Nuclear fuel is a substance that is used in nuclear power stations to
produce heat to power turbines.
Heat is created when nuclear fuel
undergoes nuclear fission.
Most nuclear fuels contain heavy fissile elements that are capable of
nuclear fission, such as
Uranium-235 or Plutonium-239. When the
unstable nuclei of these atoms are hit by a slow-moving neutron, they
split, creating two daughter nuclei and two or three more neutrons.
These neutrons then go on to split more nuclei. This creates a
self-sustaining chain reaction that is controlled in a nuclear
reactor, or uncontrolled in a nuclear weapon.
The processes involved in mining, refining, purifying, using, and
disposing of nuclear fuel are collectively known as the nuclear fuel
Not all types of nuclear fuels create power from nuclear fission;
plutonium-238 and some other elements are used to produce small
amounts of nuclear power by radioactive decay in radioisotope
thermoelectric generators and other types of atomic batteries.
Nuclear fuel has the highest energy density of all practical fuel
1 Oxide fuel
2 Metal fuel
2.3 Molten plutonium
4 Liquid fuels
4.1 Molten salts
4.2 Aqueous solutions of uranyl salts
5 Common physical forms of nuclear fuel
5.1 PWR fuel
5.2 BWR fuel
6 Less-common fuel forms
6.2 TRISO fuel
6.3 QUADRISO fuel
6.5 CerMet fuel
6.6 Plate-type fuel
6.7 Sodium-bonded fuel
7 Spent nuclear fuel
7.1 Oxide fuel under accident conditions
8 Fuel behavior and post-irradiation examination
Radioisotope decay fuels
Radioisotope thermoelectric generators
Radioisotope heater units (RHU)
10 Fusion fuels
10.1 First-generation fusion fuel
10.2 Second-generation fusion fuel
10.3 Third-generation fusion fuel
11 See also
13 External links
13.1 PWR fuel
13.2 BWR fuel
13.4 TRISO fuel
13.5 QUADRISO fuel
13.6 CERMET fuel
13.7 Plate type fuel
13.9 Fusion fuel
For fission reactors, the fuel (typically based on uranium) is usually
based on the metal oxide; the oxides are used rather than the metals
themselves because the oxide melting point is much higher than that of
the metal and because it cannot burn, being already in the oxidized
The thermal conductivity of zirconium metal and uranium dioxide as a
function of temperature
Uranium dioxide is a black semiconducting 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 then
be converted by heating in an argon / hydrogen mixture (700 °C)
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.
Corrosion of uranium dioxide in water is controlled by similar
electrochemical processes to the galvanic corrosion of a metal
Main article: MOX fuel
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
MOX fuel is an alternative to low enriched uranium (LEU)
fuel used in the light water reactors which predominate nuclear power
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.
Reprocessing of commercial nuclear fuel to make MOX was done in the
Sellafield MOX Plant
Sellafield MOX Plant (England). As of 2015,
MOX fuel is made in France
(see Marcoule Nuclear Site), and to a lesser extent in Russia (see
Mining and Chemical Combine), India and Japan. China plans to develop
fast breeder reactors (see CEFR) and reprocessing.
The Global Nuclear Energy Partnership, was a U.S. proposal in the
George W. Bush Administration 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.Normally, with the fuel being changed every three years or so,
about half of the Pu-239 is 'burned' in the reactor, providing about
one third of the total energy. It behaves like U-235 and its fission
releases a similar amount of energy. The higher the burn-up, the more
plutonium in the spent fuel, but the lower the fraction of fissile
plutonium. Typically about one percent of the used fuel discharged
from a reactor is plutonium, and some two thirds of this is fissile
(c. 50% Pu-239, 15% Pu-241). Worldwide, some 70 tonnes of plutonium
contained in used fuel is removed when refueling reactors each
Metal fuels have the advantage of a much higher heat conductivity than
oxide fuels but cannot survive equally high temperatures. Metal fuels
have a long history of use, stretching from the Clementine reactor in
1946 to many test and research reactors. Metal fuels have the
potential for the highest fissile atom density. Metal fuels are
normally alloyed, but some metal fuels have been made with pure
Uranium alloys that have been used include uranium
aluminum, uranium zirconium, uranium silicon, uranium molybdenum, and
uranium zirconium hydride. Any of the aforementioned fuels can be made
with plutonium and other actinides as part of a closed nuclear fuel
cycle. Metal fuels have been used in water reactors and liquid metal
fast breeder reactors, such as EBR-II.
