During this era, Baghdad stood as the Islamic world's foremost hub of intellectual activity. The Abbasid
leaders in Baghdad quickly recognized their populace's limited
understanding in fields like astronomy, mathematics, and medicine. They
turned their attention to India and Persia for advanced knowledge.
Possession of Sind
provided the Abbasids with a crucial pathway to access Indian
expertise. This period saw the visit of an Indian
astronomer-mathematician and diplomat from Sind, Kanaka, to Caliph Al-Mansur's court (754–775). Intrigued by Indian astronomy and mathematics, the caliph instructed Ibrahim al-Fazari and Yaqub ibn Tariq to translate Brahmagupta's significant texts, Brahmasphutasiddhanta and Khandakhadyaka. These translations, named Sindhind and Arkand, introduced the concept of Indian numerals
to the Islamic world. Similarly, Persian astronomical tables influenced
by Indian astronomy, Zig-I shahriyarr, were translated into Arabic as
Zijashshahriyar. The ninth-century scholar al-Khwarizmi, who learned Sanskrit, played a pivotal role in disseminating the Indian numeral system globally. Another contemporary scholar, al-Kindi, authored four books on Indian numerals.
Indian medical practices and pharmaceuticals were also highly
sought after in the Islamic world. Numerous Sanskrit medical texts were
translated into Arabic, sponsored by Khalid ibn Barmak, Al-Mansur's vizier. Khalid, originally from a Buddhist family in Balkh, converted to Islam after the Arab conquest. His family, known as the Barmakis of Baghdad, showed a keen interest in Indian innovations. Under Caliph Harun al-Rashid (788–809), the translation of Susruta Samhita into Arabic was commissioned. Furthermore, the notable Arabic medical work Kitab Al-Hawi, later translated into Latin as Liber continens in the 13th century, was penned by al-Razi, or Rhazes (865–925), incorporating substantial Indian medical knowledge.
History
For the best part of a millennium, from the Seleucid era and through to the Sassanid period, there had been an exchange of scholarship between the Greek, Persian and Indian cultural spheres. The origin of the number zero and the place-value system notably falls into this period; its early use originates in Indian mathematics of the 5th century (Lokavibhaga), influencing Sassanid era Persian scholars during the 6th century.
The sudden Islamic conquest of Persia
in the 640s drove a wedge between the Mediterranean and Indian
traditions, but scholarly transfer soon resumed, with translations of
both Greek and Sanskrit works into Arabic during the 8th century. This
triggered the flourishing of Abbasid-era scholarship centered in Baghdad in the 9th century, and the eventual resumption of transmission to the west via Muslim Spain and Sicily by the 10th century.
There was continuing contact between Indian and Perso-Arabic scholarship during the 9th to 11th centuries while the Muslim conquest of India was temporarily halted.
Al Biruni in the early 11th century traveled widely in India and became an important source of knowledge about India in the Islamic world during that time.
With the establishment of the Delhi Sultanate in the 13th century, northern India fell under Perso-Arabic dominance and the native Sanskrit tradition fell into decline, while at about the same time the "Golden Age of Islam" of the Arab caliphates gave way to Turko-Mongol dominance, leading to the flourishing of a secondary "Golden Age" of Turko-Persian literary tradition during the 13th to 16th centuries, exemplified on either side of Timurid Persia by the Ottoman Empire in the west and the Mughal Empire in the east.
Astronomy
The mathematical astronomy text Brahmasiddhanta of Brahmagupta (598-668) was received in the court of Al-Mansur (753–774). It was translated by Alfazari into Arabic as Az-Zīj ‛alā Sinī al-‛Arab, popularly called Sindhind. This translation was the means by which the Hindu numerals were transmitted from the Sanskrit to the Arabic tradition. According to Al-Biruni,
As Sindh was under the actual rule
of the Khalif Mansur (AD 753–774), there came embassies from that part
of India to Bagdad and among them scholars, who brought with them two
books. With the help of these Pandits Alfazari, perhaps also Yaqūb ibn Tāriq,
translated them. Both works have been largely used, and have exercised a
great influence. It was on this occasion that the Arabs first became
acquainted with a scientific system of astronomy. They learned from
Brahmagupta earlier than Ptolemy.
— Alberuni (Ed. & trans. Edward Sachau), Alberuni's India [The Indika of Alberuni] (1910)
Alberuni's translator and editor Edward Sachau wrote: "It is Brahmagupta who taught Arabs mathematics before they got acquainted with Greek science."
Al-Fazari also translated the Khandakhadyaka (Arakand) of Brahmagupta.
