Kamis, 22 Agustus 2013

........ Warheads as a Source of Nuclear Fuel..??? >>> ... Weapons-grade uranium and plutonium surplus to military requirements in the USA and Russia is being made available for use as civil fuel. Weapons-grade uranium is highly enriched, to over 90% U-235 (the fissile isotope). Weapons-grade plutonium has over 93% Pu-239 and can be used, like reactor-grade plutonium, in fuel for electricity production. Highly-enriched uranium from weapons stockpiles has been displacing some 9720 tonnes of U3O8 production from mines each year, and meets about 13% of world reactor requirements through to 2013...>> ..... Over one third of the energy produced in most nuclear power plants comes from plutonium. It is created in the reactor as a by-product. Plutonium has occurred naturally, but except for trace quantities it is not now found in the Earth's crust. There are several tonnes of plutonium in our biosphere, a legacy of atmospheric weapons testing in the 1950s and 1960s.....>> ...Plutonium is formed in nuclear power reactors from uranium. When operating, a typical 1000 MWe nuclear power reactor contains within its uranium fuel load several hundred kilograms of plutonium....>>> All plutonium isotopes are fissionable with fast neutrons, though only two are fissile (with slow neutrons). For this reason all are significant in a fast neutron reactor (FNR), but only one – Pu-239 - has a major role in a conventional light water power reactor....>>> Pu-238, (half-lifea 88 years, alpha decay to U-234) Pu-239, fissile (half-life 24,000 years, alpha decay to U-235) Pu-240, fertile (half-life 6,560 years, alpha decay to U-236) Pu-241, fissile (half-life 14.4 years, beta decay to Am-241) Pu-242, (half-life 374,000 years, alpha decay to U-238) (Periodic tables show an atomic mass of 244 for plutonium, suggesting Pu-244 as the most stable isotope with the longest half-life - 82 million years. It is the only one found in trace quantities in nature, apparently cosmogenic in origin from the formation of the Earth. It is not very relevant to this paper. It alpha decays to U-240.)..>>> Used nuclear fuel has long been reprocessed to extract fissile materials for recycling and to reduce the volume of high-level wastes. New reprocessing technologies are being developed to be deployed in conjunction with fast neutron reactors which will burn all long-lived actinides. A significant amount of plutonium recovered from used fuel is currently recycled into MOX fuel; a small amount of recovered uranium is recycled...>>. Powerful accelerators can produce neutrons by spallationa. This process may be linked to conventional nuclear reactor technology in an accelerator-driven system (ADS) to transmute long-lived radioisotopes in used nuclear fuel into shorter-lived fission products. There is also increasing interest in the application of ADSs to running subcritical nuclear reactors powered by thorium.....>>

Plutonium

(Updated March 2012)
  • Over one third of the energy produced in most nuclear power plants comes from plutonium. It is created in the reactor as a by-product.
  • Plutonium has occurred naturally, but except for trace quantities it is not now found in the Earth's crust.
  • There are several tonnes of plutonium in our biosphere, a legacy of atmospheric weapons testing in the 1950s and 1960s.
In practical terms, there are two different kinds of plutonium to be considered: reactor-grade and weapons-grade. The first is recovered as a by-product of typical used fuel from a nuclear reactor, after the fuel has been irradiated ('burned') for about three years. The second is made specially for the military purpose, and is recovered from uranium fuel that has been irradiated for only 2-3 months in a plutonium production reactor. The two kinds differ in their isotopic composition but must both be regarded as a potential proliferation risk, and managed accordingly.
Plutonium, both that routinely made in power reactors and that from dismantled nuclear weapons, is a valuable energy source when integrated into the nuclear fuel cycle. In a conventional nuclear reactor, one kilogram of Pu-239 can produce sufficient heat to generate nearly 10 million kilowatt-hours of electricity.

Plutonium and nuclear power

Plutonium is formed in nuclear power reactors from uranium. When operating, a typical 1000 MWe nuclear power reactor contains within its uranium fuel load several hundred kilograms of plutonium.

Reaction in standard UO2 fuel

Like all other heavy elements, plutonium has a number of isotopes, differing in the number of neutrons in the nucleus. All 15 plutonium isotopes are radioactive, because they are to some degree unstable and therefore decay, emitting particles and some gamma radiation as they do so.
All plutonium isotopes are fissionable with fast neutrons, though only two are fissile (with slow neutrons). For this reason all are significant in a fast neutron reactor (FNR), but only one – Pu-239 - has a major role in a conventional light water power reactor.
The main isotopes of plutonium are:
  • Pu-238, (half-lifea 88 years, alpha decay to U-234)
  • Pu-239, fissile (half-life 24,000 years, alpha decay to U-235)
  • Pu-240, fertile (half-life 6,560 years, alpha decay to U-236)
  • Pu-241, fissile (half-life 14.4 years, beta decay to Am-241)
  • Pu-242, (half-life 374,000 years, alpha decay to U-238)
  • (Periodic tables show an atomic mass of 244 for plutonium, suggesting Pu-244 as the most stable isotope with the longest half-life - 82 million years. It is the only one found in trace quantities in nature, apparently cosmogenic in origin from the formation of the Earth. It is not very relevant to this paper. It alpha decays to U-240.)
The most common isotope formed in a typical nuclear reactor is the fissile Pu-239 isotope, formed by neutron capture from U-238 (followed by beta decay), and which yields much the same energy as the fission of U-235. Well over half of the plutonium created in the reactor core is 'burned' in situ and is responsible for about one third of the total heat output of a light water reactor (LWR). Of the rest, about one sixth through neutron capture becomes Pu-240 (and Pu-241). The approximately 1.15% of plutonium in the spent fuel removed from a commercial LWR power reactor (burn-up of 42 GWd/t) consists of about 53% Pu-239, 25% Pu-240, 15% Pu-241, 5% Pu-242 and 2% of Pu-238, which is the main source of heat and radioactivity.b
Examples of the types of variation in plutonium composition produced from different sources 1  
Reactor type Mean fuel burn-up
(MW d/t)
Percentage of Pu isotopes at discharge Fissile content
%
Pu-238 Pu-239 Pu-240 Pu-241 Pu-242
PWR 33000 1.3 56.6 23.2 13.9 4.7 70.5
43000 2.0 52.5 24.1 14.7 6.2 67.2
53000 2.7 50.4 24.1 15.2 7.1 65.6
BWR 27500 2.6 59.8 23.7 10.6 3.3 70.4
30400 N/A 56.8 23.8 14.3 5.1 71.1
CANDU 7500 N/A 66.6 26.6 5.3 1.5 71.9
AGR 18000 0.6 53.7 30.8 9.9 5.0 63.6
Magnox 3000 0.1 80 16.9 2.7 0.3 82.7
5000 N/A 68.5 25.0 5.3 1.2 73.8

Plutonium-240 is the second most common isotope, formed by occasional neutron capture by Pu-239. Its concentration in nuclear fuel builds up steadily, since it does not undergo fission to produce energy in the same way as Pu-239. (In a fast neutron reactor it is fissionablec, which means that such a reactor can utilise recycled plutonium more effectively than a LWR.) While of a different order of magnitude to the fission occurring within a nuclear reactor, Pu-240 has a relatively high rate of spontaneous fission with consequent neutron emissions. This makes reactor-grade plutonium entirely unsuitable for use in a bomb (see section on Plutonium and weapons below). Reactor-grade plutonium is defined as that with 19% or more of Pu-240.
Plutonium-238, Pu-240 and Pu-242 emit neutrons as a few of their nuclei spontaneously fission, albeit at a low rate. They and Pu-239 also decay, emitting alpha particles and heat. The decay heat of Pu-238 (0.56 W/g) enables its use as an electricity source in the radioisotope thermoelectric generators (RTGs) of some cardiac pacemakers, space satellites, navigation beacons, etc. Plutonium has powered 24 US space vehicles and enabled the Voyager spacecraft to send back pictures of distant planets. These spacecraft have operated for 20 years and may continue for another 20. The Cassini spacecraft carries three generators providing 870 watts power as it orbits around Saturn.
A 1000 MWe light water reactor gives rise to about 25 tonnes of used fuel a year, containing up to 290 kilograms of plutonium. If the plutonium is extracted from used reactor fuel it can be used as a direct substitute for U-235 in the usual fuel, the Pu-239 being the main fissile part, but Pu-241 also contributing. In order to extract it for recycle, the used fuel is reprocessed and the recovered plutonium oxide is mixed with depleted uranium oxide to produce MOX fuel, with about 8% Pu-239 (this corresponds with uranium enriched to 5% U-235; see page on Mixed Oxide (MOX) Fuel).
Plutonium can also be used in fast neutron reactors, where all the plutonium isotopes fission, and so function as a fuel. As with uranium, the energy potential of plutonium is more fully realised in a fast reactor. Four of the six 'Generation IV' reactor designs currently under development are fast neutron reactors and will thus utilize plutonium in some way (see page on Generation IV Nuclear Reactors). In these, plutonium production will take place in the core, where burn-up is high and the proportion of plutonium isotopes other than Pu-239 will remain high.
Developments under the Global Nuclear Energy Partnership (GNEP) make it very likely that the some military plutonium will be used in fast reactors in the USA (see page on Global Nuclear Energy Partnership).
Plutonium stored over several years becomes contaminated with the Pu-241 decay product americium-241 (see page on Smoke Detectors and Americium), which interferes with normal fuel fabrication procedures. After long storage, Am-241 must be removed before the plutonium can be used in a MOX fuel fabrication plant because it emits intense gamma radiation (in the course of its alpha decay to Np-237).
In commercial power plants and research applications, plutonium generally exists as plutonium oxide (PuO2), a stable ceramic material with an extremely low solubility in water and with a high melting point (2,390 ºC). In pure form plutonium exists in six allotropic forms or crystal structure - more than any other element. As temperature changes, it switches forms - each has significantly different mechanical and electrical properties. One is nearly twice the density of lead (19.8 g/cm3). It melts at 640°C into a very corrosive liquid. The alpha phase is hard and brittle, like cast iron, and if finely divided it spontaneously ignites in air to form PuO2. Beta, gamma and delta phases are all less dense. Alloyed with gallium, plutonium becomes more workable.
Apart from its formation in today's nuclear reactors, plutonium was formed by the operation of naturally-occurring nuclear reactors in uranium deposits at Oklo in what is now west Africa, some two billion years ago.2

