρCpα
If t1/2 is defined as the time required for the non illuminated surface to experience half its final temperature rise then.
- α = 0.1388 L2 / t1/2
L is the thickness of the disc
For details see 
Uranium dioxide is a black semiconductor solid. It can be made by reacting uranyl nitrate with a base (ammonia) to form a solid (ammonium uranate). It is heated (calcined) to form U3O8 that can than be converted by heating in an argon / hydrogen mixture (700 oC) to form UO2. The UO2 is then mixed with an organic binder and pressed into pellets, these pellets are then fired at a much higher temperature (in H2/Ar) to sinter the solid. The aim is to form a dense solid which has few pores.
The thermal conductivity of uranium dioxide is very low compared with that of zirconium metal, and it goes down as the temperature goes up.
Mixed oxide, or MOX fuel, is a blend of plutonium and natural or depleted uranium which behaves similarly (though not identically) to the enriched uranium feed for which most nuclear reactors were designed. MOX fuel is an alternative to low enriched uranium (LEU) fuel used in the light water reactors which predominate nuclear power generation.
Some concern has been expressed that used MOX cores will introduce new disposal challenges, though MOX is itself a means to dispose of surplus plutonium by transmutation.
Currently (March, 2005) reprocessing of commercial nuclear fuel to make MOX is done in England and France, and to a lesser extent in Russia, India and Japan. China plans to develop fast breeder reactors and reprocessing.
The Global Nuclear Energy Partnership, is a U.S. plan to form an international partnership to see spent nuclear fuel reprocessed in a way that renders the plutonium in it usable for nuclear fuel but not for nuclear weapons. Reprocessing of spent commercial-reactor nuclear fuel has not been permitted in the United States due to nonproliferation considerations. All of the other reprocessing nations have long had nuclear weapons from military-focused "research"-reactor fuels except for Japan.
Metal fuels have the advantage of a much higher heat conductivity than oxide fuels but cannot survive equally high temperatures.
TRIGA fuel is used in TRIGA (Training, Research, Isotopes, General Atomics) reactors. TRIGA fuel consists of a uranium zirconium hydride matrix. It is inherently safe in that if it reaches a high temperature, the hydrogen's cross section in the fuel is shifted to higher energies, allowing more neutrons to be lost, and less to be thermalized. Most cores that use this fuel are "high leakage" cores where the excess leaked neutrons can be utilized for research.
In a fast neutron reactor the minor actinides produced by neutron capture of uranium and plutonium can be used as fuel. Metal actinide fuel is typically an alloy of zirconium , uranium, plutonium and the minor actinides. It can be made inherently safe as thermal expansion of the metal alloy will increase neutron leakage.
Less common chemical forms
Ceramic fuels have the advantage of a high heat conductivities and melting points, but they are more prone to swelling than oxide fuels and are much less well understood.
This is often the fuel of choice for reactor designs that NASA produces, one advantage is that UN has a better thermal conductivity than UO2. Uranium nitride has a very high melting point. This fuel has the disadvantage that unless 15N was used (in place of the more common 14N) that a large amount of 14C would be generated from the nitrogen by the pn reaction. As the nitrogen required for such a fuel would be so expensive it is likely that the fuel would have to be reprocessed by a pyro method to enable to the 15N to be recovered. It is likely that if the fuel was processed and dissolved in nitric acid that the nitrogen enriched with 15N would be diluted with the common 14N.
Much of what is known about uranium carbide is in the form of pin-type fuel elements for liquid metal fast breeder reactors during their intense study during the 60's and 70's. However, recently there has been a revived interest in uranium carbide in the form of plate fuel and most notably, micro fuel particles (such as TRISO particles).
The high thermal conductivity and high melting point make uranium carbide an attractive fuel. In addition, because of the absence of oxygen in this fuel (during the course of radiation, excess gas pressure can build from the formation O2 or other gases) as well as the ability to compliment a ceramic coating (a ceramic-ceramic interface has structural and chemical advantages), uranium carbide could be the ideal fuel candidate for certain Generation IV reactors such as the gas-cooled fast reactor (GFR).
