Rabu, 28 Maret 2012

... INDIA and NUCLEAR POWER... SO INCREDIBLE..???!! >> ...India's nuclear plants produce 32,000 million units of power..>>

CHENNAI: India today achieved a milestone in nuclear power 
generation, touching 32,000 million units, a target that nuclear 
had set for this fiscal. "We had our annual target of electricity 
production, which we met today. We achieved 32,000 million units
 this morning. 

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

Nuclear reprocessing technology was developed to chemically separate 
and recover fissionable plutonium from irradiated nuclear fuel.[1] 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.[2] 
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
closing the nuclear fuel cycle and potentially multiplying the energy extracted
 from natural uranium by about 60 times.[3][4]
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.[5] 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.[6]

Separated components and disposition

The potentially useful components dealt with in nuclear reprocessing 
comprise specific actinides (plutonium, uranium, and someminor 
actinides). The lighter elements components include fission products
plutonium, minor actinidesreprocessed uraniumfission in fastfusion, or subcritical reactor
reprocessed uranium, cladding, filtersless stringent storage as intermediate-level waste
long-lived fission and activation productsnuclear transmutation or geological repository
medium-lived fission products 137Cs and 90Srmedium-term storage as high-level waste
useful radionuclides and noble metalsindustrial 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 
extraction of the plutonium (free of fission-productcontamination) 
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 BismuthPhosphate process, 
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.[7]
Reprocessing of civilian fuel has long been employed in Europe, 
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, 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, 
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.[8] 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.[9]
In March 1999, the U.S. Department of Energy (DOE) 
reversed its own policy and signed a contract with a consortium
 of Duke EnergyCOGEMA, 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.[10]

Separation technologies

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 
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-gradeplutonium, which can later be recovered using 
PUREX. Because of this, PUREX chemicals are monitored.[citation needed]

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,
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
 contains elements other than carbonhydrogen,nitrogen
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 
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.
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.[11][12][13] 
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 
troublesome radioisotopes (Sr, Cs and minor actinides) from 
the raffinate remaining after the extraction of uranium and 
plutonium from used nuclear fuel.[14][15] The chemistry is 
based upon the interaction of caesium and strontium with 
(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 trifluoromethylsulfone[18] 
have been suggested as well.

Electrochemical methods

An exotic method using electrochemistry and ion exchange in 
ammonium carbonate has been reported.[19]

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 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 
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 sulfatewas 
added, and the plutonium re-precipitated again using a bismuth 
phosphate carrier precipitate. Then lanthanum salts andfluoride 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.[20]
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 
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.[21][22]
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 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. Electrorefiningdistillation
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.[23]
  • 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[24] 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.[25]) 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. 
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 ascaesiumzirconium and strontium
remain in the salt.[26][27][28] As alternatives to the molten cadmium 
electrode it is possible to use a molten bismuth cathode, or a solid
aluminium cathode.[29]
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.[30]
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.[31]
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 
atransmuter 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 (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.[24] 
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 iodinetechnetium, 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 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:[32]


Tritium is not mentioned in this paper.

[edit]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-85Rb, 
Sr-90Zr, Ru-106Pd, Sb-125Te, Cs-137Ba, Ce-144Nd, 
Sm-151Eu, Eu-155Gdvisible.
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
 from each other by fractional distillation or selectivereductionUranium 
hexafluoride andtechnetium 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 asiodinetellurium and molybdenum; notable differences are 
that technetium is volatilized, but caesium is not.
Some transuranium elements such as plutoniumneptunium and 
americium can form volatile fluorides, but these compounds are not 
stable when the fluorine partial pressure is decreased.[33] 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-
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 returnactinides 
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,[24] 
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
curiumlanthanidesstrontiumcaesium are more soluble than 

[edit]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)[34].


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[35] 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%,[36] 
though it still takes over a thousand years for the level of radioactivity 
to approach that of natural uranium.[37] However the level 
of transuranic elements, includingplutonium-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.[38][39]
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.[40][41]

List of sites

CountryReprocessing siteFuel typeProcedureReprocessing
capacity tU/yr
or operating period
 BelgiumMolLWR, MTR (Material test reactor)80[42]1966–1974[42]
 Chinaintermediate pilot plant[43]60–1001968-early 1970s
 ChinaPlant 404[44]502004
 GermanyKarlsruhe, WAKLWR[45]35[42]1971–1990[42]
 FranceMarcoule, UP 1Military1,200[42]1958[42]-1997[46]
 FranceMarcoule, CEA APMFBRPUREX DIAMEX SANEX[47]6[45]1988–present[45]
 FranceLa Hague, UP 2LWR[45]PUREX[48]900[42]1967–1974[42]
 FranceLa Hague, UP 2–400LWR[45]PUREX[48]400[42]1976–1990[42]
 FranceLa Hague, UP 2–800LWRPUREX[48]800[42]1990[42]
 FranceLa Hague, UP 3LWRPUREX[48]800[42]1990[42]
 IndiaKalpakkamPHWR and FBTR100[50]1998[50]
 PakistanNew LabsRawalpindiMilitary/Plutonium/Thorium80[54]1982–Present
 PakistanKhushab Nuclear Complex,Atomic City of PakistanHWR/Military/Tritium22 kg [55]1986–Present
 RussiaMayak Plant BMilitary4001948-196?[56]
 RussiaMayak Plant BB, RT-1LWR[45]PUREX + Np separation[57]400[42]1978[42]
 RussiaZheleznogorsk (Krasnoyarsk-26), RT-2VVER1,500[42]under construction
 USA,WAHanford SiteMilitarybismuth phospate, REDOX, PUREX1944–present[58]
 USASCSavannah River SiteMilitary/LWR/HWR/TritiumPUREX, REDOX, THOREX, Np separation5000[59]1952–2002
 USANYWest ValleyLWR[45]PUREX300[42]1966–1972[42]


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