Plutonium and weapons 

It takes about 10 kilograms of nearly pure Pu-239 to make a bomb. Producing this requires 30 megawatt-years of reactor operation, with frequent fuel changes and reprocessing of the 'hot' fuel. Hence 'weapons-grade' plutonium is made in special production reactors by burning natural uranium fuel to the extent of only about 100 MWd/t (effectively three months), instead of the 45,000 MWd/t typical of LWR power reactors. Allowing the fuel to stay longer in the reactor increases the concentration of the higher isotopes of plutonium, in particular the Pu-240 isotope. For weapons use, Pu-240 is considered a serious contaminant, due to higher neutron emission and higher heat production. It is not feasible to separate Pu-240 from Pu-239. An explosive device could be made from plutonium extracted from low burn-up reactor fuel (i.e. if the fuel had only been used for a short time), but any significant proportions of Pu-240 in it would make it hazardous to the bomb makers, as well as probably unreliable and unpredictable. Typical 'reactor-grade' plutonium recovered from reprocessing used power reactor fuel has about one third non-fissile isotopes (mainly Pu-240)d.
In the UK, the Magnox reactors were designed for the dual use of generating commercial electricity as well as being able to produce plutonium for the country's defence programme. A report released by the UK's Ministry of Defence (MoD) says that both the Calder Hall and the Chapelcross power stations, which started up in 1956 and 1958 respectively, were operated on this basis3. The government confirmed in April 1995 that production of plutonium for defence purposes had ceased at these two stations, which& are both now permanently shutdown. The other UK Magnox reactors were civil stations subject to full international safeguards.
International safeguards arrangements applied to traded uranium extend to the plutonium arising from it, ensuring constant audits even of reactor-grade material. This addresses uncertainty as to the explosive potential of reactor-grade plutonium and the weapons proliferation potential of it. (All we know for sure is that it has never been made to explode.)
The International Atomic Energy Agency (IAEA) is conservative on this matter so that, for the purpose of applying IAEA safeguards measures, all plutonium (other than plutonium comprising 80% or more of the isotope Pu-238) is defined by the IAEA as a 'direct-use' material, that is, "nuclear material that can be used for the manufacture of nuclear explosives components without transmutation or further enrichment".  The IAEA is not saying that all plutonium is suitable for making weapons, simply that on the basis of calculations and under certain technically-demanding conditions it might be made to explode. The 'direct use' definition applies also to plutonium which has been incorporated into commercial MOX fuel, which as such certainly could not be made to explode.

Type Composition Origin Use
Reactor-grade from high-burnup fuel 55-70% Pu-239; more than 19% Pu-240 (typically about 30% non-fissile Pu) Comprises about 1% of spent fuel from normal operation of civil nuclear reactors used for electricity generation As ingredient (c. 5%) of MOX fuel for normal reactor
Weapons-grade Pu-239 with less than 8% Pu-240 From military 'production' reactors specifically designed and operated for production of low burn-up Pu Nuclear weapons (can be recycled as fuel in fast neutron reactor or as ingredient of MOX)

Resources of plutonium

Total world generation of reactor-grade plutonium in spent fuel is some 70 tonnes per year. About 1300 tonnes have been produced so far, and most of this remains in the used fuel, with some 370 tonnes extracted. About one third of the separated Pu (130 t) has been used in MOX fuel over the last 30 years. Currently 8-10 tonnes of Pu is used in MOX each year (see page on Mixed Oxide (MOX) Fuel).
Three US reactors are able to run fully on MOX, as can Canadian heavy water (CANDU) reactors. All Western and the later Russian light water reactors can use 30% MOX in their fuel. Some 32 European reactors are licensed to use MOX fuel, and several in France are using it as 30% of their fuel. Areva's EPR design is capable of running a full core load of MOX.
About 22 tonnes of reactor-grade plutonium is separated by reprocessing plants in the OECD each year and this is set to increase. Eventually its usage in MOX is expected to outstrip this level of production so that stockpiles diminish.
At the end of 2007 the UK had 77 tonnes of separated civil plutonium4 and this stockpile from historic and current operations is expected to grow to around 101 tonnes - including some 83t from Magnox fuel and 15t from AGR fuel. In addition, the UK will hold around 34 tonnes on behalf of foreign utilities once reprocessing contracts have been fulfilled5. Using all of UK's plutonium in MOX fuel rather than immobilising it as waste is expected to yield a £700-1200 million resource cost saving to UK, along with 300 billion kWh of electricity (about one year's UK supply). The civil plutonium stockpile could be consumed in two 1000 MWe light water reactors using 100% MOX fuel over 35 years.
At the end of 2006 France held nearly 50 tonnes of separated plutonium and Russia 41 tonnes. Worldwide stocks were estimated as just over 250 tonnes.
Disarmament will give rise to some 150-200 tonnes of weapons-grade plutonium, over half of it in former USSR. Discussions are progressing as to what should be done with it. The main options for the disposal of weapons-grade plutonium are:
  • Vitrification with high-level waste - treating plutonium as waste.
  • Fabrication with uranium oxide as MOX fuel for burning in existing reactors.
  • Fuelling fast-neutron reactors.
In June 2000, the USA and Russia agreed to dispose of 34 tonnes each of weapons-grade plutonium by 2014, and since then the US government has released further surplus weapons plutonium. The US government planned to pursue the first two options above, though it has since dropped the first one for any significant amount of material. Construction on the Mixed Oxide Fuel Fabrication Facility at the Savannah River Site near Aiken, South Carolina commenced in August 2007. The plant is designed to convert 3.5 t/yr of weapons-grade plutonium into mixed oxide (MOX) fuel. Initial trials of MOX fuel made with weapons plutonium have been successful. Russia plans to use all its military plutonium in fast-neutron reactors, and the USA will contribute $400 million towards effecting this. The 2000 agreement was reaffirmed in 2010.
Developments under the Global Nuclear Energy Partnership (GNEP)/Advanced Fuel Cycle Initiative (AFCI) make it very likely that the some military plutonium will be used in fast reactors in the USA.
Generation IV reactor designs are under development through an international project. Four of the six designs are fast neutron reactors and will thus utilize plutonium in some way. In these, plutonium production will take place in the core, where burn-up is high and the proportion of plutonium isotopes other than Pu-239 will remain high.

Toxicity and health effects

Despite being toxic both chemically and because of its ionising radiation, plutonium is far from being "the most toxic substance on Earth" or so hazardous that "a speck can kill". On both counts there are substances in daily use that, per unit of mass, have equal or greater chemical toxicity (arsenic, cyanide, caffeine) and radiotoxicity (smoke detectors).
There are three principal routes by which plutonium can get into human beings who might be exposed to it:
  • Ingestion.
  • Contamination of open wounds.
  • Inhalation.
Ingestion is not a significant hazard, because plutonium passing through the gastro-intestinal tract is poorly absorbed and is expelled from the body before it can do harm.
Contamination of wounds has rarely occurred although thousands of people have worked with plutonium. Their health has been protected by the use of remote handling, protective clothing and extensive health monitoring procedures.
The main threat to humans comes from inhalation. While it is very difficult to create airborne dispersion of a heavy metal like plutonium, certain forms, including the insoluble plutonium oxide, at a particle size less than 10 microns (0.01 mm), are a hazard. If inhaled, much of the material is immediately exhaled or is expelled by mucous flow from the bronchial system into the gastro-intestinal tract, as with any particulate matter. Some however will be trapped and readily transferred, first to the blood or lymph system and later to other parts of the body, notably the liver and bones. It is here that the deposited plutonium's alpha radiation may eventually cause cancer.
However, the hazard from Pu-239 is similar to that from any other alpha-emitting radionuclides which might be inhaled. It is less hazardous than those which are short-lived and hence more radioactive, such as radon daughters, the decay products of radon gas, which (albeit in low concentrations) are naturally common and widespread in the environment.
In the 1940s some 26 workers at US nuclear weapons facilities became contaminated with plutonium. Intensive health checks of these people have revealed no serious consequence and no fatalities that could be attributed to the exposure. In the 1990s plutonium was injected into and inhaled by some volunteers, without adverse effects. In the 1950s Queen Elizabeth II was visiting Harwell and was handed a lump of plutonium (presumably Pu-239) in a plastic bag and invited to feel how warm it was.
Plutonium is one among many toxic materials that have to be handled with great care to minimise the associated but well understood risks.