These include fuels where the fuel is dissolved in the coolant. They were used in the molten salt reactor experiment and numerous other liquid core reactor experiments. The liquid fuel for the molten salt reactor was LiF-BeF2-ThF4-UF4 (72-16-12-0.4 mol%), it had a peak operating temperature of 705 °C in the experiment but could have gone to much higher temperatures since the boiling point of the molten salt was in excess of 1400 °C.
The Aqueous Homogeneous Reactors uses a solution of uranyl sulfate or other uranium salt in water. This homogenous reactor type has not been used for any large power reactors. One of its disadvantages is that the fuel is in a form which is easy to disperse in the event of an accident.
Common physical forms of nuclear fuel
For use as nuclear fuel, enriched UF6 is converted into uranium dioxide (UO2) powder that is then processed into pellet form. The pellets are then fired in a high-temperature, sintering furnace to create hard, ceramic pellets of enriched uranium. The cylindrical pellets then undergo a grinding process to achieve a uniform pellet size. The pellets are stacked, according to each nuclear core's design specifications, into tubes of corrosion-resistant metal alloy. The tubes are sealed to contain the fuel pellets: these tubes are called fuel rods. The finished fuel rods are grouped in special fuel assemblies that are then used to build up the nuclear fuel core of a power reactor.
The metal used for the tubes depends on the design of the reactor - stainless steel was used in the past, but most reactors now use a zirconium alloy. For the most common types of reactors (BWRs and PWRs) the tubes are assembled into bundles with the tubes spaced precise distances apart. These bundles are then given a unique identification number, which enables them to be tracked from manufacture through use and into disposal
Pressurized water reactor (PWR) fuel consists of cylindrical rods put into bundles. A uranium oxide ceramic is formed into pellets and inserted into Zircaloy tubes that are bundled together. The Zircaloy tubes are about 1 cm in diameter, and the fuel cladding gap is filled with helium gas to improve the conduction of heat from the fuel to the cladding. There are about 179-264 fuel rods per fuel bundle and about 121 to 193 fuel bundles are loaded into a reactor core. Generally, the fuel bundles consist of fuel rods bundled 14x14 to 17x17. PWR fuel bundles are about 4 meters in length. In PWR fuel bundles, control rods are inserted through the top directly into the fuel bundle. The fuel bundles usually are enriched several percent in 235U. The uranium oxide is dried before inserting into the tubes to try to eliminate moisture in the ceramic fuel that can lead to corrosion and hydrogen embrittlement. The Zircoloy tubes are pressurized with helium to try to minimize pellet cladding interaction (PCI) which can lead to fuel rod failure over long periods.
In boiling water reactors (BWR), the fuel is similar to PWR fuel except that the bundles are "canned". That is there is a thin tube surrounding each bundle. This is primarily done to prevent local density variations from effecting neutronics and thermal hydraulics of the nuclear core on a global scale. In BWR fuel bundles, there are about 500-800 fuel rods per assembly. Each BWR fuel rod is back filled with helium to a pressure of about three atmospheres (300 kPa).
CANDU fuel bundles are about a half meter in length and 30 cm in diameter. They consist of sintered (UO2) pellets in Zirconium tubes, welded to Zirconium end plates. Each bundle is roughly 20 kg, and a typical core loading is on the order of 4500 bundles. Modern types typically have 37 identical fuel pins radially arranged about the long axis of the bundle, but in the past several different configurations and numbers of pins have been used. The CANFLEX bundle has 43 fuel elements, with two element sizes. It is about 10 cm (four inches) in diameter, 0.5 m (20 inches long) and weighs about 20 kg (44 lbs) and replaces 37-pin standard bundle. It has been designed specifically to increase fuel performance by utilizing two different pin diameters. Current CANDU designs do not need enriched uranium to achieve criticality (due to their more efficient heavy water moderator), however, some newer concepts call for low enrichment to help reduce the size of the reactors.
Less common fuel forms
Various other nuclear fuel forms find use in specific applications, but lack the widespread use of those found in BWRs, PWRs, and CANDU power plants. Many of these fuel forms are only found in research reactors, or have military applications.