Through the resulting Arabic translations of Sindhind and Arakand, the use of Indian numerals became established in the Islamic world.
Mathematics
The etymology of the word "sine" comes from the Latin mistranslation of the word jiba, which is an Arabic transliteration of the Sanskrit word for half the chord, jya-ardha.
The sin and cos functions of trigonometry, were important mathematical concepts, imported from the Gupta period of Indian astronomy namely the jyā and koṭi-jyā functions via translation of texts like the Aryabhatiya and Surya Siddhanta, from Sanskrit to Arabic, and then from Arabic to Latin, and later to other European languages.
Al-Khowarizmi
(ca. 840) contributed a work on algebra and an account of the
Hindu—Arabic numerals including the use of zero as a place-holder...the
history of early Hindu mathematics has always presented considerable
problems for the West...it is still not possible to form a clear picture
of either method or motivation in Hindu mathematics...this, together
with the absence of any formalised proof structure, militated against
continuous mathematical development...much of the Hindu approach to
mathematics was certainly conveyed to western Europe through Arabs. The
Algebraic method formerly considered to have been invented by Al Khowarizimi
can now be seen to stem from Hindu sources. The place-value system
involving the use of nine numerals and a zero as place-holder is
undoubtedly of Hindu origin and its transmission to the West had a
profound influence on the whole course of mathematics.
As in the rest of mathematical
science so in Trigonometry, were the Arabs pupils of the Hindus and
still more of the Greeks, but not without important devices of their
own.
For over five hundred years Arabic writers and others continued to apply to works on arithmetic the name Indian.
Another important early treatise that publicized decimal numbers was the Iranian mathematician and astronomer Kushyar ibn Labban's leading arithmetic book Kitab fi usul hisab al-hind (principals of Hindu reckoning).
Abu'l-Hasan al-Uqlidisi a scholar in the Abbasid caliphate wrote al-Fusul fi al-Hisab al-Hindi ("chapters in Indian calculation") to address the difficulty in procedures for calculation from the Euclid's Elements
and endorsed the use of Indian calculation. He highlighted its ease of
use, speed, fewer requirements of memory and the focused scope on the
subject.
Medical texts
Manka, an Indian physician at the court of Harun al-Rashid translated the Sushruta (the classical (Gupta-era) Sanskrit text on medicine) into Persian.
A large number of Sanskrit medical,
pharmacological and toxicological texts were translated into Arabic
under the patronage of Khalid, the vizier of Al-Mansur. Khalid was the
son of a chief priest of a Buddhist monastery at Balkh. Some of his
family was killed when the Arabs captured Balkh; others including Khalid survived by converting to Islam. They were to be known as the Barmakids
of Baghdad who were fascinated by the new ideas from India. Indian
medical knowledge was given a further boost under the Caliph Harun al
Rashid (788–809) who ordered the translation of Susruta Samhita into Arabic.
We know of Yahya ibn Khalid al Barmaki
(805) as a patron of physicians and, specifically, of the translation
of Hindu medical works into both Arabic and Persian. In all likelihood
however, his activity took place in the orbit of the caliphate court in
Iraq, where at the behest of Harun al Rashid (786–809), such books were
translated into Arabic. Thus Khurasan and Transoxania were effectively
bypassed in this transfer of learning from India to Islam, even though,
undeniably the Barmakis cultural outlook owed something to their land of
origin, northern Afghanistan, and Yahya al Barmaki's interest in
medicine may have derived from no longer identifiable family tradition.
The Caraka Saṃhitā was translated into Persian and subsequently into Arabic by Abd-Allah ibn Ali in the ninth century.
Probably the first Islamic hospital (Bimaristan or Maristan) was established in Baghdad Yahya ibn Khalid ibn Barmak, tutor and subsequently vizier of Harun al-Rashid when the latter became Khalif
in 786. Yahya ibn Khalid ibn Barmak's hospital, usually referred to as
the Barmakid Hospital must have been established before 803, the year in
which the Barmakid family fell from power. The hospital is mentioned in
two places in the Fihrist
which was written in 997. Ibn Dahn, Al Hindi, who administered the
Bimaristan of the Barmak. He translated from the Indian language into
Arabic. Yahya ibn Khalid ordered Mankah (Kankah), the Indian, to
translate it (an Indian book of medicine) at the hospital to render it
in the form of a compilation.
Al-Razi's Al-Hawi (liber continens) of c. 900 is said to contain "much Indian knowledge" from texts such as the Susruta Samhita.
Geography
The Indian geographical knowledge that was transmitted and influenced the Arabs included the view of Aryabhata
that the apparent daily rotation of the heavens was caused by the
rotation of the earth on its own axis, the idea that the proportion of
land and sea on the surface of the earth was half and half and the land
mass as being dome shaped and covered on all sides by water.