Further Information
 Notes
a. Half-life is the time it takes for a radionuclide to lose half of its own radioactivity. The fissile isotopes can be used as fuel in a nuclear reactor, others are capable of absorbing neutrons and becoming fissile (i.e. they are 'fertile'). Alpha decays are generally accompanied by gamma radiation. [Back]
b. Comparable isotopic ratios are found in the spent fuel of CANDU heavy water reactors at much lower burnups (8 GWd/t), due to their use of natural uranium fuel and high thermal neutron spectrum. From gas graphite Magnox reactors the plutonium has more Pu-239 - about 65%, plus 25% Pu-240, 5% Pu-241, 1% Pu-242 and negligible Pu-238. [Back]
c. The term 'fissionable' applies to isotopes that can be made to undergo fission. If a fissionable isotope only requires neutrons with low kinetic energy to undergo fission, then it is said to 'fissile'. Thus, all fissile isotopes are fissionable. Pu-240 is fissionable, as it undergoes fission in a fast neutron reactor - but it is not a fissile isotope. [Back]
d. In 1962 a nuclear device using low-burnup plutonium from a UK Magnox reactor was detonated in the USA. The isotopic composition of this plutonium has not been officially disclosed, but it was evidently about 85% Pu-239 - what would since 1971 have been called 'fuel-grade' plutonium. The plutonium used in the bomb test was almost certainly derived from the Calder Hall/Chapelcross reactors operating as military plutonium production reactors (see Reference 3 below). As part of the UK's 1998 Strategic Defence Review, a UK Ministry of Defence document (The United Kingdom's Defence Nuclear Weapons Programme) states: "The US Government has given assurances that UK plutonium transferred to the US since 1964 was not used in the US nuclear weapons programme. It is theoretically possible, but very unlikely, that some UK civil plutonium may have been transferred to the US and used in the US nuclear weapons programme before 1964." [Back]
 References
1. Data taken from NDA Plutonium Options, Nuclear Decommissioning Authority (2008). [Back]
2. Information on the Oklo natural reactors is on the Swedish Nuclear Fuel and Waste Management Company (Svensk Kärnbränslehantering, SKB) website (www.skb.se). See also I. Gurban and M. Laaksoharju, Uranium transport around the reactor zone at Okelobondo (Oklo), Data evaluation with M3 and HYTEC, SKB Technical Report TR-99-36 (December 1999). [Back]
3. Plutonium and Aldermaston - An Historical Account, UK Ministry of Defence (2000). [Back]
4. Communication Received from the United Kingdom of Great Britain and Northern Ireland Concerning Its Policies Regarding the Management of Plutonium - Statement on the Management of Plutonium and of High Enriched Uranium, IAEA Information Circular INFCIRC/549/Add.8/11 (02 July 2008). [Back]
5. NDA Plutonium Options, Nuclear Decommissioning Authority (2008). [Back]
General sources
HV Henderickz, Plutonium: blessing or curse?, Copper Beech 1998 (ISBN: 9782930221083)
Management of Separated Plutonium -The Technical Options, NEA/OECD Paris 1997 (ISBN: 9789264154100)
Management of separated plutonium,  The Royal Society, February 1998 (ISBN: 9780854035144)
Plutonium Fuel: An Assessment, OECD/NEA Paris 1989 (ISBN: 9789264132658)
Plutonium Management in the Medium Term - A Review by the OECD/NEA Working Party on the Physics of Plutonium Fuels and Innovative Fuel Cycles (WPPR), OECD/NEA Paris 2003 (ISBN: 9264021515)
Managing the Plutonium Surplus: Applications and Technical Options, NATO 1994 (ISBN 9780792331247)
The Toxicity of Plutonium, Medical Research Council, HMSO London 1975 (ISBN: 0114500304)
Plutonium articles in Revue Generale Nucleaire, June 1995
D Fishlock, Drama of Plutonium, Nuclear Engineering International (2005)
J Carlson, Introduction to the Concept of Proliferation Resistance, International Commission on Nuclear Non-proliferation and Disarmament (ICNND, January 2009). See also the website of the ICNND (www.icnnd.org) for more information on nuclear non-proliferation.
Related information pages

Mixed Oxide (MOX) Fuel

(Updated May 2013)
  • Mixed oxide (MOX) fuel provides about 2% of the new nuclear fuel used today.
  • MOX fuel is manufactured from plutonium recovered from used reactor fuel.
  • MOX fuel also provides a means of burning weapons-grade plutonium (from military sources) to produce electricity.
In every nuclear reactor there is both fission of isotopes such as uranium-235, and the formation of new, heavier isotopes due to neutron capture, primarily by U-238. Most of the fuel mass in a reactor is U-238. This can become plutonium-239 and by successive neutron capture Pu-240, Pu-241 and Pu-242 as well as other transuranic isotopes (see page on Plutonium). Pu-239 and Pu-241 are fissile, like U-235. (Very small quantities of Pu-236 and Pu-238 are formed similarly from U-235.)
Normally, with the fuel being changed every three years or so, about half of the Pu-239 is 'burned' in the reactor, providing about one third of the total energy. It behaves like U-235 and its fission releases a similar amount of energy. The higher the burn-up, the less fissile plutonium remains in the used fuel. Typically about one percent of the used fuel discharged from a reactor is plutonium, and some two thirds of this is fissile (c. 50% Pu-239, 15% Pu-241). Worldwide, some 70 tonnes of plutonium contained in used fuel is removed when refuelling reactors each year.
The plutonium (and uranium) in used fuel can be recovered through reprocessing. The plutonium could then be used in the manufacture mixed oxide (MOX) nuclear fuel, to provide energy through electricity generation. A single recycle of plutonium in the form of MOX fuel increases the energy derived from the original uranium by some 12%, and if the uranium is also recycled this becomes about 22% (based on light water reactor fuel with burn-up of 45 GWd/tU).

Reaction in standard UO2 fuel

Today there is a significant amount of separated uranium and plutonium which may be recycled, including from ex-military sources. It is equivalent to about three years' supply of natural uranium from world mines.

Inventory of separated recyclable materials 1  
  Quantity (tonnes) Natural U equivalent (tonnes)
Plutonium from reprocessed fuel 320 60,000
Uranium from reprocessed fuel 45,000 50,000
Ex-military plutonium 70 15,000
Ex-military high-enriched uranium 230 70,000

In addition, there is about 1.6 million tonnes of enrichment tails, with recoverable fissile uranium.

MOX use

MOX fuel was first used in a thermal reactor in 1963, but did not come into commercial use until the 1980s. So far about 2000 tonnes of MOX fuel has been fabricated and loaded into power reactors. In 2006 about 180 tonnes of MOX fuel was loaded into over 30 reactors (mostly PWR) in Europe.
Today MOX is widely used in Europe and in Japan. Currently about 40 reactors in Europe (Belgium, Switzerland, Germany and France) are licensed to use MOX, and over 30 are doing so.  In Japan about ten reactors are licensed to use it and several do so. These reactors generally use MOX fuel as about one third of their core, but some will accept up to 50% MOX assemblies. France aims to have all its 900 MWe series of reactors running with at least one third MOX. Japan also plans to use MOX in one third of its reactors in the near future and expects to start up a 1383 MWe (gross) reactor with a complete fuel loading of MOX at the Ohma plant in late 2014.2 Other advanced light water reactors such as the EPR or AP1000 are able to accept complete fuel loadings of MOX if required.
In the USA there was significant development work in 1960s and 19790s, and MOX fuel was used in several demonstration projects (San Onofre, Ginna PWRs, Dresden, Quad Cities and Big Rock Point). It performed acceptably and similar to uranium oxide fuel. In 2005 four MOX test assemblies made by Melox in France were tested successfully at the Catawba power station.
The use of up to 50% of MOX does not change the operating characteristics of a reactor, though the plant must be designed or adapted slightly to take it. More control rods are needed. For more than 50% MOX loading, significant changes are necessary and a reactor needs to be designed accordingly, as several new designs are. Burn-up of MOX fuel is about the same as that for UOX fuel.
An advantage of MOX is that the fissile concentration of the fuel can be increased easily by adding a bit more plutonium, whereas enriching uranium to higher levels of U-235 is relatively expensive. As reactor operators seek to burn fuel harder and longer, increasing burnup from around 30,000 MW days per tonne a few years ago to over 50,000 MWd/t now, MOX use becomes more attractive.
Reprocessing to separate plutonium for recycle as MOX becomes more economic as uranium prices rise. MOX use also becomes more attractive as the need to reduce the volume of spent fuel increase. Seven UO2 fuel assemblies give rise to one MOX assembly plus some vitrified high-level waste, resulting in only about 35% of the volume, mass and cost of disposal.

Recycling normal used fuel

If used fuel is to be recycled, the first step is separating the plutonium and the remaining uranium (about 96% of the spent fuel) from the fission products with other wastes (together about 3%). The plutonium then needs to be separated from most or all of the uranium. All this is undertaken at a reprocessing plant (see information page on Processing of Used Nuclear Fuel).
The plutonium, as an oxide, is then mixed with depleted uranium left over from an enrichment plant to form fresh mixed oxide fuel (MOX, which is UO2+PuO2). MOX fuel, consisting of about 7-10% plutonium mixed with depleted uranium, is equivalent to uranium oxide fuel enriched to about 4.5% U-235, assuming that the plutonium has about two thirds fissile isotopes. If weapons plutonium is used (>90% Pu-239), only about 5% plutonium is needed in the mix. The plutonium content of commercial MOX fuel varies up to 10.8% depending on the design of the fuel, and averages about 9.5%. Fuel in an EPR with 30% MOX having less than 10.8% Pu is equivalent to 4.2% enriched uranium fuel. An EPR with 100% MOX fuel can use a wider variety of used fuel material (burnup, initial enrichment, Pu quality) than with only 30% MOX.

