Nuclear reprocessing technology was developed to
chemically separate and recover fissionable plutonium from irradiated nuclear
fuel. Reprocessing serves multiple purposes, whose relative importance has
changed over time. Originally reprocessing was used solely to extract plutonium
for producing nuclear weapons. With the commercialization of nuclear
power, the reprocessed plutonium was recycled back into MOX
nuclear fuel for thermal reactors. The reprocessed uranium, which constitutes the bulk
of the spent fuel material, can in principle also be re-used as fuel, but that
is only economic when uranium prices are high. Finally, a breeder
reactor is not restricted to using recycled plutonium and uranium. It can
employ all the actinides,
closing the nuclear fuel cycle and potentially multiplying
the energy extracted from natural uranium by about 60 times.
Nuclear reprocessing reduces the volume of high-level waste,
but by itself does not reduce radioactivity or heat generation and therefore
does not eliminate the need for a geological waste repository. Reprocessing has
been politically controversial because of the potential to contribute to nuclear proliferation, the potential
vulnerability to nuclear terrorism, the political challenges of
repository siting (a problem that applies equally to direct disposal of spent
fuel), and because of its high cost compared to the once-through fuel cycle. In
the United States, the Obama administration stepped back from President Bush's
plans for commercial-scale reprocessing and reverted to a program focused on
reprocessing-related scientific research. Nuclear fuel reprocessing is
performed routinely in Europe, Russia and Japan.
The potentially useful components dealt with in nuclear
reprocessing comprise specific actinides (plutonium, uranium, and some minor
actinides). The lighter elements
components include fission products, activation products, and cladding.
History
The first large-scale nuclear reactors were built during World War
II. These reactors were designed for the production of plutonium for use in
nuclear
weapons. The only reprocessing required, therefore, was the extraction of
the plutonium
(free of fission-product contamination) from the spent natural
uranium fuel. In 1943, several methods were proposed for separating the
relatively small quantity of plutonium from the uranium and fission products.
The first method selected, a precipitation process called the bismuth phosphate process, was developed
and tested at the Oak Ridge National Laboratory (ORNL)
between 1943 and 1945 to produce quantities of plutonium for evaluation and use
in the US weapons programs. ORNL produced the
first macroscopic quantities (grams) of separated plutonium with these
processes.
The bismuth phosphate process was first operated on a large
scale at the Hanford Site, in the later part of 1944. It was
successful for plutonium separation in the emergency situation existing then,
but it had a significant weakness: the inability to recover uranium.
The first successful solvent extraction process for the
recovery of pure uranium and plutonium was developed at ORNL in 1949. The PUREX process is the
current method of extraction. Separation plants were also constructed at Savannah River Site and a smaller plant at West Valley Reprocessing Plant which
closed by 1972 because of its inability to meet new regulatory requirements.
Reprocessing of civilian fuel has long been employed at the COGEMA La Hague site in France, the Sellafield
site in the United Kingdom, the Mayak Chemical Combine in Russia, and at sites such as the
Tokai plant in Japan, the Tarapur plant in India, and briefly at the West Valley Reprocessing Plant in
the United States.
In October 1976, concern of nuclear weapons proliferation
(especially after India
demonstrated nuclear weapons capabilities using reprocessing technology) led
President Gerald
Ford to issue a Presidential directive to indefinitely
suspend the commercial reprocessing and recycling of plutonium in the U.S. On 7
April 1977, President Jimmy Carter banned the reprocessing of commercial
reactor spent nuclear fuel. The key issue driving this
policy was the serious threat of nuclear weapons proliferation by diversion of
plutonium from the civilian fuel cycle, and to encourage other nations to
follow the USA lead. After that, only countries that already had large
investments in reprocessing infrastructure continued to reprocess spent nuclear
fuel. President Reagan lifted the ban in 1981, but did not provide the
substantial subsidy that would have been necessary to start up commercial
reprocessing.