The Arabs utilized the Indian cartographic system in which the
northern hemisphere was considered to be the inhabited part of the earth
and divided into nine parts. Its four geographical limits were djamakut
in the east, rum in the west, Ceylon as the cupola (dome) and Sidpur.
Indians believed that the prime meridian passes through ujjain
and calculated their longitudes from Ceylon. The Arabs adopted this idea
of Ceylon's being the cupola of the earth but later mistakenly believed
ujjain to be the cupola.
Nuclear reprocessing is the chemical separation of fission products and actinides from spent nuclear fuel. Originally, reprocessing was used solely to extract plutonium for producing nuclear weapons. With commercialization of nuclear power, the reprocessed plutonium was recycled back into MOX nuclear fuel for thermal reactors. The reprocessed uranium,
also known as the spent fuel material, can in principle also be re-used
as fuel, but that is only economical when uranium supply is low and
prices are high. Nuclear reprocessing may extend beyond fuel and include
the reprocessing of other nuclear reactor material, such as 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.
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. 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.
re-use for zircalloy cladding or storage as intermediate level waste
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 risk of nuclear weapons proliferation by diversion of plutonium from the civilian fuel cycle, and to encourage other nations to follow the US 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.
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. 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.
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 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 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. 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 UNiversalEXtraction 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 cobaltcarboraneanion (known as chlorinated cobalt dicarbollide). The actinides are extracted by CMPO, and the diluent is a polar aromatic such as nitrobenzene. Other diluents such as meta-nitrobenzotrifluoride and phenyl trifluoromethyl sulfone have been suggested as well.
Electrochemical and ion exchange methods
An exotic method using electrochemistry and ion exchange in ammoniumcarbonate has been reported.
Other methods for the extraction of uranium using ion exchange in
alkaline carbonate and "fumed" lead oxide have also been reported.
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) to produce PuO22+. The plutonium was maintained in the +6 oxidation state by addition of a 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.
Hexone or REDOX
This is a liquid-liquid extraction process which uses methyl isobutyl ketone codenamed hexone 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 was used in 1952-1956 on the Hanford plant T and has been replaced by the PUREX process.
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 nitrate)
to increase the nitrate concentration in the aqueous phase to obtain a
reasonable distribution ratio. This process was used at Windscale
in 1951-1964. This process has been replaced by PUREX, which was shown
to be a superior technology for larger scale reprocessing.
Sodium acetate
The sodium uranyl acetate process was used by the early Soviet nuclear industry to recover plutonium from irradiated fuel. It was 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.
Pyroprocessing
The most developed, though commercially unfielded, alternative reprocessing method, is Pyroprocessing, suggested as part of the depicted metallic-fueled, Integral fast reactor (IFR) a sodium fast reactor concept of the 1990s. After the spent fuel is dissolved in molten salt, all of the recyclable 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.
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.
Alternatively, voloxidation (see below)
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 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 isotopes of palladium.
Palladium derived from the decay of fission ruthenium and rhodium will
be nonradioactive, but fission Palladium contains significant
contamination with long-lived 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 radioisotope thermoelectric generators
(RTGs) that are mostly decayed in spent fuel, that has significantly
aged, can be recovered in sufficient quantities to make their use
worthwhile. Examples include materials with half lives around two years
such as 134 Cs, 125 Sb, 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.
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
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 pyrometallurgical 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.
Working with "fresh" spent fuel requires more shielding and better
ways to deal with heat production than working with "aged" spent fuel
does. If the facilities are built in such a way as to require high specific activity material, they cannot handle older "legacy waste" except blended with fresh spent fuel
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.
Experimental electro refinement cell at Argonne National Laboratory
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 anodebasket
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. 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 moltenalloy 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 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. Tritium is a difficult contaminant to
remove from aqueous solution, as it cannot be separated from water
except by isotope separation. However, tritium is also a valuable
product used in industry science and 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
The process is simple and requires no complex machinery or chemicals above and beyond that required in all reprocessing (hot cells, remote handling equipment)
Products such as krypton-85 or tritium, as well as xenon (whose isotope are either stable, very nearly stable, or quickly decay), can be recovered and sold for use in industry, science or medicine
Driving off volatile fission products allows for safer storage in interim storage or deep geological repository
Radioactive material is not chemically mobilized beyond what should
be accounted for in long-term storage anyway. Substances that are inert
as native elements or oxides remain so
The product can be used as fuel in a CANDU reactor or even downblended with similarly treated spent CANDU fuel if too much fissile material is left in the spent fuel.