Reaction in MOX Fuel

Plutonium from reprocessed fuel is usually fabricated into MOX as soon as possible to avoid problems with the decay of short-lived plutonium isotopes. In particular, Pu-241 (half-life 14 years) decays to Am-241 which is a strong gamma emitter, giving rise to a potential occupational health hazard if separated plutonium over five years old is used in a normal MOX plant. The Am-241 level in stored plutonium increases about 0.5% per year, with corresponding decrease in fissile value of the plutonium. Pu-238 (half-life 88 years), a strong alpha emitter and a source of spontaneous neutrons, is increased in high-burnup fuel. Pu-239, Pu-240 and Pu-242 are long-lived and hence little changed with prolonged storage. (See also information page on Plutonium).
Fast neutron reactors allow multiple recycling of plutonium, since all transuranic isotopes there are fissionable, but in thermal reactors isotopic degradation limits the plutonium recycle potential and most spent MOX fuel is stored pending the greater deployment of fast reactors. (The plutonium isotopic composition of used MOX fuel at 45 GWd/tU burnup is about 37% Pu-239, 32% Pu-240, 16% Pu-241, 12% Pu-242 and 4% Pu-238.)
Recovered uranium from a reprocessing plant may be re-enriched on its own for use as fresh fuel. Because it contains some neutron-absorbing U-234 and U-236, reprocessed uranium must be enriched significantly (e.g. one-tenth) more than is required for natural uranium. Thus reprocessed uranium from low-burn-up fuel is more likely to be suitable for re-enrichment, while that from high burn-up fuel is best used for blending or MOX fabrication.
Reprocessing of 850 tonnes of French used fuel per year (about 15 years after discharge) produces 8.5 tonnes of plutonium (immediately recycled as 100 tonnes of MOX) and 810 tonnes of reprocessed uranium (RepU). Of this about two-thirds is converted into stable oxide form for storage. One-third of the RepU is re-enriched and EdF has demonstrated its use in 900 MWe power reactors.

MOX production

Two plants currently produce commercial quantities of MOX fuel – in France and UK. In 2006 a 40 t/yr Belgian plant closed3 and in April 2007 the French Melox plant was licensed for an increase in production from 145 to 195 t/yr. Also the Sellafield MOX Plant in UK was downrated from 128 to 40 t/yr, and in August 2011 the Nuclear Decommissioning Authority announced that it had reassessed the plant's prospects and would close it.
Japan is planning to start up a 130 t/yr J-MOX plant at Rokkasho in 2015. Meanwhile, construction on a MOX fabrication facility at the Savannah River Site in the USA is underway for 2016 start-up – see section below on MOX and disposition of weapons plutonium.

World mixed oxide fuel fabrication capacities (t/yr)
  2009 2015
France, Melox 195 195
Japan, Tokai 10 10
Japan, Rokkasho 0 130
Russia, Mayak, Ozersk 5 5
 Russia, Zheleznogorsk 0 60?
UK, Sellafield 40 0
Total for LWR 250  400

MOX is also used in fast neutron reactors in several countries, particularly France and Russia. It was first developed for this purpose, with experimental work being done in USA, Russia, UK, France, Germany, Belgium and Japan. Today, Russia leads the way in fast reactor development and has long-term plans to build a new generation of fast reactors fuelled by MOX. The world's largest fast reactor – the 800 MWe BN-800 – is currently under construction at Beloyarsk in the Urals and due to start up in 2014.
At present the output of reprocessing plants exceeds the rate of plutonium usage in MOX, resulting in inventories of (civil) plutonium in several countries. These stocks are expected to exceed 250 tonnes before they start to decline after 2010 as MOX use increases, with MOX then expected to supply about 5% of world reactor fuel requirements.
The UK is investigating the incorporation of its 120 tonnes of reactor-grade plutonium into CANMOX fuel which would be used in four Candu EC6 reactors. The fuel would have 2% plutonium and four UK units (2800 MWe) would require about 400 t/yr of it. The used fuel would be stored for a hundred years and then sent to a repository.

MOX and disposition of weapons plutonium

Under the Plutonium Management and Disposition Agreement, Russia and the USA agreed in 2000 to each dispose of (or immobilise) 34 tonnes of weapons-grade plutonium deemed surplus to requirements (see page on Military Warheads as a Source of Nuclear Fuel). 
The Mixed Oxide Fuel Fabrication Facility (MFFF) at the Savannah River Site in South Carolina began construction in August 2007 and will convert the US plutonium to MOX fuel. Expected to begin operations in 2016, the MFFF is designed to turn 3.5 t/yr of weapons-grade plutonium into about 150 MOX fuel assemblies, both PWR and BWR. The contract to design, build and operate the MFFF was awarded to the Shaw AREVA MOX Services consortium in 1999, with the $2.7 billion construction option being exercised in May 2008.4 Four MOX fuel lead test assemblies manufactured from US weapons plutonium and fabricated at the Melox plant in France were successfully burned on a trial basis at the Catawba plant.
Meanwhile, following several years of dispute, in November 2007 the USA and Russia agreed that Russia would dispose of its 34 t of weapons-grade plutonium by conversion to MOX fuel, which would be burned in the BN-600 reactor at the Beloyarsk nuclear plant, and in the BN-800 under construction at the same site.5 Under this plan, Russia would begin disposition in the BN-600 reactor in the 2012 timeframe. Disposition in the BN-800 would follow soon thereafter. Once disposition begins, the two reactors could dispose of approximately 1.5 t of Russian weapons plutonium per year. The USA agreed to contribute $400 million to the project. A 60 t/yr commercial MOX Fuel Fabrication Facility (MFFF) is scheduled to start up at Zheleznogorsk by 2014, operated by the Mining & Chemical Combine (MCC). It will make MOX granules and pelletised MOX for 400 fuel assemblies per year for the BN-800 and future fast reactors. The capacity is designed to supply five BN-800 units. This is likely to use ex-weapons plutonium. Another MOX plant for military plutonium was planned for Seversk, Siberia, but this appears to have been displaced by the MCC one.

MOX reprocessing and further use

Used MOX fuel reprocessing has been demonstrated since 1992 in France, at the La Hague plant. In 2004 the first reprocessing of used MOX fuel was undertaken on a larger scale with continuous process. Ten tonnes of MOX irradiated to about 35,000 MWd/t and with Pu content of about 4% was involved. The main problem of fully dissolving PuO2 was overcome. Since 2004 an increasing amount of MOX from German and Swiss reactors has been reprocessed, totaling about 70 tonnes, with a wide range of composition. As MOX is repeatedly recycled it is mixed with substantial proportions (70-80%) of plutonium from UOX fuel.
At present the French policy is not to reprocess used MOX fuel, but to store it and await the advent of fuel cycle developments related to Generation IV fast neutron reactor designs.

Plutonium-thorium fuel

Since the early 1990s Russia has had a programme to develop a thorium-uranium fuel, which more recently has moved to have a particular emphasis on utilisation of weapons-grade plutonium in a thorium-plutonium fuel. The programme is described in the information page on Thorium. With an estimated 150 tonnes of surplus weapons plutonium in Russia, the thorium-plutonium project would not necessarily cut across existing plans to make MOX fuel.

Further information

References

1. OECD/NEA 2007, Management of Recyclable Fissile and Fertile Materials, NEA #6107 (ISBN: 9789264032552). [Back]
2. J-Power reschedules Ohma start-up, World Nuclear News, 11 November 2008. [Back]
3. Belgonucleaire's decision to close its MOX plant was explained in its 2005 Annual Report – see http://www.belgonucleaire.be/files/JAARVERSLAG2005EN.pdf [Back]
4. Final contract for US MOX, World Nuclear News, 27 May 2008. [Back]
5. Russia and USA confirm plutonium plan, World Nuclear News, 20 November 2007. [Back]

General sources

Australian Safeguards and Non-Proliferation Office, Annual Report 1999
NATO 1994, Managing the Plutonium Surplus: Applications and Technical Options (ISBN 9780792331247)
OECD NEA 1997, Management of Separated Plutonium, the technical options (ISBN 9264154108)
Nuclear Europe Worldscan, European Nuclear Society, March/April 1997 (several articles)
Nuclear Engineering International, Europeans & MOX , July 1997
D Albright and K Kramer, Tracking Plutonium Inventories, Plutonium Watch, July (revised August) 2005 – see http://www.isis-online.org/global_stocks/end2003/plutonium_watch2005.pdf
International Atomic Energy Agency, Status and Advances in MOX Fuel Technology, Technical Review Series # 415 (2003)
www.moxproject.com, the website for the Mixed Oxide Fuel Fabrication Facility (MFFF) at the Savannah River Site
Marc Arslan, 2012, Fuel Cycle Strategies to Optimise the use of MOX Fuels, WNFC Helsinki.