TRIGA fuel is used in
TRIGA (Training, Research, Isotopes, General
Atomics) reactors. The
TRIGA reactor uses UZrH fuel, which has a
prompt negative fuel temperature coefficient of reactivity, meaning
that as the temperature of the core increases, the reactivity
decreases—so it is highly unlikely for a meltdown to occur. Most
cores that use this fuel are "high leakage" cores where the excess
leaked neutrons can be utilized for research.
TRIGA fuel was
originally designed to use highly enriched uranium, however in 1978
the U.S. Department of Energy launched its Reduced Enrichment for
Research Test Reactors program, which promoted reactor conversion to
low-enriched uranium fuel. A total of 35
TRIGA reactors have been
installed at locations across the USA. A further 35 reactors have been
installed in other countries.
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 minor
actinides. It can be made inherently safe as thermal expansion of the
metal alloy will increase neutron leakage.
Molten plutonium, alloyed with other metals to lower its melting point
and encapsulated in tantalum, was tested in two experimental
reactors, LAMPRE I and LAMPRE II, at LANL in the 1960s. "LAMPRE
experienced three separate fuel failures during operation."
Ceramic fuels other than oxides have the advantage of high heat
conductivities and melting points, but they are more prone to swelling
than oxide fuels and are not understood as well.
This is often the fuel of choice for reactor designs that NASA
produces, one advantage is that UN has a better thermal conductivity
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 (n,p) 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 pyroprocessing to enable 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
Much of what is known about uranium carbide is in the form of pin-type
fuel elements for liquid metal fast reactors during their intense
study during the 1960s and 1970s. 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 makes uranium
carbide an attractive fuel. In addition, because of the absence of
oxygen in this fuel (during the course of irradiation, excess gas
pressure can build from the formation of O2 or other gases) as well as
the ability to complement 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.
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Liquid fuels are liquids containing dissolved nuclear fuel and have
been shown to offer numerous operational advantages compared to
traditional solid fuel approaches.
Liquid-fuel reactors offer significant safety advantages due to their
inherently stable "self-adjusting" reactor dynamics. This provides two
major benefits: - virtually eliminating the possibility of a run-away
reactor meltdown, - providing an automatic load-following capability
which is well suited to electricity generation and high temperature
industrial heat applications.
Another major advantage of the liquid core is its ability to be
drained rapidly into a passively safe dump-tank. This advantage was
conclusively demonstrated repeatedly as part of a weekly shutdown
procedure during the highly successful 4 year ORNL MSRE program.
Another huge advantage of the liquid core is its ability to release
xenon gas which normally acts as a neutron absorber and causes
structural occlusions in solid fuel elements (leading to early
replacement of solid fuel rods with over 98% of the nuclear fuel
unburned, including many long lived actinides). In contrast Molten
Salt Reactors (MSR) are capable of retaining the fuel mixture for
significantly extended periods, which not only increases fuel
efficiency dramatically, but also incinerates the vast majority of its
own waste as part of the normal operational characteristics.
Molten salt fuels have nuclear fuel dissolved directly in the molten
salt coolant. Molten salt-fueled reactors, such as the liquid fluoride
thorium reactor (LFTR), are different from molten salt-cooled reactors
that do not dissolve nuclear fuel in the coolant.
Molten salt fuels were used in the LFTR known as the Molten Salt
Reactor Experiment, as well as other liquid core reactor experiments.
The liquid fuel for the molten salt reactor was a mixture of lithium,
beryllium, thorium and uranium fluorides: 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 operated at much higher
temperatures, since the boiling point of the molten salt was in excess
of 1400 °C.
