Nuclear reprocessing

Source: Wikipedia, the free encyclopedia.

Sellafield nuclear reprocessing site, UK

Nuclear reprocessing is the chemical separation of

Zircaloy
cladding.

The high

radioactivity of spent nuclear material means that reprocessing must be highly controlled and carefully executed in advanced facilities by specialized personnel. Numerous processes exist, with the chemical based PUREX process dominating. Alternatives include heating to drive off volatile elements, burning via oxidation, and fluoride volatility (which uses extremely reactive Fluorine). Each process results in some form of refined nuclear product, with radioactive waste as a byproduct. Because this could allow for weapons grade nuclear material, nuclear reprocessing is a concern for nuclear proliferation
and is thus tightly regulated.

Relatively high cost is associated with spent fuel reprocessing compared to the once-through fuel cycle, but fuel use can be increased and waste volumes decreased.[3] Nuclear fuel reprocessing is performed routinely in Europe, Russia, and Japan. 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.[4] Not all nuclear fuel requires reprocessing; 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.[5][6]

Separated components and disposition

See also: Spent nuclear fuel

The potentially useful components dealt with in nuclear reprocessing comprise specific

cladding
.

material disposition
plutonium, minor actinides, reprocessed uranium
fast, fusion fission hybrid, or subcritical reactor or use as MOX fuel
reprocessed uranium, filters less stringent storage as
intermediate-level waste
long-lived fission and activation products
geological repository
137Cs and 90Sr
 medium-term storage as high-level waste; decay heat could be used to drive a Stirling engine
useful radionuclides,
rare earths (lanthanides), and noble metals
 industrial and medical uses
cladding, fission product zirconium re-use for
zircalloy
cladding or storage as intermediate level waste

History

The first large-scale nuclear reactors were built during

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.

West Valley Reprocessing Plant which closed by 1972 because of its inability to meet new regulatory requirements.[8]

Reprocessing of civilian fuel has long been employed at the

West Valley Reprocessing Plant
in the United States.

In October 1976,[9] 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 risk of nuclear weapons proliferation by diversion of plutonium from the civilian fuel cycle, and to encourage other nations to follow the US lead.[10][11][12] 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.[13]

In March 1999, the

Fukushima Daiichi.[15]

Separation technologies

Water and organic solvents

PUREX

Main article: 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, 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 considered "weapons-grade" plutonium, ideal for use in a nuclear weapon. Nevertheless, highly reliable nuclear weapons can be built at all levels of technical sophistication using reactor-grade plutonium.[16] Moreover, reactors that are capable of refueling frequently can be used to produce weapon-grade plutonium, which can later be recovered using PUREX. Because of this, PUREX chemicals are monitored.[17]

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.[citation needed
]

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 US 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 (DIAMide EXtraction) process has the advantage of avoiding the formation of organic waste which contains elements other than

CEA. The process is sufficiently mature that an industrial plant could be constructed with the existing knowledge of the process.[18]
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

bis-triazinyl pyridine (BTP) based process.[19][20][21]
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

aromatic such as nitrobenzene. Other diluents such as meta-nitrobenzotrifluoride and phenyl trifluoromethyl sulfone[26]
have been suggested as well.

Electrochemical and ion exchange methods

An exotic method using electrochemistry and ion exchange in ammonium carbonate has been reported.[27] Other methods for the extraction of uranium using ion exchange in alkaline carbonate and "fumed" lead oxide have also been reported.[28]

Obsolete methods

Bismuth phosphate

The

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

dichromate
salt.

The bismuth phosphate was next re-precipitated, leaving the plutonium in solution, and an iron(II) salt (such as

ferrous sulfate) was added. The plutonium was again re-precipitated using a bismuth phosphate carrier and a combination of lanthanum salts and fluoride added, forming a solid lanthanum fluoride carrier for the plutonium. Addition of an alkali produced an oxide. The combined lanthanum plutonium oxide was collected and extracted with nitric acid to form plutonium nitrate.[29]

Hexone or REDOX

This is a liquid-liquid extraction process which uses

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 was used in 1952-1956 on the Hanford plant T and has been replaced by the PUREX process.[30][31]

Pu4+ + 4NO−3 + 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

Windscale in 1951-1964. This process has been replaced by PUREX, which was shown to be a superior technology for larger scale reprocessing.[32]

Sodium acetate

The sodium uranyl acetate process was used by the early Soviet nuclear industry to recover plutonium from irradiated fuel.[33] It was a never used in the West; the idea is to dissolve the fuel in nitric acid, alter the oxidation state of the plutonium, and then add acetic acid and base. This would convert the uranium and plutonium into a solid acetate salt.