Related information pages

The Nuclear Fuel Cycle
Plutonium
Processing of Used Nuclear Fuel
Military Warheads as a Source of Nuclear Fuel
Japanese Waste and MOX Shipments From Europe

Accelerator-driven Nuclear Energy

(Updated October 2011)
  • Powerful accelerators can produce neutrons by spallationa
  • This process may be linked to conventional nuclear reactor technology in an accelerator-driven system (ADS) to transmute long-lived radioisotopes in used nuclear fuel into shorter-lived fission products.
  • There is also increasing interest in the application of ADSs to running subcritical nuclear reactors powered by thorium.
Used fuel from a conventional nuclear power reactor contains a number of radionuclides, most of which (notably fission products) decay rapidly, so that their collective radioactivity is reduced to less than 0.1% of the original level 50 years after being removed from the reactor. However, a significant proportion of the wastes contained in used nuclear fuel is long-lived actinides (particularly neptunium, americium and curium). In recent years, interest has grown in the possibility of separating (or partitioning) the long-lived radioactive waste from the used fuel and transmuting it into shorter-lived radionuclides so that the management and eventual disposal of this waste is easier and less expensive.
The transmutation of long-lived radioactive waste can be carried out in an accelerator-driven system (ADS), where neutrons produced by an accelerator are directed at a blanket assembly containing the waste along with fissionable fuel. Following neutron capture, the heavy isotopes in the blanket assembly subsequently fission, producing energy in doing so. ADSs could also be used to to generate power from the abundant element thorium.

Accelerator-driven systems

High-current, high-energy accelerators or cyclotrons are able produce neutrons from heavy elements by spallation. A number of research facilities exist which explore this phenomenon, and there are plans for much larger ones. In this process, a beam of high-energy protons (usually >500 MeV) is directed at a high-atomic number target (e.g. tungsten, tantalum, depleted uranium, thorium, zirconium, lead, lead-bismuth, mercury) and up to one neutron can be produced per 25 MeV of the incident proton beam. (These numbers compare with 200-210 MeV released by the fission of one uranium-235 or plutonium-239 atomb.) A 1000 MeV beam will create 20-30 spallation neutrons per proton.
The spallation neutrons have only a very small probability of causing additional fission events in the target. However, the target still needs to be cooled due to heating caused by the accelerator beam.
If the spallation target is surrounded by a blanket assembly of nuclear fuel, such as fissile isotopes of uranium or plutonium (or thorium-232 which can breed to U-233), there is a possibility of sustaining a fission reaction. This is described as an accelerator-driven system (ADS)c. In such a system, the neutrons produced by spallation would cause fission in the fuel, assisted by further neutrons arising from that fission. Up to 10% of the neutrons could come from the spallation, though it would normally be less, with the rest of the neutrons arising from fission events in the blanket assembly. An ADS can only run when neutrons are supplied to it because it burns material which does not have a high enough fission-to-capture ratio for neutrons to maintain a fission chain reaction. One then has a nuclear reactor which could be turned off simply by stopping the proton beam, rather than needing to insert control rods to absorb neutrons and make the fuel assembly subcritical. Because they stop when the input current is switched off, accelerator-driven systems are seen as safer than normal fission reactors.

Thorium utilisation

For many years there has been interest in utilising thorium-232 as a nuclear fuel since it is three to five times as abundant in the Earth's crust as uranium. A thorium reactor would work by having Th-232 capture a neutron to become Th-233 which decays to uranium-233, which fissions. (The process of converting fertile isotopes such as Th-232 to fissile ones is known as 'breeding'.) The problem is that insufficient neutrons are generated to keep the reaction going, and so driver fuel is needed – either plutonium or enriched uranium. Just as with uranium, if all of it and not a mere 0.7% of uranium is to be used as fuel, fast neutron reactors are required in the system. (A fast neutron spectrum enables maximum fission with minimum build-up of new actinides due to neutron capture.)
An alternative is provided by the use of accelerator-driven systems. The concept of using an ADS based on the thorium-U-233 fuel cycle was first proposed by Professor Carlo Rubbia, but at a national level, India is the country with most to gain, due to its very large thorium resources. India is actively researching ADSs as an alternative to its main fission program focused on thorium.
The core of an ADS is mainly thorium, located near the bottom of a 25 metre high tank. It is filled with some 8000 tonnes of molten lead or lead-bismuth at high temperature – the primary coolant, which circulates by convection around the core. Outside the main tank is an air gap to remove heat if needed. The accelerator supplies a beam of high-energy protons down a beam pipe to the spallation target inside the core, and the neutrons produced enter the fuel and transmute the thorium into protactinium, which soon decays to U-233 which is fissile. The neutrons also cause fission in uranium, plutonium and possibly transuranics present, releasing energy. A 10 MW proton beam might thus produce 1500 MW of heat (and thus 600 MWe of electricity, some 30 MWe of which drives the accelerator). With a different, more subcritical, core a 25 MW proton beam would be required for the same result. Today's accelerators are capable of only 1 MW beams.
There have been several proposals to develop a prototype reactor of this kind, sometimes popularly called an energy amplifier.
A UK-Swiss proposal for an accelerator-driven thorium reactor (ADTR) has gone to feasibility study stage, for a 600 MWe lead-cooled fast reactor. This envisages a ten-year self-sustained thorium fuel cycle, using plutonium as a fission starter. Molten lead is both the spallation target and the coolant. In contrast to other designs with neutron multiplication coefficients of 0.95 - 0.98 and requiring more powerful accelerators, this ADTR has a coefficient of 0.995 and requires only a 3-4 MW accelerator, with fast-acting shutdown rods, control rods, and precise measurement of neutron flux.
A 2008 Norwegian study summarised the advantages and disadvantages of an ADS fuelled by thorium, relative to a conventional nuclear power reactor, as follows, and said that such a system was not likely to operate in the next 30 years:1
Advantages Disadvantages
Much smaller production of long-lived actinides More complex (with accelerator)
Minimal probability of runaway reaction Less reliable power production due to accelerator downtime
Efficient burning of minor actinides Large production of volatile radioactive isotopes in the spallation target
Low pressure system The beam tube may break containment barriers

Waste incinerator

An ADS can be used to destroy heavy isotopes contained in the used fuel from a conventional nuclear reactor – particularly actinidesd. Here the blanket assembly is actinide fuel and/or used nuclear fuel. One approach is to start with fresh used fuel from conventional reactors in the outer blanket region and progressively move it inwards. It is then removed and reprocessed, with the uranium recycled and most fission products separated as waste. The actinides are then placed back in the system for further 'incineration'e.
ADSs could also be used to destroy longer-lived fission products contained in used nuclear fuel, such as Tc-99 and I-129 (213,000 and 16 million years half-lives, respectively). These isotopes can acquire a neutron to become Tc-100 and I-130 respectively, which are very short-lived, and beta decay to Ru-100 and Xe-130, which are stable.
Commercial application of partitioning and transmutation (P&T), which is attractive particularly for actinides, is still a long way off, since reliable separation is needed to ensure that stable isotopes are not transmuted into radioactive ones. New reprocessing methods would be required, including electrometallurgical ones (pyroprocessing). The cost and technology of the partitioning together with the need to develop the necessary high-intensity accelerators seems to rule out early use. An NEA study showed that multiple recycling of the fuel would be necessary to achieve major (e.g. 100-fold) reductions in radiotoxicity, and also that the full potential of a transmutation system can be exploited only with commitment to it for 100 years or more2.
The French Atomic Energy Commission is funding research on the application of this process to nuclear wastes from conventional reactors, as is the US Department of Energy. The Japanese Omega (Options Making Extra Gain from Actinides) project envisages an accelerator transmutation plant for nuclear wastes operated in conjunction with ten or so large conventional reactors. The French concept similarly links a transmutation - energy amplifying system with about eight large reactors. Other research has been proceeding in USA, Russia and Europe.
Another area of current interest in the use of ADSs is in their potential to dispose of weapons-grade plutonium, as an alternative to burning it as mixed oxide fuel in conventional reactors. Two alternative strategies are envisaged: the plutonium and minor actinides being managed separately, with the latter burned in ADSs while plutonium is burned in fast reactors; and the plutonium and minor actinides being burned together in ADSs, providing better proliferation resistance but posing some technical challenges. Both can achieve major reduction in waste radiotoxicity, and the first would add only 10-20% to electricity costs (compared with the once-through fuel cycle).

ADS research and development

What was claimed to be the world’s first ADS experiment was begun in March 2009 at the Kyoto University Research Reactor Institute (KURRI), utilizing the Kyoto University Critical Assembly (KUCA). The research project was commissioned by Japan’s Ministry of Education, Culture, Sports, Science and Technology (MEXT) six years earlier. The experiment irradiates a high-energy proton beam (100 MeV) from the accelerator on to a heavy metal target set within the critical assembly, after which the neutrons produced by spallation are bombarded into a subcritical fuel core.
The Indian Atomic Energy Commission is proceeding with design studies for a 200 MWe PHWR accelerator-driven system (ADS) fuelled by natural uranium and thoriumf. Uranium fuel bundles would be changed after about 7 GWd/t burn-up, but thorium bundles would stay longer, with the U-233 formed adding reactivity. This would be compensated for by progressively replacing some uranium with thorium, so that ultimately there is a fully-thorium core with in situ breeding and burning of thorium. This is expected to mean that the reactor needs only 140 tU through its life and achieves a high burnup of thorium - about 100 GWd/t. A 30 MW accelerator would be required to run it.
The Belgian Nuclear Research Centre (SCK.CEN) is planning to begin construction on the MYRRHA (Multipurpose Hybrid Research Reactor for High-tech Applications) research reactor at Mol in 2015. Initially it will be a 57 MWt ADS, consisting of a proton accelerator delivering a 600 MeV, 2.5 mA (or 350 MeV, 5 mA) proton beam to a liquid lead-bismuth (Pb-Bi) spallation target that in turn couples to a Pb-Bi cooled, subcritical fast nuclear core (see Research and development section in the information page on Nuclear Power in Belgium).