In March 1999, the U.S. Department of Energy (DOE) reversed
its policy and signed a contract with a consortium
of Duke
Energy, COGEMA,
and Stone & Webster (DCS) to design and operate
a mixed oxide
(MOX) fuel fabrication facility. Site preparation at the Savannah River
Site (South Carolina) began in October 2005. In 2011 the New York Times
reported "...11 years after the government awarded a construction
contract, the cost of the project has soared to nearly $5 billion. The vast
concrete and steel structure is a half-finished hulk, and the government has
yet to find a single customer, despite offers of lucrative subsidies." TVA
(currently the most likely customer) said in April 2011 that it would delay a
decision until it could see how MOX fuel performed in the nuclear accident at Fukushima
Daiichi.
Separation technologies
Water and organic solvents
PUREX
PUREX, the current standard method, is an acronym
standing for Plutonium and Uranium Recovery by EXtraction.
The PUREX process is a liquid-liquid extraction method used to
reprocess spent nuclear fuel, in order to extract uranium and plutonium,
independent of each other, from the fission
products. This is the most developed and widely used process in the industry at
present. When used on fuel from commercial power reactors the plutonium
extracted typically contains too much Pu-240 to be useful in a nuclear weapon.
However, reactors that are capable of refuelling frequently can be used to
produce weapon-grade plutonium, which can later be recovered
using PUREX. Because of this, PUREX chemicals are monitored.
Plutonium Processing
Modifications of PUREX
UREX
The PUREX process can be modified to make a UREX (URanium
EXtraction) process which could be used to save space inside high level nuclear
waste disposal sites, such as the Yucca Mountain nuclear waste
repository, by removing the uranium which makes up the vast majority of the
mass and volume of used fuel and recycling it as reprocessed uranium.
The UREX process is a PUREX process which has been modified
to prevent the plutonium from being extracted. This can be done by adding a
plutonium reductant
before the first metal extraction step. In the UREX process, ~99.9% of the
uranium and >95% of technetium are separated from each other and the other
fission products and actinides. The key is the addition of acetohydroxamic acid (AHA) to the extraction
and scrub sections of the process. The addition of AHA greatly diminishes the
extractability of plutonium and neptunium, providing somewhat greater proliferation
resistance than with the plutonium extraction stage of the PUREX process.
TRUEX
Adding a second extraction agent, octyl(phenyl)-N, N-dibutyl
carbamoylmethyl phosphine oxide(CMPO) in combination with tributylphosphate,
(TBP), the PUREX process can be turned into the TRUEX (TRansUranic
EXtraction) process. TRUEX was invented in the USA by Argonne National
Laboratory and is designed to remove the transuranic metals (Am/Cm) from waste.
The idea is that by lowering the alpha
activity of the waste, the majority of the waste can then be disposed of
with greater ease. In common with PUREX this process operates by a solvation
mechanism.
DIAMEX
As an alternative to TRUEX, an extraction process using a
malondiamide has been devised. The DIAMEX (DIAMideEXtraction)
process has the advantage of avoiding the formation of organic waste which
contains elements other than carbon, hydrogen, nitrogen, and oxygen. Such an organic waste can be burned without the
formation of acidic gases which could contribute to acid rain
(although the acidic gases could be recovered by a scrubber). The DIAMEX
process is being worked on in Europe by the French CEA. The process is sufficiently
mature that an industrial plant could be constructed with the existing
knowledge of the process. In common with PUREX this process operates by a
solvation mechanism.
SANEX
Selective ActiNide EXtraction.
As part of the management of minor actinides it has been proposed that the lanthanides
and trivalent minor actinides should be removed from the PUREX raffinate by
a process such as DIAMEX or TRUEX. In order to allow the actinides such as
americium to be either reused in industrial sources or used as fuel, the lanthanides
must be removed. The lanthanides have large neutron cross sections and hence
they would poison a neutron driven nuclear reaction. To date the extraction
system for the SANEX process has not been defined, but currently several
different research groups are working towards a process. For instance the
French CEA is working on a bis-triazinyl pyridine (BTP) based process.