Explosion of the crystallized acetates-nitrates in a non-cooled waste tank caused the Kyshtym disaster in 1957.

Alternatives to PUREX

As there are some downsides to the PUREX process, there have been efforts to develop alternatives to the process, some of them compatible with PUREX (i.e. the residue from one process could be used as feedstock for the other) and others wholly incompatible. None of these have (as of the 2020s) reached widespread commercial use, but some have seen large scale tests or firm commitments towards their future larger scale implementation.[34]

Pyroprocessing

actinides, consisting largely of plutonium and uranium though with important minor constituents, are extracted using electrorefining/electrowinning. The resulting mixture keeps the plutonium at all times in an unseparated gamma and alpha emitting actinide form, that is also mildly self-protecting in theft scenarios.[36]

Electrorefining, distillation
, and solvent-solvent extraction are common steps.

These processes are not currently in significant use worldwide, but they have been pioneered at Argonne National Laboratory[37][38] with current research also taking place at CRIEPI in Japan, the Nuclear Research Institute of Řež in Czech Republic, Indira Gandhi Centre for Atomic Research in India and KAERI in South Korea.[39][40][41][42]

Advantages of pyroprocessing

  • The principles behind it are well understood, and no significant technical barriers exist to their adoption.[43]
  • 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
    fission product tritium and the activation product carbon-14
    in dilute solutions that cannot be separated later.
  • 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
    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.[45]) 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 increases energy production per kg of fuel, as well as reducing the long-term radioactivity of the wastes.
  • Fluoride volatility (see below) produces salts that can readily be used in molten salt reprocessing such as pyroprocessing
  • The ability to process "fresh" spent fuel reduces the needs for spent fuel pools (even if the recovered short lived radionuclides are "only" sent to storage, that still requires less space as the bulk of the mass, uranium, can be stored separately from them). Uranium – even higher specific activity reprocessed uranium – does not need cooling for safe storage.
  • Short lived radionuclides can be recovered from "fresh" spent fuel allowing either their direct use in industry science or medicine or the recovery of their decay products without contamination by other isotopes (for example: ruthenium in spent fuel decays to rhodium all isotopes of which other than 103
    Rh     further decay to stable
    107
    Pd
    . Ruthenium-107 and rhodium-107 both have half lives on the order of minutes and decay to palladium-107 before reprocessing under most circumstances)
  • Possible fuels for
    147
    Pm. While those would perhaps not be suitable for lengthy space missions, they can be used to replace diesel generators in off-grid locations where refueling is possible once a year.[a] Antimony would be particularly interesting because it forms a stable alloy with lead and can thus be transformed relatively easily into a partially self-shielding and chemically inert form. Shorter lived RTG fuels present the further benefit of reducing the risk of orphan sources
    as the activity will decline relatively quickly if no refueling is undertaken.

Disadvantages of pyroprocessing

Electrolysis

The electrolysis methods are based on the difference in the

standard potentials of uranium, plutonium and minor actinides in a molten salt. The standard potential of uranium is the lowest, therefore when a potential is applied, the uranium will be reduced at the cathode out of the molten salt solution before the other elements.[46]

Experimental electro refinement cell at Argonne National Laboratory

PYRO-A and -B for IFR

These processes were developed by

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 electric 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.[47][48][49] As alternatives to the molten cadmium electrode it is possible to use a molten bismuth cathode, or a solid aluminium cathode.[50]

As an alternative to electrowinning, the wanted metal can be isolated by using a

electropositive metal and a less reactive metal.[51]

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.[52]

The mixed actinides produced by pyrometallic processing can be used again as nuclear fuel, as they are virtually all either

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

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

nuclear weapons, so recovery of a stream of hydrogen or water with a high tritium content can make targeted recovery economically worthwhile. 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.