Further Information

Notes

a. Spallation is the process where nucleons are ejected from a heavy nucleus being hit by a high energy particle. In this case, a high-enery proton beam directed at a heavy target expels a number of spallation particles, including neutrons. [Back]
b. An average fission event of U-235 releases 200 MeV of energy and is accompanied by the release of an average of 2.43 neutrons. [Back]
c. Accelerator-driven systems are also referred to as energy amplifiers since more energy is released from the fission reactions in the blanket assembly than is needed to power the particle accelerator. Professor Carlo Rubbia, a former director of the international CERN laboratory, is credited with proposing the concept of the energy amplifier, using natural thorium fuel. [Back]
d. In the case of atoms of odd-numbered isotopes heavier than thorium-232, they have a high probability of absorbing a neutron and subsequently undergoing nuclear fission, thereby producing some energy and contributing to the multiplication process. Even-numbered isotopes can capture a neutron, perhaps undergo beta decay, and then fission. Therefore in principle, the subcritical nuclear reactor may be able to convert all transuranic elements into (generally) short-lived fission products and yield some energy in the process. [Back]
e. As well as fission products, the process generates spallation products from the target material, in direct proportion to the energy of the proton beam. Some of these are volatile and will find their way into the cover gas system above the coolant, posing a major maintenance challenge. Their radiotoxicity is likely to exceed that of the fission products in the short term, which is relevant to operation and storage rather than final disposal. Ultimately the burning of actinides means that overall radiotoxicity of them is reduced greatly by the time 1000 years has elapsed, and is then less than that of the equivalent uranium ore. [Back]
f. India is already running a very small research reactor on U-233 fuel extracted from thorium which has been irradiated and bred in another reactor. When this started in 1996 it was hailed as a first step towards the thorium cycle there, utilizing 'near breeder' reactors. [Back]
References
1. Thorium as an Energy Source – Opportunities for Norway, Thorium Report Committee, Norwegian Ministry of Petroleum and Energy (2008). See also Thorium committee submits report: Neither dismisses nor embraces thorium fuel, The Research Council of Norway (21 February 2008) and Norway's thorium option 'should be kept open', World Nuclear News (18 February 2008) [Back]
2. Accelerator-driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles – A Comparative Study, OECD Nuclear Energy Agency (2002), available on the NEA webpage on Accelerator-driven Systems (ADS) and Fast Reactors (FR) in Advanced Nuclear Fuel Cycles (http://www.nea.fr/html/ndd/reports/2002/nea3109.html) [Back]
3. Two articles on the ADTR power station, Nuclear Future 7,3, May-June 2011.
General sources
Accelerator driven nuclear energy systems, J.W. Boldeman, Australian Academy of Technological Sciences and Engineering Symposium, Energy for Ever – Technological Challenges of Sustainable Growth (November 1997)
P&T: A long-term option for radioactive waste disposal?, E. Bertel, L. Van den Durpel, NEA News No. 20.2, p20 (2002)
The answer is No – Does transmutation of spent nuclear fuel produce more hazardous material then it destroys?, H. Treulle, Radwaste Solutions (July-August 2002)


Processing of Used Nuclear Fuel

(Updated June 2013)
  • Used nuclear fuel has long been reprocessed to extract fissile materials for recycling and to reduce the volume of high-level wastes.
  • New reprocessing technologies are being developed to be deployed in conjunction with fast neutron reactors which will burn all long-lived actinides.
  • A significant amount of plutonium recovered from used fuel is currently recycled into MOX fuel; a small amount of recovered uranium is recycled.
A key, nearly unique, characteristic of nuclear energy is that used fuel may be reprocessed to recover fissile and fertile materials in order to provide fresh fuel for existing and future nuclear power plants. Several European countries, Russia and Japan have had a policy to reprocess used nuclear fuel, although government policies in many other countries have not yet addressed the various aspects of reprocessing.
Over the last 50 years the principal reason for reprocessing used fuel has been to recover unused uranium and plutonium in the used fuel elements and thereby close the fuel cycle, gaining some 25% to 30% more energy from the original uranium in the process and thus contributing to energy security. A secondary reason is to reduce the volume of material to be disposed of as high-level waste to about one fifth. In addition, the level of radioactivity in the waste from reprocessing is much smaller and after about 100 years falls much more rapidly than in used fuel itself.
In the last decade interest has grown in recovering all long-lived actinides together (i.e. with plutonium) so as to recycle them in fast reactors so that they end up as short-lived fission products. This policy is driven by two factors: reducing the long-term radioactivity in high-level wastes, and reducing the possibility of plutonium being diverted from civil use – thereby increasing proliferation resistance of the fuel cycle. If used fuel is not reprocessed, then in a century or two the built-in radiological protection will have diminished, allowing the plutonium to be recovered for illicit use (though it is unsuitable for weapons due to the non-fissile isotopes present).
Reprocessing used fuela to recover uranium (as reprocessed uranium, or RepU) and plutonium (Pu) avoids the wastage of a valuable resource. Most of it – about 96% – is uranium, of which less than 1% is the fissile U-235 (often 0.4-0.8%); and up to 1% is plutonium. Both can be recycled as fresh fuel, saving up to 30% of the natural uranium otherwise required. The materials potentially available for recycling (but locked up in stored used fuel) could conceivably run the US reactor fleet of about 100 GWe for almost 30 years with no new uranium input.
So far, almost 90,000 tonnes (of 290,000 t discharged) of used fuel from commercial power reactors has been reprocessed. Annual reprocessing capacity is now some 4000 tonnes per year for normal oxide fuels, but not all of it is operational.
Between now and 2030 some 400,000 tonnes of used fuel is expected to be generated worldwide, including 60,000 t in North America and 69,000 t in Europe.

World commercial reprocessing capacity1,2
(tonnes per year)
LWR fuel France, La Hague 1700
 

 
UK, Sellafield (THORP)
900
 
Russia, Ozersk (Mayak)
400
 
Japan (Rokkasho)
800*
 
Total LWR (approx)
 
3800
Other nuclear fuels UK, Sellafield (Magnox) 1500
 
India (PHWR, 4 plants)
330
 
Total other (approx)
 
1830
Total civil capacity    
5630
* now expected to start operation in October 2013

Products of reprocessing

Used fuel contains a wide array of nuclides in varying valency states. Processing it thus inherently complex chemically, and made more difficult because many of those nuclides are also radioactive.
The composition of reprocessed uranium (RepU) depends on the initial enrichment and the time the fuel has been in the reactor, but it is mostly U-238. It will normally have less than 1% U-235 (typically about 0.5% U-235) and also smaller amounts of U-232 and U-236 created in the reactor. The U-232, though only in trace amounts, has daughter nuclides which are strong gamma-emitters, making the material difficult to handle. However, once in the reactor, U-232 is no problem (it captures a neutron and becomes fissile U-233). It is largely formed through alpha decay of Pu-236, and the concentration of it peaks after about 10 years of storage.
The U-236 isotope is a neutron absorber present in much larger amounts, typically 0.4% to 0.6% – more with higher burn-up – which means that if reprocessed uranium is used for fresh fuel in a conventional reactor it must be enriched significantly more (e.g. up to one-tenth more) than is required for natural uraniumb. Thus RepU from low burn-up fuel is more likely to be suitable for re-enrichment, while that from high burn-up fuel is best used for blending or MOX fuel fabrication.
The other minor uranium isotopes are U-233 (fissile), U-234 (from original ore, enriched with U-235, fertile), and U-237 (short half-life beta emitter). None of these affects the use of handling of the reprocessed uranium significantly. In the future, laser enrichment techniques may be able to remove these isotopes.
Reprocessed uranium (especially from earlier military reprocessing) may also be contaminated with traces of fission products and transuranics. This will affect its suitability for recycling either as blend material or via enrichment. Over 2002-06 USEC successfully cleaned up 7400 tonnes of technetium-contaminated uranium from the US Department of Energy.
Most of the separated uranium (RepU) remains in storage, though its conversion and re-enrichment (in UK, Russia and Netherlands) has been demonstrated, along with its re-use in fresh fuel. Some 16,000 tonnes of RepU from Magnox reactors in UK has been usedc to make about 1650 tonnes of enriched AGR fuel. In Belgium, France, Germany and Switzerland over 8000 tonnes of RepU has been recycled into nuclear power plants. In Japan the figure is over 335 tonnes in tests and in India about 250 t of RepU has been recycled into PHWRs. Allowing for impurities affecting both its treatment and use, RepU value has been assessed as about half that of natural uranium.
Plutonium from reprocessing will have an isotopic concentration determined by the fuel burn-up level. The higher the burn-up levels, the less value is the plutonium, due to increasing proportion of non-fissile isotopes and minor actinides, and depletion of fissile plutonium isotopesd. Whether this plutonium is separated on its own or with other actinides is a major policy issue relevant to reprocessing (see section on Reprocessing policies below).
Most of the separated plutonium is used almost immediately in mixed oxide (MOX) fuel. World MOX production capacity is currently around 200 tonnes per year, nearly all of which is in France (see page on Mixed Oxide (MOX) Fuel).
Inventory of separated recyclable materials worldwide3
  Quantity (tonnes) Natural U equivalent (tonnes)
Plutonium from reprocessed fuel 320 60,000
Uranium from reprocessed fuel 45,000 50,000
Ex-military plutonium 70 15,000
Ex-military high-enriched uranium 230 70,000