The resulting product can be further processed by any of the other
processes mentioned above and below. Removal of volatile fission
products means that transportation becomes slightly easier compared to
spent fuel with damaged or removed cladding
All volatile products of concern (while helium will be present in the spent fuel, there won't be any radioactive isotopes of helium) can in principle be recovered in a cold trap cooled by liquid nitrogen
(temperature: 77 K (−196.2 °C; −321.1 °F) or lower). However, this
requires significant amounts of cooling to counteract the effect of
decay heat from radioactive volatiles like krypton-85. Tritium will be
present in the form of tritiated water, which is a solid at the temperature of liquid nitrogen.
Technetium heptoxide
can be removed as a gas by heating above its boiling point of 392.6 K
(119.5 °C; 247.0 °F) which reduces the issues presented by Technetium
contamination in processes like fluoride volatility or PUREX; ruthenium tetroxide (gaseous above 313.1 K (40.0 °C; 103.9 °F)) can likewise be removed from the spent fuel and recovered for sale or disposal
Disadvantages
Further processing is needed if the resulting product is to be used for re-enrichment or fabrication of MOX-fuel
If volatile fission products escape to the environment this presents a radiation hazard, mostly due to 129 I, Tritium and 85 Kr. Their safe recovery and storage requires further equipment.
An oxidizing agent / reducing agent has to be used for reduction/oxidation steps whose recovery can be difficult, energy consuming or both
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:
Can in theory be done "self heating" via the decay heat of sufficiently "fresh" spent fuel
Caesium-137 has uses in food irradiation and can be used to power radioisotope thermoelectric generators. However, its contamination with stable 133 Cs and long lived 135 Cs reduces efficiency of such uses while contamination with 134 Cs in relatively fresh spent fuel makes the curve of overall radiation and heat output much steeper until most of the 134 Cs has decayed
Can potentially recover elements like ruthenium
whose ruthenate ion is particularly troublesome in PUREX and which has
no isotopes significantly longer lived than a year, allowing possible
recovery of the metal for use
A "third phase recovery" can be added to the process if substances
that melt but don't vaporize at the temperatures involved are drained to
a container for liquid effluents and allowed to re-solidify. To avoid
contamination with low-boiling products which melt at low temperatures, a
melt plug could be used to open the container for liquid effluents only
once a certain temperature is reached by the liquid phase.
Strontium, which is present in the form of the particularly troublesome mid-lived fission product 90 Sr is liquid above 1,050 K (780 °C; 1,430 °F). However, Strontium oxide
remains solid below 2,804 K (2,531 °C; 4,588 °F) and if strontium oxide
is to be recovered with other liquid effluents, it has to be reduced to the native metal before the heating step. Both Strontium and Strontium oxide form soluble Strontium hydroxide and hydrogen upon contact with water, which can be used to separate them from non-soluble parts of the spent fuel.
As there are little to no chemical changes in the spent fuel, any
chemical reprocessing methods can be used following this process
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 thermal stress. This combines with radiation damage to material and possibly neutron embrittlement if neutron sources such as 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.
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 decayKr-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 (628 °F) and 114.1 °C
(237.4 °F). 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.
Chlorine (and to a lesser extent fluorine) is a readily available industrial chemical that is produced in mass quantity
Fractional distillation allows many elements to be separated from
each other in a single step or iterative repetition of the same step
Uranium will be produced directly as Uranium hexafluoride, the form used in enrichment
Many volatile fluorides and chlorides are volatile at relatively
moderate temperatures reducing thermal stress. This is especially
important as the boiling point of uranium hexafluoride is below that of
water, allowing to conserve energy in the separation of high boiling
fission products (or their fluorides) from one another as this can take
place in the absence of uranium, which makes up the bulk of the mass
Some fluorides and chlorides melt at relatively low temperatures
allowing a "liquid phase separation" if desired. Those low melting salts
could be further processed by molten salt electrolysis.
Fluorides and chlorides differ in water solubility depending on the
cation. This can be used to separate them by aqueous solution. However,
some fluorides violently react with water, which has to be taken into
account.
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
thousands of tons of uranium would require smaller facilities than
processing billions of 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 alpha emitters induces some (α,n) reactions in fluorine, producing both radioactive 22 Na and neutrons.
This effect can be reduced by separating alpha emitters and fluorine as
fast as feasible. Interactions between chlorine's two stable isotopes 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.
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, 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 was the focus of much debate over the first
decade of the 2000s. 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 agreed that under then-current economic conditions the
reprocessing-recycle option is the more costly one. 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.
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.