Tristructural-isotropic (TRISO) fuel is a type of micro fuel particle. It consists of a fuel kernel composed of UOX (sometimes UC or UCO) in the center, coated with four layers of three isotropic materials. The four layers are a porous buffer layer made of carbon, followed by a dense inner layer of pyrolytic carbon (PyC), followed by a ceramic layer of SiC to retain fission products at elevated temperatures and to give the TRISO particle more structural integrity, followed by a dense outer layer of PyC. TRISO fuel particles are designed to not crack due to the stresses from processes (such as differential thermal expansion or fission gas pressure) at temperatures beyond 1600°C, and therefore can contain the fuel in the worst of accident scenarios in a properly designed reactor. Two such reactor designs are pebble bed modular reactor (PBMR), in which thousands of TRISO fuel particles are dispersed into graphite pebbles, and a prismatic-block gas cooled reactor (such as the GT-MHR), in which the the TRISO fuel particles are fabricated into compacts and placed in a graphite block matrix. Both of these reactor designs are high-temperature gas-cooled reactors (HTGR), which is a type of very high temperature reactors (VHTR), one of the six classes of reactor designs in the Generation IV initiative.
TRISO fuel particles were originally developed in Germany for high-temperature gas-cooled reactors (HTGR). The first nuclear reactor to use TRISO fuels was the AVR and the first powerplant was the THTR-300. Currently, TRISO fuel compacts are being used in the experimental reactors, the HTR-10 in China, and the HTTR in Japan.
RBMK reactor fuel was used in soviet designed and built RBMK type reactors. This is a low enriched uranium oxide fuel. The fuel elements in an RBMK are extremely long, on the order of 7 meters. The Chernobyl reactor was a 1GWe RBMK reactor.
CerMet fuel consists of ceramic fuel particles (usually uranium oxide) embedded in a metal matrix. It is hypothesized that this type of fuel is what is used in US Navy reactors. This fuel has high heat transport characteristics and can withstand a large amount of expansion.
Plate type fuel
Plate type fuel has grown out of favor over the years. It is currently used in the Advanced Test Reactor (ATR) at Idaho National Laboratory.
Spent nuclear fuel
Used nuclear fuel is a complex mixture of the fission products, uranium, plutonium and the transplutonium metals. In fuel which has been used at high temperature in power reactors it is common for the fuel to not be homogenous often the fuel will contain nanoparticles of platinum group metals such as palladium. Also the fuel may well have cracked, swelled and been used close to its melting point. Despite the fact that the used fuel can be cracked it is very insoluble in water, and is able to retain the vast majority of the actinides and fission products within the uranium dioxide crystal lattice.
Oxide fuel under accident conditions
Two main modes of release exist, the fission products can be vapourised or small particles of the fuel can be dispersed. For details see the main article at Nuclear fuel and reactor accidents.
Post Irradiation Examination (PIE) and fuel behavior
It is common that experimental and production fuel will be examined after use in a reactor.    Due to the intensely radioactive nature of the used fuel this is done in a hot cell. A combination of nondestructive and destructive methods are used.
The PIE is used to check that the fuel is both safe and effective. After major accidents the core (or what is left of it) is normally subject to PIE in order to find out what happened. One site where PIE is done is the ITU which is the EU centre for the study of highly radioactive materials.
- Fission gas release
- Cracking of the fuel
- This is due to the fact that as the fuel expands on heating, the core of the pellet expands more than the rim. Because of the thermal stress thus formed the fuel cracks, the cracks tend to go from the centre to the edge in a star shaped pattern.
The temperature of the fuel varies as a function of the distance from the centre to the rim. At distance x from the centre the temperature (Tx) is described by the equation where ρ is the power density (W m-3) and Kf is the thermal conductivity.
- Tx = TRim + ρ (rpellet2 - x2) (4 Kf)-1
To explain this for a series of fuel pellets being used with a rim temperature of 200 oC (typical for a BWR) with different diameters and power densities of 250 Wm-3 have been modeled using the above equation. Note that these fuel pellets are rather large; it is normal to use oxide pellets which are about 10 mm in diameter.