Aqueous solutions of uranyl salts
The aqueous homogeneous reactors (AHRs) use a solution of uranyl
sulfate or other uranium salt in water. Historically, AHRs have all
been small research reactors, not large power reactors. An AHR known
as the Medical Isotope Production System is being considered for
production of medical isotopes.
Common physical forms of nuclear fuel
See also: Active fuel length
Uranium dioxide (UO2) powder is compacted to cylindrical pellets and
sintered at high temperatures to produce ceramic nuclear fuel pellets
with a high density and well defined physical properties and chemical
composition. A grinding process is used to achieve a uniform
cylindrical geometry with narrow tolerances. Such fuel pellets are
then stacked and filled into the metallic tubes. 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 which,
in addition to being highly corrosion-resistant, has low neutron
absorption. The tubes containing the fuel pellets are sealed: these
tubes are called fuel rods. The finished fuel rods are grouped into
fuel assemblies that are used to build up the core of a power reactor.
Cladding is the outer layer of the fuel rods, standing between the
coolant and the nuclear fuel. It is made of a corrosion-resistant
material with low absorption cross section for thermal neutrons,
Zircaloy or steel in modern constructions, or magnesium with
small amount of aluminium and other metals for the now-obsolete Magnox
reactors. Cladding prevents radioactive fission fragments from
escaping the fuel into the coolant and contaminating it.
Nuclear Regulatory Commission
Nuclear Regulatory Commission (NRC) photo of unirradiated (fresh) fuel
NRC photo of fresh fuel pellets ready for assembly.
NRC photo of fresh fuel being inspected.
PWR fuel assembly (also known as a fuel bundle) This fuel assembly is
from a pressurized water reactor of the nuclear-powered passenger and
cargo ship NS Savannah. Designed and built by the Babcock &
Pressurized water reactor
Pressurized water reactor (PWR) fuel consists of cylindrical rods put
into bundles. A uranium oxide ceramic is formed into pellets and
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 14×14 to
17×17. PWR fuel bundles are about 4 meters long. 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
Zircaloy tubes are pressurized with helium
to try to minimize pellet-cladding interaction 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 affecting neutronics and thermal hydraulics of
the reactor core. In modern BWR fuel bundles, there are either 91, 92,
or 96 fuel rods per assembly depending on the manufacturer. A range
between 368 assemblies for the smallest and 800 assemblies for the
largest U.S. BWR forms the reactor core. Each BWR fuel rod is
backfilled with helium to a pressure of about three atmospheres (300
CANDU fuel bundles Two
Deuterium Uranium") fuel
bundles, each about 50 cm long, 10 cm in diameter.
CANDU fuel bundles are about a half meter long and 10 cm in
diameter. They consist of sintered (UO2) pellets in zirconium alloy
tubes, welded to zirconium alloy end plates. Each bundle is roughly
20 kg, and a typical core loading is on the order of 4500-6500
bundles, depending on the design. 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 also about 10 cm (4 inches) in
diameter, 0.5 m (20 in) long and weighs about 20 kg
(44 lb) and replaces the 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.
A magnox fuel rod
Magnox reactors are pressurised, carbon dioxide–cooled,
graphite-moderated reactors using natural uranium (i.e. unenriched) as
Magnox alloy as fuel cladding. Working pressure varies from
6.9 to 19.35 bar for the steel pressure vessels, and the two
reinforced concrete designs operated at 24.8 and 27 bar.
consists mainly of magnesium with small amounts of aluminium and other
metals—used in cladding unenriched uranium metal fuel with a
non-oxidising covering to contain fission products.
Magnox is short
Magnesium non-oxidising. This material has the advantage of a low
neutron capture cross-section, but has two major disadvantages:
It limits the maximum temperature, and hence the thermal efficiency,
of the plant.
It reacts with water, preventing long-term storage of spent fuel under
Magnox fuel incorporated cooling fins to provide maximum heat transfer
despite low operating temperatures, making it expensive to produce.