Advantages

Disadvantages

Volatilization in isolation

Simply heating spent oxide fuel in an inert atmosphere or vacuum at a temperature between 700 °C (1,292 °F) and 1,000 °C (1,830 °F) 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 has a similar half-life). The estimated overall mass balance for 20,000 g of processed fuel with 2,000 g of cladding is:[53]

Input Residue Zeolite
filter		 Carbon
filter		 Particle
filters
Palladium 28 14 14
Tellurium 10 5 5
Molybdenum 70 70
Caesium 46 46
Rubidium 8 8
Silver 2 2
Iodine 4 4
Cladding 2000 2000
Uranium 19218 19218 ?
Others 614 614 ?
Total 22000 21851 145 4 0

Advantages

Disadvantages

  • At temperatures above 1,000 K (730 °C; 1,340 °F) the native metal form of several actinides, including neptunium (melting point: 912 K (639 °C; 1,182 °F)) and plutonium (melting point: 912.5 K (639.4 °C; 1,182.8 °F)), are molten. This could be used to recover a liquid phase, raising proliferation concerns, given that uranium metal remains a solid until 1,405.3 K (1,132.2 °C; 2,069.9 °F). While neptunium and plutonium cannot be easily separated from each other by different melting points, their differing solubility in water can be used to separate them.
  • If "nuclear self heating" is employed, the spent fuel with have much higher specific activity, heat production and radiation release. If an external heat source is used, significant amounts of external power are needed, which mostly go to heat the uranium.
  • Heating and cooling the vacuum chamber and/or the piping and vessels to collect volatile effluents induces
    californium-252
    are present to a significant extent.
  • In the commonly used oxide fuel, some elements will be present both as oxides and as native elements. Depending on their chemical state, they may end up in either the volatalized stream or in the residue stream. If an element is present in both states to a significant degree, separation of that element may be impossible without converting it all to one chemical state or the other
  • The temperatures involved are much higher than the melting point of lead (600.61 K (327.46 °C; 621.43 °F)) which can present issues with radiation shielding if lead is employed as a shielding material
  • If filters are used to recover volatile fission products, those become
    low-
    to intermediate level waste.

Fluoride volatility

Main article: Fluoride volatility
Eu-155→Gd
visible.

In the fluoride volatility process,

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

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.

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-

tin tetrachloride have relatively low boiling points of 331 °C (628 °F) and 114.1 °C (237.4 °F). Chlorination has even been proposed as a method for removing zirconium fuel cladding,[44]
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

lanthanides, strontium, caesium are more soluble than those of uranium, neptunium, plutonium, and zirconium
.

Advantages of halogen volatility

Disadvantages of halogen volatility

  • Many compounds of fluorine or chlorine as well as the native elements themselves are toxic, corrosive and react violently with air, water or both
  • Uranium hexafluoride and Technetium hexafluoride have very similar boiling points (329.6 K (56.5 °C; 133.6 °F) and 328.4 K (55.3 °C; 131.4 °F) respectively), making it hard to completely separate them from one another by distillation.
  • Fractional distillation as used in
    petroleum refining
    requires large facilities and huge amounts of energy. To process tons of uranium would require similarly large facilities as processing tons of petroleum - however, unlike petroleum refineries, the entire process would have to take place inside radiation shielding and there would have to be provisions made to prevent leaks of volatile, poisonous and radioactive fluorides
  • Plutonium hexafluoride boils at 335 K (62 °C; 143 °F) this means that any facility capable of separating uranium hexafluoride from Technetium hexafluoride is capable of separating plutonium hexafluoride from either, raising proliferation concerns
  • The presence of
    35
    Cl and 37
    Cl     on the one hand and alpha particles on the other are of lesser concern as they do not have as high a cross section and do not produce neutrons or long lived radionuclides.[58]
  • If carbon is present in the spent fuel it'll form
    halogenated hydrocarbons which are extremely potent greenhouse gases
    , and hard to chemically decompose. Some of those are toxic as well.