History of reprocessing

A great deal of hydrometallurgical reprocessing has been going on since the 1940s, originally for military purposes, to recover plutonium for weapons (from low burn-up used fuel, which has been in a reactor for only a very few months). In the UK, metal fuel elements from the Magnox generation gas-cooled commercial reactors have been reprocessed at Sellafield for about 50 yearse. The 1500 t/yr Magnox reprocessing plant undertaking this has been successfully developed to keep abreast of evolving safety, hygiene and other regulatory standards. From 1969 to 1973 oxide fuels were also reprocessed, using part of the plant modified for the purpose, and the 900 t/yr Thermal Oxide Reprocessing Plant (THORP) at Sellafield was commissioned in 1994.
In the USA, no civil reprocessing plants are now operating, though three have been built. The first, a 300 t/yr plant at West Valley, New York, was operated successfully from 1966-72. However, escalating regulation required plant modifications which were deemed uneconomic, and the plant was shut down. The second was a 300 t/yr plant built at Morris, Illinois, incorporating new technology which, although proven on a pilot-scale, failed to work successfully in the production plant. It was declared inoperable in 1974. The third was a 1500 t/yr plant at Barnwell, South Carolina, which was aborted due to a 1977 change in government policy which ruled out all US civilian reprocessing as one facet of US non-proliferation policy. In all, the USA has over 250 plant-years of reprocessing operational experience, the vast majority being at government-operated defence plants since the 1940s.
In France a 400 t/yr reprocessing plant operated for metal fuels from gas-cooled reactors at Marcoule until 1997. At La Hague, reprocessing of oxide fuels has been done since 1976, and two 800 t/yr plants are now operating, with an overall capacity of 1700 t/yr.
French utility EDF has made provision to store reprocessed uranium (RepU) for up to 250 years as a strategic reserve. Currently, reprocessing of 1150 tonnes of EDF used fuel per year produces 8.5 tonnes of plutonium (immediately recycled as MOX fuel) and 815 tonnes of RepU. Of this about 650 tonnes is converted into stable oxide form for storage. EDF has demonstrated the use of RepU in its 900 MWe power plants, but it is currently uneconomic due to conversion costing three times as much as that for fresh uranium, and enrichment needing to be separate because of U-232 and U-236 impurities. The presence of the gamma-emitting U-232 requires shielding and so should be handled in dedicated facilities; and the presence of the neutron-absorbing U-236 isotope means that a higher level of enrichment is required compared with fresh uranium.
The plutonium is immediately recycled via the dedicated Melox mixed oxide (MOX) fuel fabrication plant. The reprocessing output in France is co-ordinated with MOX plant input, to avoid building up stocks of plutonium. If plutonium is stored for some years the level of americium-241, the isotope used in household smoke detectors, will accumulate and make it difficult to handle through a MOX plant due to the elevated levels of gamma radioactivity.
India has two 100 t/yr oxide fuel plants operating at Tarapur with another at Kalpakkam and a smaller one at Trombay, and Japan is starting up a major (800 t/yr) plant at Rokkasho while having had most of its used fuel reprocessed in Europe meanwhile. To 2006 it had a small (90 t/yr) reprocessing plant operating at Tokai Mura. 
Russia has an old 400 t/yr RT-1 oxide fuel reprocessing plant at Ozersk (Chelyabinsk), and the partly-built 3000 t/yr RT-2 plant at Zheleznogorsk (Krasnoyarsk) is being redesigned for completion about 2025. An underground military reprocessing plant there is decommissioned.

Reprocessing policies 

Conceptually reprocessing can take several courses, separating certain elements from the remainder, which becomes high-level waste. Reprocessing options include:
  • Separate U, Pu, (as today).
  • Separate U, Pu+U (small amount of U).
  • Separate U, Pu, minor actinidesf.
  • Separate U, Pu+Np, Am+Cm.
  • Separate U+Pu all together.
  • Separate U, Pu+actinides, certain fission products.
In today's reactors, reprocessed uranium (RepU) needs to be enriched, whereas plutonium goes straight to mixed oxide (MOX) fuel fabrication. This situation has two perceived problems: the separated plutonium is a potential proliferation risk, and the minor actinides remain in the separated waste, which means that its radioactivity is longer-lived than if it comprised fission products only.
As there is no destruction of minor actinides, recycling through light water reactors delivers only part of the potential waste management benefit. For the future, the focus is on removing the actinides from the final waste and burning them with the recycled uranium and plutonium in fast neutron reactors. (The longer-lived fission products may also be separated from the waste and transmuted in some other way.) Hence the combination of reprocessing followed by recycling in today’s reactors should be seen as an interim phase of nuclear power development, pending widespread use of fast neutron reactors.
All but one of the six Generation IV reactors being developed have closed fuel cycles which recycle all the actinides. Although US policy has been to avoid reprocessing, the US budget process for 2006 included $50 million to develop a plan for "integrated spent fuel recycling facilities", and a program to achieve this with fast reactors has become more explicit since.
In November 2005 the American Nuclear Society released a position statement4 saying that it "believes that the development and deployment of advanced nuclear reactors based on fast-neutron fission technology is important to the sustainability, reliability and security of the world's long-term energy supply." This will enable "extending by a hundred-fold the amount of energy extracted from the same amount of mined uranium". The statement envisages on-site reprocessing of used fuel from fast reactors and says that "virtually all long-lived heavy elements are eliminated during fast reactor operation, leaving a small amount of fission product waste which requires assured isolation from the environment for less than 500 years."
In February 2006 the US government announced the Global Nuclear Energy Partnership (GNEP) through which it would "work with other nations possessing advanced nuclear technologies to develop new proliferation-resistant recycling technologies in order to produce more energy, reduce waste and minimise proliferation concerns." GNEP goals included reducing US dependence on imported fossil fuels, and building a new generation of nuclear power plants in the USA. Two significant new elements in the strategy were new reprocessing technologies at advanced recycling centres, which separate all transuranic elements together (and not plutonium on its own) ­starting with the UREX+ process (see section on Developments of PUREX below), and 'advanced burner reactors' to consume the result of this while generating power.
GE Hitachi Nuclear Energy (GEH) is developing this concept by combining electrometallurgical separation (see section on Electrometallurgical 'pyroprocessing' below) and burning the final product in one or more of its PRISM fast reactors on the same site. The first two stages of the separation remove uranium which is recycled to light water reactors, then fission products which are waste, and finally the actinides including plutonium.
In mid-2006 a report5 by the Boston Consulting Group for Areva and based on proprietary Areva information showed that recycling used fuel in the USA using the COEX aqueous process (see Developments of PUREX below) would be economically competitive with direct disposal of used fuel. A $12 billion, 2500 t/yr plant was considered, with total capital expenditure of $16 billion for all related aspects. This would have the benefit of greatly reducing demand on space at the planned Yucca Mountain repository.
Boston Consulting Group gave four reasons for reconsidering US used fuel strategy which has applied since 1977:
  • Cost estimates for direct disposal at Yucca Mountain had risen sharply and capacity was limited (even if doubled)
  • Increased US nuclear generation, potentially from 103 to 160 GWe
  • The economics of reprocessing and associated waste disposal have improved
  • There is now a lot of experience with civil reprocessing.
Soon after this the US Department of Energy said that it might start the GNEP program using reprocessing technologies that "do not require further development of any substantial nature" such as COEX while others were further developed. It also flagged detailed siting studies on the feasibility of this accelerated "development and deployment of advanced recycling technologies by proceeding with commercial-scale demonstration facilities."

Reprocessing today – PUREX

All commercial reprocessing plants use the well-proven hydrometallurgical PUREX (plutonium uranium extraction) process. This involves dissolving the fuel elements in concentrated nitric acid. Chemical separation of uranium and plutonium is then undertaken by solvent extraction steps (neptunium – which may be used for producing Pu-238 for thermo-electric generators for spacecraft – can also be recovered if required). The Pu and U can be returned to the input side of the fuel cycle – the uranium to the conversion plant prior to re-enrichment and the plutonium straight to MOX fuel fabrication.

THORP
The Thermal Oxide Reprocessing Plant (THORP) at Sellafield , UK
The smaller black building to the rear is the vitrification plant.
(Sellafield Ltd.)

 Chemistry of Purex (see flowsheet below)
The used fuel is chopped up and dissolved in hot concentrated nitric acid. The first stage separates the uranium and plutonium in the aqueous nitric acid stream from the fission products and minor actinides by a countercurrent solvent extraction process, using tributyl phosphate dissolved in kerosene or dodecane. In a pulsed column uranium and plutonium enter the organic phase while the fission products and other elements remain in the aqueous raffinate. 
In a second pulsed column uranium is separated from plutonium by reduction with excess U4+ added to the aqueous stream. Plutonium is then transferred to the aqueous phase while the mixture of U4+ and U6+ remains in the organic phase. It is then stripped from the organic solvent with dilute nitric acid.
The plutonium nitrate is concentrated by evaporation then subject to an oxalate precipitation process followed by calcination to produce PuO2 in powder form. The uranium nitrate is concentrated by evaporation and calcined to produce UO3 in powder form. It is then converted to UO2 product by reduction in hydrogen. 



Alternatively, some small amount of recovered uranium can be left with the plutonium which is sent to the MOX plant, so that the plutonium is never separated on its own. This is known as the COEX (co-extraction of actinides) process, developed in France as a 'Generation III' process, but not yet in use (see next section). Japan's new Rokkasho plant uses a modified PUREX process to achieve a similar result by recombining some uranium before denitration, with the main product being 50:50 mixed oxides.
In either case, the remaining liquid after Pu and U are removed is high-level waste, containing about 3% of the used fuel in the form of fission products and minor actinides (Np, Am, Cm). It is highly radioactive and continues to generate a lot of heat. It is conditioned by calcining and incorporation of the dry material into borosilicate glass, then stored pending disposal. In principle any compact, stable, insoluble solid is satisfactory for disposal.