While the use of uranium metal rather than oxide made reprocessing
more straightforward and therefore cheaper, the need to reprocess fuel
a short time after removal from the reactor meant that the fission
product hazard was severe. Expensive remote handling facilities were
required to address this danger.
TRISO fuel particle which has been cracked, showing the multiple
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 not to
crack due to the stresses from processes (such as differential thermal
expansion or fission gas pressure) at temperatures up to and 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 the pebble-bed reactor (PBR), in which thousands of TRISO
fuel particles are dispersed into graphite pebbles, and the
prismatic-block gas-cooled reactor (such as the GT-MHR), in which the
TRISO fuel particles are fabricated into compacts and placed in a
graphite block matrix. Both of these reactor designs are high
temperature gas reactors (HTGRs). These are also the basic reactor
designs of very-high-temperature reactors (VHTRs), one of the six
classes of reactor designs in the Generation IV initiative that is
attempting to reach even higher HTGR outlet temperatures.
TRISO fuel particles were originally developed in the United Kingdom
as part of the
Dragon reactor project. The inclusion of the SiC as
diffusion barrier was first suggested by D. T. Livey. The first
nuclear reactor to use TRISO fuels was the
Dragon reactor 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
High-temperature engineering test reactor in Japan. Spherical fuel
elements utilizing a TRISO particle with a UO2 and UC solid solution
kernel are being used in the Xe-100 in the United States.
In QUADRISO particles a burnable neutron poison (europium oxide or
erbium oxide or carbide) layer surrounds the fuel kernel of ordinary
TRISO particles to better manage the excess of reactivity. If the core
is equipped both with TRISO and QUADRISO fuels, at beginning of life
neutrons do not reach the fuel of the QUADRISO particles because they
are stopped by the burnable poison. During irradiation, the poison
depletes and more neutrons are able to stream into the fuel kernel of
QUADRISO particles inducing fission reactions. This mechanism
compensates for the normal depletion of fissile material during fuel
burnup. In the generalized QUADRISO fuel concept the poison can
eventually be mixed with the fuel kernel or the outer pyrocarbon. The
QUADRISO  concept has been conceived at Argonne National
RBMK reactor fuel rod holder 1 – distancing armature; 2 – fuel
rods shell; 3 – fuel tablets.
RBMK reactor fuel was used in Soviet-designed and built RBMK-type
reactors. This is a low-enriched uranium oxide fuel. The fuel elements
RBMK are 3 m long each, and two of these sit back-to-back on
each fuel channel, pressure tube.
Reprocessed uranium from Russian
VVER reactor spent fuel is used to fabricate
RBMK fuel. Following the
Chernobyl accident, the enrichment of fuel was changed from 2.0% to
2.4%, to compensate for control rod modifications and the introduction
of additional absorbers.
CerMet fuel consists of ceramic fuel particles (usually uranium oxide)
embedded in a metal matrix. It is hypothesized[by whom?] that this
type of fuel is what is used in United States Navy reactors. This fuel
has high heat transport characteristics and can withstand a large
amount of expansion.
ATR Core The
Advanced Test Reactor
Advanced Test Reactor at
Idaho National Laboratory
Idaho National Laboratory uses
plate-type fuel in a clover leaf arrangement. The blue glow around the
core is known as Cherenkov radiation.
Plate-type fuel has fallen out of favor over the years. Plate-type
fuel is commonly composed of enriched uranium sandwiched between metal
cladding. Plate-type fuel is used in several research reactors where a
high neutron flux is desired, for uses such as material irradiation
studies or isotope production, without the high temperatures seen in
ceramic, cylindrical fuel. It is currently used in the Advanced Test
Reactor (ATR) at Idaho National Laboratory, and the nuclear research
reactor at the University of Massachusetts Lowell Radiation
Sodium-bonded fuel consists of fuel that has liquid sodium in the gap
between the fuel slug (or pellet) and the cladding. This fuel type is
often used for sodium-cooled liquid metal fast reactors. It has been
used in EBR-I, EBR-II, and the FFTF. The fuel slug may be metallic or
ceramic. The sodium bonding is used to reduce the temperature of the
Spent nuclear fuel
Main article: 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 be heterogeneous; often the fuel will contain nanoparticles of
platinum group metals such as palladium. Also the fuel may well have
cracked, swollen, and been heated 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
Main article: Behavior of nuclear fuel during a reactor accident
Two main modes of release exist, the fission products can be vaporised
or small particles of the fuel can be dispersed.