Other systems such as the dithiophosphinic acids are being worked on by some
other workers.
UNEX
The UNiversal EXtraction process
was developed in Russia
and the Czech Republic; it is designed to completely remove
the most troublesome radioisotopes (Sr, Cs and minor
actinides) from the raffinate remaining after the extraction of uranium and
plutonium from used nuclear fuel. The chemistry is based upon the
interaction of caesium
and strontium
with polyethylene glycol) and a cobalt carborane anion (known as
chlorinated cobalt dicarbollide). The actinides are extracted by CMPO, and the diluent is a
polar aromatic
such as nitrobenzene.
Other dilents such as meta-nitrobenzotrifluoride and
phenyl trifluoromethyl sulfone have been suggested as well.
Electrochemical methods
Obsolete methods
Bismuth phosphate
The bismuth phosphate process is an obsolete
process that adds significant unnecessary material to the final radioactive
waste. The bismuth phosphate process has been replaced by solvent extraction
processes. The bismuth phosphate process was designed to extract plutonium
from aluminium-clad nuclear fuel rods, containing uranium. The fuel
was decladded by boiling it in caustic
soda. After decladding, the uranium metal was dissolved in nitric acid.
The plutonium at this point is in the +4 oxidation state. It
was then precipitated out of the solution by the addition of bismuth nitrate
and phosphoric acid to form the bismuth phosphate. The
plutonium was coprecipitated with this. The supernatant
liquid (containing many of the fission
products) was separated from the solid. The precipitate was then dissolved
in nitric acid before the addition of an oxidant such as potassium permanganate which converted the
plutonium to PuO22+ (Pu VI), then a dichromate
salt was added to maintain the plutonium in the +6 oxidation state.
The bismuth phosphate was next re-precipitated leaving the
plutonium in solution. Then an iron (II) salt such as ferrous
sulfate was added, and the plutonium re-precipitated again using a bismuth
phosphate carrier precipitate. Then lanthanum
salts and fluoride
were added to create solid lanthanum fluoride which acted as a carrier for the
plutonium. This was converted to the oxide by the action of an alkali. The
lanthanum plutonium oxide was next collected and extracted with nitric acid to
form plutonium nitrate.
Hexone or redox
This is a liquid-liquid extraction process which uses methyl isobutyl ketone as the extractant.
The extraction is by a solvation mechanism. This process has the
disadvantage of requiring the use of a salting-out reagent (aluminium nitrate) to
increase the nitrate concentration in the aqueous phase to obtain a reasonable
distribution ratio (D value). Also, hexone is degraded by concentrated nitric
acid. This process has been replaced by the PUREX process.
Pu4+ + 4 NO3− + 2S → [Pu(NO3)4S2]
Butex, β,β'-dibutyoxydiethyl ether
A process based on a solvation extraction process using the
triether extractant named above. This process has the disadvantage of requiring
the use of a salting-out reagent (aluminium nitrate) to
increase the nitrate concentration in the aqueous phase to obtain a reasonable
distribution ratio. This process was used at Windscale
many years ago. This process has been replaced by PUREX.
Pyroprocessing
Pyroprocessing is a generic term for
high-temperature methods. Solvents are molten
salts (e.g. LiCl+KCl or LiF+CaF2) and molten metals (e.g. cadmium, bismuth,
magnesium) rather than water and organic compounds. Electrorefining,
distillation,
and solvent-solvent extraction are common steps.
These processes are not currently in significant use
worldwide, but they have been researched and developed at Argonne National Laboratory and
elsewhere.
Advantages
- The principles behind them are well understood, and no significant technical barriers exist to their adoption.
- Readily applied to high-burnup spent fuel and requires little cooling time, since the operating temperatures are high already.
