CHENNAI: India today achieved a milestone in nuclear power
generation, touching 32,000 million units, a target that nuclear
operatorNuclear Power Corporation of India Limited(NPCIL)
had set for this fiscal. "We had our annual target of electricity
production, which we met today. We achieved 32,000 million units
So as a company, we have met our annual target," Shiv
Abhilash Bhardwaj, Director (Technical), NPCIL told reporters
here. However, he said that some plants are operating at lower
power because of fuel mismatch. "And very soon, that problem
is going to be over because Tummalapalli mine (Andhra
Pradesh) is right now getting commissioned and in a couple
of months we will start getting Uranium from that," he said.
India, which has a total capacity of 4680 MW from nuclear power,
is looking forward to install a "very large number" nuclear plants,
he said. "We are now talking to France. We are quite in advanced
stage of talking to France on EPR technology with a capacity of
1,650 MW unit in Jaitapur. As for South Korea, India has
signed a Inter-governmental agreement last July. We are in
the status of defining our specifications. South Korea has 1,400
MW reactor and they have to change some designs as per our
requirements," he said, asked about India's ongoing collaborations
with other countries.
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
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, the breeder reactor can employ not only
the recycled plutonium and uranium in spent fuel, but all theactinides,
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
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. 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.
Separated components and disposition
The potentially useful components dealt with in nuclear reprocessing
comprise specific actinides (plutonium, uranium, and someminor
|plutonium, minor actinides, reprocessed uranium||fission in fast, fusion, or subcritical reactor|
|reprocessed uranium, cladding, filters||less stringent storage as intermediate-level waste|
|long-lived fission and activation products||nuclear transmutation or geological repository|
|medium-lived fission products 137Cs and 90Sr||medium-term storage as high-level waste|
|useful radionuclides and noble metals||industrial and medical uses|
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
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,
was developed and tested at the Oak Ridge National Laboratory
(ORNL) in the 1943–1945 period to produce quantities of plutonium
for evaluation and use in 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 latter 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, New York which closed by 1972 because
of its inability to meet new regulatory requirements.
Reprocessing of civilian fuel has long been employed in Europe,
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, fear 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 April 7, 1977,
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 own policy and signed a contract with a consortium
design and operate a Mixed Oxide (MOX) fuel fabrication
facility. Site preparation at the Savannah River Site
(South Carolina) began in October 2005.
Water and organic solvents
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
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-gradeplutonium, which can later be recovered using
PUREX. Because of this, PUREX chemicals are monitored.
Modifications of PUREX
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
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 ofacetohydroxamic acid (AHA) to the
extraction and scrub sections of the process. The addition
of AHA greatly diminishes the extractability of plutonium
and neptunium, providing greater proliferation resistance
than with the plutonium extraction stage of the PUREX process.
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.
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
and oxygen. Such an organic waste can be burned without
the formation of acidic gases which could contribute to
acid rain. The DIAMEX process is being worked on in
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.
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
Other systems such as the dithiophosphinic acids are being worked on
by some other workers.
The UNiversal EXtraction process was developed in Russia and
the Czech Republic; it is designed to completely remove the most
the raffinate remaining after the extraction of uranium and
(known as chlorinated cobalt dicarbollide). The actinides
such as nitrobenzene. Other dilents such as
have been suggested as well.
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
The fuel was decladded by boiling it incaustic 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 bismuthnitrate
and phosphoric acid to form the bismuth phosphate. The plutonium
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
added, and the plutonium re-precipitated again using a bismuth
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 solvationmechanism. This process has the disadvantage
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
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
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 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
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.
- 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
- 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.
- 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.
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.
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
electrode it is possible to use a molten bismuth cathode, or a solid
As an alternative to electrowinning, the wanted metal can be
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
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
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 (for volumetric oxidation) involves heating oxide
fuel with oxygen, sometimes with alternating oxidation and reduction,
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
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 cesium-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.
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 inuranium 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
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,
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
rutheniumhexafluoride is relatively stable and volatile.
Distillation of the residue at higher temperatures can separate lower-
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 returnactinides
to the molten fuel mixture for eventual fission, while removing
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
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,
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).
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
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
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
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
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
List of sites
|Country||Reprocessing site||Fuel type||Procedure||Reprocessing|
or operating period
|Belgium||Mol||LWR, MTR (Material test reactor)||80||1966–1974|
|China||intermediate pilot plant||60–100||1968-early 1970s|
|France||Marcoule, UP 1||Military||1,200||1958-1997|
|France||Marcoule, CEA APM||FBR||PUREX DIAMEX SANEX||6||1988–present|
|France||La Hague, UP 2||LWR||PUREX||900||1967–1974|
|France||La Hague, UP 2–400||LWR||PUREX||400||1976–1990|
|France||La Hague, UP 2–800||LWR||PUREX||800||1990|
|France||La Hague, UP 3||LWR||PUREX||800||1990|
|India||Kalpakkam||PHWR and FBTR||100||1998|
|Pakistan||New Labs, Rawalpindi||Military/Plutonium/Thorium||80||1982–Present|
|Pakistan||Khushab Nuclear Complex,Atomic City of Pakistan||HWR/Military/Tritium||22 kg ||1986–Present|
|Russia||Mayak Plant B||Military||400||1948-196?|
|Russia||Mayak Plant BB, RT-1||LWR||PUREX + Np separation||400||1978|
|Russia||Zheleznogorsk (Krasnoyarsk-26), RT-2||VVER||1,500||under construction|
|USA,WA||Hanford Site||Military||bismuth phospate, REDOX, PUREX||1944–present|
|USA, SC||Savannah River Site||Military/LWR/HWR/Tritium||PUREX, REDOX, THOREX, Np separation||5000||1952–2002|
|USA, NY||West Valley||LWR||PUREX||300||1966–1972|