Fuel behavior and post-irradiation examination
Main article: Post Irradiation Examination
Post-Irradiation Examination (PIE) is the study of used nuclear
materials such as nuclear fuel. It has several purposes. It is known
that by examination of used fuel that the failure modes which occur
during normal use (and the manner in which the fuel will behave during
an accident) can be studied. In addition information is gained which
enables the users of fuel to assure themselves of its quality and it
also assists in the development of new fuels. After major accidents
the core (or what is left of it) is normally subject to PIE 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.
Materials in a high-radiation environment (such as a reactor) can
undergo unique behaviors such as swelling  and non-thermal creep.
If there are nuclear reactions within the material (such as what
happens in the fuel), the stoichiometry will also change slowly over
time. These behaviors can lead to new material properties, cracking,
and fission gas release.
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 center 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.
According to the
International Nuclear Safety Center
International Nuclear Safety Center  the thermal
conductivity of uranium dioxide can be predicted under different
conditions by a series of equations.
The bulk density of the fuel can be related to the thermal
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
Kf = Ko(1 − p/1 + (s − 1)p)
Rather than measuring the thermal conductivity using the traditional
methods such as Lees' disk, the Forbes' method, or Searle's bar, it is
common to use
Laser Flash Analysis
Laser Flash Analysis 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 calculate
and determine the thermal conductivity.
λ = ρCpα
λ thermal conductivity
Cp heat capacity
α thermal diffusivity
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 
Radioisotope decay fuels
Main article: atomic battery
The terms atomic battery, nuclear battery and radioisotope battery are
used interchangeably to describe a device which uses the radioactive
decay to generate electricity. 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 plutonium-238,
curium-242, curium-244 and strontium-90 have been used. Tritium,
nickel-63, promethium-147, and technetium-99 have been tested.
There are two main categories of atomic batteries: thermal and
non-thermal. The non-thermal atomic batteries, which have many
different designs, exploit charged alpha and beta particles. These
designs include the direct charging generators, betavoltaics, the
optoelectric nuclear battery, and the radioisotope piezoelectric
generator. The thermal atomic batteries on the other hand, convert the
heat from the radioactive decay to electricity. These designs include
thermionic converter, thermophotovoltaic cells, alkali-metal thermal
to electric converter, and the most common design, the radioisotope
Radioisotope thermoelectric generators
Inspection of Cassini spacecraft RTGs before launch
A radioisotope thermoelectric generator (RTG) is a simple electrical
generator which converts heat into electricity from a radioisotope
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 for this application.
Photo of a disassembled RHU
Radioisotope heater units (RHU)
Radioisotope heater units normally provide about 1 watt of heat each,
derived from the decay of a few grams of plutonium-238. This heat is
given off continuously for several decades.
Their function is to provide highly localised heating of sensitive
equipment (such as electronics in outer space). The Cassini–Huygens
Saturn contains 82 of these units (in addition to its 3
main RTG's for power generation). The Huygens probe to Titan contains
Fusion power § Fuels
Fusion fuels include deuterium (2H) and tritium (3H) as well as
helium-3 (3He). Many other elements can be fused together, but the
larger electrical charge of their nuclei means that much higher
temperatures are required. Only the fusion of the lightest elements is
seriously considered as a future energy source. Fusion of the lightest
atom, 1H hydrogen, as is done in the Sun and stars, has also not been
considered practical on Earth. Although the energy density of fusion
fuel is even higher than fission fuel, and fusion reactions sustained
for a few minutes have been achieved, utilizing fusion fuel as a net
energy source remains only a theoretical possibility.