- Does not use solvents containing hydrogen and carbon, which are neutron moderators creating risk of criticality accidents and can absorb the fission product tritium and the activation product carbon-14 in dilute solutions that cannot be separated later.
- Alternatively, voloxidation can remove 99% of the tritium from used fuel and recover it in the form of a strong solution suitable for use as a supply of tritium.
- More compact than aqueous methods, allowing on-site reprocessing at the reactor site, which avoids transportation of spent fuel and its security issues, instead storing a much smaller volume of fission products on site as high-level waste until decommissioning. For example, the Integral Fast Reactor and Molten Salt Reactor fuel cycles are based on on-site pyroprocessing.
- It can separate many or even all actinides at once and produce highly radioactive fuel which is harder to manipulate for theft or making nuclear weapons. (However, the difficulty has been questioned.) In contrast the PUREX process was designed to separate plutonium only for weapons, and it also leaves the minor actinides (americium and curium) behind, producing waste with more long-lived radioactivity.
- Most of the radioactivity in roughly 102 to 105 years after the use of the nuclear fuel is produced by the actinides, since there are no fission products with half-lives in this range. These actinides can fuel fast reactors, so extracting and reusing (fissioning) them reduces the long-term radioactivity of the wastes.
Disadvantages
- Reprocessing as a whole is not currently (2005) in favor, and places that do reprocess already have PUREX plants constructed. Consequently, there is little demand for new pyrometalurgical systems, although there could be if the Generation IV reactor programs become reality.
- The used salt from pyroprocessing is less suitable for conversion into glass than the waste materials produced by the PUREX process.
- If the goal is to reduce the longevity of spent nuclear fuel in burner reactors, then better recovery rates of the minor actinides need to be achieved.
Electrolysis
PYRO-A and -B for IFR
These processes were developed by Argonne National Laboratory and used in
the Integral Fast Reactor project.
PYRO-A is a means of separating actinides (elements
within the actinide
family, generally heavier than U-235) from non-actinides. The spent fuel is
placed in an anode
basket which is
immersed in a molten salt electrolyte. An electrical current is applied,
causing the uranium metal (or sometimes oxide, depending on the spent fuel) to
plate out on a solid metal cathode while the other actinides (and the rare earths)
can be absorbed into a liquid cadmium cathode. Many of the fission products (such as caesium, zirconium and
strontium)
remain in the salt. As alternatives to the molten cadmium electrode it is
possible to use a molten bismuth cathode, or a solid aluminium cathode.
As an alternative to electrowinning, the wanted metal can be
isolated by using a molten
alloy of an electropositive
metal and a less reactive metal.
Since the majority of the long term radioactivity,
and volume, of spent fuel comes from actinides, removing the actinides produces
waste that is more compact, and not nearly as dangerous over the long term. The
radioactivity of this waste will then drop to the level of various naturally
occurring minerals and ores within a few hundred, rather than thousands of,
years.
The mixed actinides produced by pyrometallic processing can
be used again as nuclear fuel, as they are virtually all either fissile, or fertile,
though many of these materials would require a fast breeder reactor in order to be burned
efficiently. In a thermal neutron spectrum, the concentrations of
several heavy actinides (curium-242 and plutonium-240)
can become quite high, creating fuel that is substantially different from the
usual uranium or mixed uranium-plutonium oxides (MOX) that most current
reactors were designed to use.
Another pyrochemical process, the PYRO-B process, has
been developed for the processing and recycling of fuel from a transmuter reactor ( a fast breeder reactor designed to convert
transuranic nuclear waste into fission products ). A typical transmuter fuel is
free from uranium and contains recovered transuranics
in an inert matrix such as metallic zirconium. In
the PYRO-B processing of such fuel, an electrorefining
step is used to separate the residual transuranic elements from the fission
products and recycle the transuranics to the reactor for fissioning. Newly
generated technetium and iodine are extracted for incorporation into
transmutation targets, and the other fission products are sent to waste.