Radioanalytical separations

To determine the distribution of radioactive metals for analytical purposes,

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).[59]

Economics

The relative economics of reprocessing-waste disposal and interim storage-direct disposal was the focus of much debate over the first decade of the 2000s. Studies[60] 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 agreed that under then-current economic conditions the reprocessing-recycle option is the more costly one.[61] While the uranium market - particularly its short term fluctuations - has only a minor impact on the cost of electricity from nuclear power, long-term trends in the uranium market do significantly affect the economics of nuclear reprocessing. If uranium prices were to rise and remain consistently high, "stretching the fuel supply" via MOX fuel, breeder reactors or even the thorium fuel cycle could become more attractive. However, if uranium prices remain low, reprocessing will remain less attractive.[citation needed
]

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%,

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.[64][65]

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.[66][67]

List of sites

Country Reprocessing site Fuel type Procedure Reprocessing
capacity tHM/yr 		Commissioning
or operating period
 Belgium Mol
LWR
, MTR (Material test reactor)		 80[68] 1966–1974[68]
 China intermediate pilot plant[69] 60–100 1968-early 1970s
 China Plant 404[70] 50 2004
 Germany Karlsruhe, WAK LWR[71] 35[68] 1971–1990[68]
 France Marcoule, UP 1 Military 1200[68] 1958[68]-1997[72]
 France Marcoule, CEA APM
FBR
 PUREX DIAMEX SANEX[73] 6[71] 1988–present[71]
 France La Hague, UP 2 LWR[71] PUREX[74] 900[68] 1967–1974[68]
 France La Hague, UP 2–400 LWR[71] PUREX[74] 400[68] 1976–1990[68]
 France
La Hague
, UP 2–800		 LWR PUREX[74] 800[68] 1990[68]
 France
La Hague
, UP 3		 LWR PUREX[74] 800[68] 1990[68]
 UK Windscale, B204 Magnox, LWR BUTEX 750[68] 1956–1962,[68] 1969-1973
 UK Sellafield, Magnox Reprocessing Plant Magnox,[71] LWR, FBR PUREX 1500[68] 1964[68]-2022
 UK Dounreay FBR[71] 8[68] 1980[68]
 UK
THORP
AGR
, LWR		 PUREX 900[68][75] 1994[68][75]-2018
 Italy Rotondella Thorium 5[68] 1968[68] (shutdown)
 India Trombay Military PUREX[76] 60[68] 1965[68]
 India Tarapur PHWR PUREX 100[68] 1982[68]
 India Kalpakkam PHWR and FBTR PUREX 100[77] 1998[77]
 India Tarapur PHWR 100[78] 2011[78]
 Israel Dimona Military 60–100[79] ~1960–present
 Japan Tokai LWR[80] 210[68] 1977[68]-2006[81]
 Japan Rokkasho LWR[71] 800[68][75] under construction (2024)[82]
 Pakistan New Labs, Rawalpindi Military/Plutonium/Thorium 80[83] 1982–present
 Pakistan
Atomic City of Pakistan
HWR/Military/Tritium
 22 kg[84] 1986–present
 Russia Mayak Plant B Military 400 1948-196?[85]
 Russia Mayak Plant BB, RT-1 LWR[71] PUREX + Np separation[86] 400[68][75] 1978[68]
 Russia
Tomsk-7
Radiochemical Plant		 Military 6000[79] 1956[87]
 Russia Zheleznogorsk (Krasnoyarsk-26) Military 3500[79] 1964–~2010[88]
 Russia Zheleznogorsk, RT-2 VVER 800[68] under construction (2030)
 USA, WA Hanford Site Military bismuth phosphate, REDOX, PUREX 1944–1988[89]
 USA, SC Savannah River Site Military/LWR/HWR/Tritium PUREX, REDOX, THOREX, Np separation 5000[90] 1952–2002
 
NY
West Valley
 LWR[71] PUREX 300[68] 1966–1972[68]
 USA, NC Barnwell LWR PUREX 1500 never permitted to operate[91]
 
ID
 INL LWR PUREX –

See also

References

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Notes

  1. ^ a radioisotope with a two year half life will retain 0.5^0.5 or over 70% of its power after a year - all those isotopes have half lives longer than two years and would thus retain even more power. Even if the yearly refueling window were to be missed, over half the power would still remain for the second refueling window

Further reading

External links

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