Radioisotope decay fuels
The terms atomic battery, nuclear battery and radioisotope battery are used to describe a device which uses the radioactive decay to generate electricity. Although RTG's could strictly be said to belong to this class, the term generally refers to non-thermal converters whose output power is not a function of a temperature difference. Several designs, exploiting charged alpha and beta particles are available. These include: the direct charging generators; Betavoltaics; the optoelectric nuclear battery and the radioisotope piezoelectric generator.
These systems use radioisotopes that produce low energy beta particles or sometimes alpha particles of varying energies. Low energy beta particles are needed to prevent the production of high energy penetrating Bremsstrahlung radiation that would require heavy shielding. Radioisotopes such as tritium, nickel-63, promethium-147, and technetium-99 have been tested. Plutonium-238, curium-242, curium-244 and strontium-90 have been used.
Radioisotope thermoelectric generators
A radioisotope thermoelectric generator (RTG) is a simple electrical generator which obtains its power from radioactive decay. In such a device, the heat released by the decay of a suitable radioactive material is converted into electricity using an array of thermocouples
238Pu has become the most widely used fuel for RTGs. In the form of plutonium dioxide it has a half-life of 87.7 years, reasonable energy density and exceptionally low gamma and neutron radiation levels. Some Russian terrestrial RTGs have used 90Sr; this isotope has a shorter half-life and a much lower energy density, but is cheaper. Early RTGs, first built in 1958 by the U.S. Atomic Energy Commission, have used 210Po. This fuel provides phenomenally huge energy density, (a single gram of polonium-210 generates 140 watts thermal) but has limited use because of its very short half-life and gamma production and has been phased out of use in this application.
Radioisotope heater units (RHU)
Their function is to provide highly localised heating of sensitive equipment (such as electronics) in deep space. The Cassini-Huygens orbiter to Saturn contains 82 of these units (in addition to its 3 main RTG's for power generation). The Huygens probe to Titan contains 35 devices.
Most fusion fuels fit in here. They include tritium (3H) and deuterium (2H) as well as helium three (3He). Many other elements can be fused together if they can be forced close enough to each other at high enough temperatures. In general, fusion fuels are expected to have at least three generations based on the ease of fusing light atomic nuclei together.
First generation fusion fuel
- 2H + 3H n (14.07 MeV) + 4He (3.52 MeV)
- 2H + 2H n (2.45 MeV) + 3He (0.82 MeV)
- 2H + 2H p (3.02 MeV) + 3H (1.01 MeV)
Second generation fusion fuel
Second generation fuels require either higher confinement temperatures or longer confinement time than those required of first generation fusion fuels. This group consists of deuterium and helium three. The products of these reactants are all charged particles, but there may be non-beneficial side reactions leading to radioactive activation of fusion reactor components.
- 2H + 3He p (14.68 MeV) + 4He (3.67 MeV)
Third generation fusion fuel
There are several potential third generation fusion fuels. Third generation fusion fuels produce only charged particles in the fusion process and there are no side reactions. Therefore, there would be no radioactive activation of the fusion reactor. This is often seen as the end goal of fusion research. 3He is the first 3rd generation fusion fuel that is likely to be used since it has the lowest Maxwellian reactivity in comparison to other 3rd generation fusion fuels.
- 3He + 3He 2p + 4He (12.86 MeV)
Another aneutronic fusion reaction may be the proton-boron reaction:
- p + 11B → 34He
Under reasonable assumptions, side reactions will result in about 0.1% of the fusion power being carried by neutrons. With 123 keV, the optimum temperature for this reaction is nearly ten times higher than that for the pure hydrogen reactions, the energy confinement must be 500 times better than that required for the D-T reaction, and the power density will be 2500 times lower than for D-T. If the percentage of D-T is higher, spontaneous combustion might occur and damage the system.
- INSC Material Properties Database: Thermodynamic Properties, International Nuclear Safety Center
- J.Y. Colle, J.P. Hiernaut, D. Papaioannou, C. Ronchi, A. Sasahara, Journal of Nuclear Materials, 2006, 348, 229
- Radiochemistry and Nuclear Chemistry, G. Choppin, J-O Liljenzin and J. Rydberg, 3rd Ed, 2002, Butterworth-Heinemann, ISBN 0-7506-7463-6