Voloxidation
Voloxidation (for volumetric oxidation) involves
heating oxide fuel with oxygen, sometimes with alternating oxidation and
reduction, or alternating oxidation by ozone to uranium
trioxide with decomposition by heating back to triuranium octoxide. A major purpose is to
capture tritium
as tritiated water vapor before further processing where it would be difficult
to retain the tritium. Other volatile elements leave the fuel and must be
recovered, especially iodine, technetium, and carbon-14.
Voloxidation also breaks up the fuel or increases its surface area to enhance
penetration of reagents in following reprocessing steps.
Volatilization in isolation
Simply heating spent oxide fuel in an inert atmosphere or
vacuum at a temperature between 700 °C and 1000 °C as a first
reprocessing step can remove several volatile elements, including caesium whose
isotope caesium-137
emits about half of the heat produced by the spent fuel over the following 100
years of cooling (however, most of the other half is from strontium-90
which remains). The estimated overall mass balance for 20,000 grams of
processed fuel with 2,000 grams of cladding is:
Tritium is not mentioned in this paper.
Fluoride volatility
Blue elements have volatile fluorides or are already
volatile; green elements do not but have volatile chlorides; red elements have
neither, but the elements themselves or their oxides are volatile at very high
temperatures. Yields at 100,1,2,3 years after fission,
not considering later neutron capture, fraction of 100% not 200%. Beta decay
Kr-85→Rb, Sr-90→Zr, Ru-106→Pd, Sb-125→Te, Cs-137→Ba, Ce-144→Nd, Sm-151→Eu, Eu-155→Gd visible.
In the fluoride volatility process, fluorine is
reacted with the fuel. Fluorine is so much more reactive than even oxygen that small
particles of ground oxide fuel will burst into flame when dropped into a
chamber full of fluorine. This is known as flame fluorination; the heat
produced helps the reaction proceed. Most of the uranium, which
makes up the bulk of the fuel, is converted to uranium hexafluoride, the form of uranium used
in uranium enrichment, which has a very low boiling
point. Technetium,
the main long-lived fission product, is also
efficiently converted to its volatile hexafluoride. A few other elements also
form similarly volatile hexafluorides, pentafluorides, or heptafluorides. The
volatile fluorides can be separated from excess fluorine by condensation, then
separated from each other by fractional distillation or selective reduction. Uranium hexafluoride and technetium hexafluoride have very similar
boiling points and vapor pressures, which makes complete separation more
difficult.
Many of the fission
products volatilized are the same ones volatilized in non-fluorinated,
higher-temperature volatilization, such as iodine, tellurium and
molybdenum;
notable differences are that technetium is volatilized, but caesium is not.
Some transuranium elements such as plutonium, neptunium and
americium
can form volatile fluorides, but these compounds are not stable when the
fluorine partial pressure is decreased. Most of the plutonium and some of the
uranium will initially remain in ash which drops to the bottom of the flame
fluorinator. The plutonium-uranium ratio in the ash may even approximate the
composition needed for fast neutron reactor fuel. Further
fluorination of the ash can remove all the uranium, neptunium,
and plutonium as volatile fluorides; however, some other minor
actinides may not form volatile fluorides and instead remain with the
alkaline fission products. Some noble
metals may not form fluorides at all, but remain in metallic form; however ruthenium
hexafluoride is relatively stable and volatile.
Distillation of the residue at higher temperatures can
separate lower-boiling transition metal fluorides and alkali
metal (Cs, Rb) fluorides from higher-boiling lanthanide
and alkaline earth metal (Sr, Ba) and yttrium
fluorides. The temperatures involved are much higher, but can be lowered
somewhat by distilling in a vacuum. If a carrier salt like lithium
fluoride or sodium fluoride is being used as a solvent,
high-temperature distillation is a way to separate the carrier salt for reuse.