First-generation fusion fuel
Deuterium and tritium are both considered first-generation fusion
fuels; they are the easiest to fuse, because the electrical charge on
their nuclei is the lowest of all elements. The three most commonly
cited nuclear reactions that could be used to generate energy are:
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, but generate fewer neutrons. Neutrons are an unwanted
byproduct of fusion reactions in an energy generation context, because
they are absorbed by the walls of a fusion chamber, making them
radioactive. They cannot be confined by magnetic fields, because they
are not electrically charged. This group consists of deuterium and
helium-3. The products are all charged particles, but there may be
significant side reactions leading to the production of neutrons.
2H + 3He → p (14.68 MeV) + 4He (3.67 MeV)
Third-generation fusion fuel
Main article: Aneutronic fusion
Third-generation fusion fuels produce only charged particles in the
primary reactions, and side reactions are relatively unimportant.
Since a very small amount of neutrons is produced, there would be
little induced radioactivity in the walls of the fusion chamber. This
is often seen as the end goal of fusion research. 3He has the highest
Maxwellian reactivity of any 3rd generation fusion fuel. However,
there are no significant natural sources of this substance on Earth.
3He + 3He → 2p + 4He (12.86 MeV)
Another potential aneutronic fusion reaction is the proton-boron
p + 11B → 34He (8.7 MeV)
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.
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^ "Archived copy" (PDF). Archived (PDF) from the original on
2016-10-21. Retrieved 2016-06-04.
^ "Archived copy" (PDF). Archived (PDF) from the original on
2016-04-15. Retrieved 2013-11-11.
^ "B&W Medical Isotope Production System". The Babcock &
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^ "Nuclear Fusion Power". World Nuclear Association. September 2009.
NEI fuel schematic
Picture of a PWR fuel assembly
Picture showing handling of a PWR bundle
Mitsubishi nuclear fuel Co.
Picture of a "canned" BWR assembly
Physical description of LWR fuel
Links to BWR photos from the nuclear tourist webpage
CANDU Fuel pictures and FAQ
The Evolution of
CANDU Fuel Cycles and their Potential Contribution to
CANDU Fuel-Management Course
CANDU Fuel and Reactor Specifics (Nuclear Tourist)
Candu Fuel Rods and Bundles
TRISO fuel descripción
Non-Destructive Examination of SiC Nuclear Fuel Shell using X-Ray
Fluorescence Microtomography Technique
GT-MHR fuel compact process
Description of TRISO fuel for "pebbles"
LANL webpage showing various stages of TRISO fuel production
Method to calculate the temperature profile in TRISO fuel
Conceptual Design of QUADRISO Fuel
A Review of Fifty Years of Space Nuclear Fuel Development Programs
Thoria-based Cermet Nuclear Fuel: Sintered Microsphere Fabrication by
The Use of Molybdenum-Based Ceramic-Metal (CerMet) Fuel for the
Actinide Management in LWRs
Plate type fuel
List of reactors at INL and picture of ATR core
ATR plate fuel
TRIGA fuel website
Advanced fusion fuels presentation
Radioisotope Thermoelectric Generator
Accidents and incidents
Radioisotope thermoelectric (RTG)
Safety and security
Single-photon emission (SPECT)
Positron-emission tomography (PET)
Neutron capture therapy of cancer
Estimated death tolls from attacks
States with nuclear weapons
Blue Ribbon Commission on America's Nuclear Future
Nuclear power phase-out
Uranium Naturel Graphite Gaz (UNGG)
Advanced gas-cooled (AGR)
Ultra-high-temperature experiment (UHTREX)
Pebble-bed (PBMR) (HTR-PM)
Gas-turbine modular-helium (GTMHR)
Liquid-fluoride thorium reactor (LFTR)
Molten-Salt Reactor Experiment
Molten-Salt Reactor Experiment (MSRE)
Integral Molten Salt Reactor (IMSR)
Small sealed transportable autonomous (SSTAR)
Energy Multiplier Module (EM2)
Fast Breeder Test Reactor (FBTR)
Dual fluid reactor (DFR)
Helium gas (GFR)
Organically moderated and cooled reactor
Aircraft Reactor Experiment
Reversed field pinch
Dense plasma focus