Molten salt reactor designs carry out fluoride
volatility reprocessing continuously or at frequent intervals. The goal is to
return actinides
to the molten fuel mixture for eventual fission, while removing fission
products that are neutron poisons, or that can be more securely stored
outside the reactor core while awaiting eventual transfer to permanent storage.
Chloride volatility and solubility
Many of the elements that form volatile high-valence fluorides will also form volatile
high-valence chlorides. Chlorination and distillation is another possible
method for separation. The sequence of separation may differ usefully from the
sequence for fluorides; for example, zirconium tetrachloride and tin
tetrachloride have relatively low boiling points of 331 °C and
114.1 °C. Chlorination has even been proposed as a method for removing
zirconium fuel cladding, instead of mechanical decladding.
Chlorides are likely to be easier than fluorides to later
convert back to other compounds, such as oxides.
Chlorides remaining after volatilization may also be
separated by solubility in water. Chlorides of alkaline elements like americium, curium, lanthanides,
strontium,
caesium are
more soluble than those of uranium, neptunium, plutonium, and zirconium.
Radioanalytical separations
In order to determine the distribution of radioactive metals
for analytical purposes, Solvent Impregnated Resins (SIRs)
can be used. SIRs are porous particles, which contain an extractant inside
their pores. This approach avoids the liquid-liquid separation step required in
conventional liquid-liquid extraction. For the
preparation of SIRs for radioanalytical separations, organic Amberlite XAD-4 or
XAD-7 can be used. Possible extractants are e.g. trihexyltetradecylphosphonium
chloride(CYPHOS IL-101) or
N,N0-dialkyl-N,N0-diphenylpyridine-2,6-dicarboxyamides (R-PDA; R = butyl, octy
I, decyl, dodecyl).
Economics
The relative economics of reprocessing-waste disposal and
interim storage-direct disposal has been the focus of much debate over the past
ten years. Studies have
modeled the total fuel cycle costs of a reprocessing-recycling system based on
one-time recycling of plutonium in existing thermal
reactors (as opposed to the proposed breeder
reactor cycle) and compare this to the total costs of an open fuel cycle
with direct disposal. The range of results produced by these studies is very
wide, but all are agreed that under current (2005) economic conditions the
reprocessing-recycle option is the more costly.
If reprocessing is undertaken only to reduce the radioactivity
level of spent fuel it should be taken into account that spent nuclear fuel
becomes less radioactive over time. After 40 years its radioactivity drops
by 99.9%, though it still takes over a thousand years for the level of
radioactivity to approach that of natural uranium.However the level of transuranic elements, including plutonium-239,
remains high for over 100,000 years, so if not reused as nuclear fuel, then
those elements need secure disposal because of nuclear proliferation reasons as well as
radiation hazard.
On 25 October 2011 a commission of the Japanese Atomic
Energy Commission revealed during a meeting calculations about the costs of
recycling nuclear fuel for power generation. These costs could be twice the
costs of direct geological disposal of spent fuel: the cost of extracting
plutonium and handling spent fuel was estimated at 1.98 to 2.14 yen per
kilowatt-hour of electricity generated. Discarding the spent fuel as waste
would cost only 1 to 1.35 yen per kilowatt-hour.
In July 2004 Japanese newspapers reported that the Japanese
Government had estimated the costs of disposing radioactive waste,
contradicting claims four months earlier that no such estimates had been made.
The cost of non-reprocessing options was estimated to be between a quarter and
a third ($5.5–7.9 billion) of the cost of reprocessing ($24.7 billion). At the
end of the year 2011 it became clear that Masaya Yasui, who had been director
of the Nuclear Power Policy Planning Division in 2004, had instructed his
subordinate in April 2004 to conceal the data. The fact that the data were
deliberately concealed obliged the ministry to re-investigate the case and to
reconsider whether to punish the officials involved.
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