Developments of PUREX 

A modified version of the PUREX that does not involve the isolation of a plutonium stream is the UREX (uranium extraction) process. This process can be supplemented to recover the fission products iodine, by volatilisation, and technetium, by electrolysis. Research at the French Atomic Energy Commission (Commissariat à l'énergie atomique, CEA) has shown the potential for 95% and 90% recoveries of iodine and technetium respectively. The same research effort has demonstrated separation of caesium.
The US Department of Energy was developing the UREX+ processes under the Global Nuclear Energy Partnership (GNEP) program (see page on Global Nuclear Energy Partnership). In these, only uranium and technetium are recovered initially (in the organic phase) for recycle and the residual is treated to recover plutonium with other transuranics. The fission products then comprise most of the high-level waste. The central feature of this system was to increase proliferation resistance by keeping the plutonium with other transuranics – all of which are then destroyed by recycling in fast reactors.*  However, there are chemical safety problems with the Pu-Np recovery in the aqueous phase, and the process has been abandoned since 2008.
* Several variations of UREX+ have been developed, with the differences being in how the plutonium is combined with various minor actinides, and lanthanide and non-lanthanide fission products are combined or separated. UREX+1a combines plutonium with three minor actinides, but this gives rise to problems in fuel fabrication due to americium being volatile and curium a neutron emitter. Remote fuel fabrication facilities would therefore be required, leading to high fuel fabrication costs and requiring significant technological development. An alternative process, UREX+3, was therefore considered. This left only neptunium with the plutonium and the result is closer to a conventional MOX fuel. However, it is less proliferation-resistant than UREX+1a.
Energy Solutions holds the rights to PUREX in the USA and has developed NUEX, which separates uranium and then all transuranics (including plutonium) together, with fission products separately. NUEX is similar to UREX+1a but has more flexibility in the separations process.
Areva and CEA have developed three processes on the basis of extensive French experience with PUREX:
  • The COEX process based on co-extraction and co-precipitation of uranium and plutonium (and usually neptunium) together, as well as a pure uranium stream (eliminating any separation of plutonium on its own). It is close to near-term industrial deployment, and allows high MOX performance for both light-water and fast reactors. COEX may have from 20 to 80% uranium in the product, the baseline is 50%.
  • The DIAMEX-SANEX processes involving selective separation of long-lived radionuclides (with a focus on Am and Cm separation) from short-lived fission products. This can be implemented with COEX, following separation of U-Pu-Np. U-Pu and minor actinides are recycled separately in Generation IV fast neutron reactors.
  • The GANEX (grouped extraction of actinides) process co-precipitates some uranium with the plutonium (as with COEX), but then separates minor actinides and some lanthanides from the short-lived fission products. The uranium, plutonium and minor actinides together become fuel in Generation IV fast neutron reactors, the lanthanides become waste. It is being demonstrated at ATALANTE and La Hague from 2008 as part of a French-Japanese-US Global Actinide Cycle International Demonstration (GACID) with the product transmutation being initially in France's Phenix fast reactor (see Transmutation section below) and subsequently in Japan's Monju.
Initial work is at ATALANTEg at Marcoule, which started operation in 1992 to consolidate reprocessing and recycling research from three other sites. By 2012, it is expected to have demonstrated GANEX, and fabrication of oxide fuel pins combining U, Pu, Am, Np & Cm. Then work will proceed at La Hague on partitioning and fabrication of minor actinide-bearing fuels without the curium. From 2020 these will be irradiated in the Monju fast reactor, Japan.
All three processes were to be assessed in 2012, so that two pilot plants could be built to demonstrate industrial-scale potential:
  • One – possibly based on COEX – to make the driver fuel for the Generation IV reactor planned to be built by CEA by 2020.
  • One to produce fuel assemblies containing minor actinides for testing in Japan's Monju fast reactor and in France's Generation IV fast reactor.
In the longer term, the goal is to have a technology validated for industrial deployment of Generation IV fast reactors about 2040, at which stage the present La Hague plant will be due for replacement.
US research in recent years has focused on the TALSPEAK process which would come after a modified PUREX or COEX process to separate trivalent lanthanides from trivalent actinides, but this is only at bench scale so far. Originally in 1960s it was developed to separate actinides, notably Am & Cm from lanthanides.
Another alternative reprocessing technology being developed by Mitsubishi and Japanese R&D establishments is Super-DIREX (supercritical fluid direct extraction). This is designed to cope with uranium and MOX fuels from light water and fast reactors. The fuel fragments are dissolved in nitric acid with tributyl phosphate (TBP) and supercritical CO2, which results in uranium, plutonium and minor actinides complexing with TBP.
A new reprocessing technology is part of the reduced-moderation water reactor (RMWR) concept. This is the fluoride volatility process, developed in the 1980s, which is coupled with solvent extraction for plutonium to give Hitachi's Fluorex process. In this, 90-92% of the uranium in the used fuel is volatalised as UF6, then purified for enrichment or storage. The residual is put through a Purex circuit which separates fission products and minor actinides, leaving the unseparated U-Pu mix (about 4:1) to be made into MOX fuel.
Used MOX fuel can be handled through the PUREX process, though it contains more plutonium (especially even-numbered isotopes) and minor actinides than used U oxide fuel. In 1991-92 2.1 tonnes of MOX was reprocessed at Marcoule and 4.7 tonnes was reprocessed La Hague.

Partitioning goals

Several factors give rise to a more sophisticated view of reprocessing today, and use of the term partitioning reflects this. First, new management methods for high and intermediate-level nuclear wastes are under consideration, notably partitioning-transmutation (P&T) and partitioning-conditioning (P&C), where the prime objective is to separate long-lived radionuclides from short-lived ones. Secondly, new fuel cycles such as those for fast neutron reactors (including a lead-cooled one) and fused salt reactors, and the possible advent of accelerator-driven systems, require a new approach to reprocessing. Here the focus is on electrolytic processes ('pyroprocessing') in a molten salt bath. The term 'electrometallurgical' is also increasingly used to refer to this in the USA.
The main radionuclides targeted for separation for P&T or P&C are the actinides neptunium, americium and curium (along with U & Pu), and the fission products iodine-129, technetium-99, caesium-135 and strontium-90. Removal of the latter two significantly reduces the heat load of residual conditioned wastes. In Japan, platinum group metals are also targeted, for commercial recovery. Of course any chemical process will not separate different isotopes of any particular element.
Efficient separation methods are needed to achieve low residuals of long-lived radionuclides in conditioned wastes and high purities of individual separated ones for use in transmutation targets or for commercial purposes (e.g. americium for household smoke detectors). If transumation targets are not of high purity then the results of transmutation will be uncertain. In particular fertile uranium isotopes (e.g. U-238) in a transmutation target with slow neutrons will generate further radiotoxic transuranic isotopes through neutron capture.
Achieving effective full separation for any transmutation program is likely to mean electrolytic processing of residuals from the PUREX or similar aqueous processes.
A BNFL-Cogema study in 2001 reported that 99% removal of actinides, Tc-99 & I-129 would be necessary to justify the effort in reducing the radiological load in a waste repository. A US study identified a goal of 99.9% removal of the actinides and 95% removal of technetium and iodine. In any event, the balance between added cost and societal benefits is the subject of considerable debate.

Electrometallurgical 'pyroprocessing' 

Electrolytic/electrometallurgical processing techniques ('pyroprocessing') to separate nuclides from a radioactive waste stream have been under development in the US Department of Energy laboratories, notably Argonne, as well as by the Korea Atomic Energy Research Institute (KAERI) in conjunction with work on DUPIC (see section on Recycled LWR uranium and used fuel in PHWRs below). Their main development has possibly been in Russia, where they are to be the mainstay of closing the fuel cycle fully by about 2020. There has been particular emphasis on fast reactor fuels.
So-called pyroprocessing involves several stages including: volatilisation; liquid-liquid extraction using immiscible metal-metal phases or metal-salt phases; electrolytic separation in molten salt; and fractional crystallisation. They are generally based on the use of either fused salts such as chlorides or fluorides (eg LiCl+KCl or LiF+CaF2) or fused metals such as cadmium, bismuth or aluminium. They are most readily applied to metal rather than oxide fuels, and are envisaged for fuels from Generation IV reactors.
Electrometallurgical 'pyroprocessing' can readily be applied to high burn-up fuel and fuel which has had little cooling time, since the operating temperatures are high already. However, such processes are at an early stage of development compared with hydrometallurgical processes already operational.
Separating (partitioning) the actinides contained in a fused salt bath is by electrodeposition on a cathode, so involves all the positive ions without the possibility of chemical separation of heavy elements such as in PUREX and its derivatives. This cathode product can then be used in a fast reactor.
So far only one electrometallurgical technique has been licensed for use on a significant scale. This is the IFR (integral fast reactor) electrolytic process developed by Argonne National Laboratory in the USA and used for pyroprocessing the used fuel from the EBR-II experimental fast reactor which ran from 1963-1994. This application is essentially a partitioning-conditioning process, because neither plutonium nor other transuranics are recovered for recycle. The process is used to facilitate the disposal of a fuel that could not otherwise be sent directly to a geologic repository. The used uranium metal fuel is dissolved in a LiCl+KCl molten bath, the U is deposited on a solid cathode, while the stainless steel cladding and noble metal fission products remain in the salt, and are consolidated to form a durable metallic waste. The transuranics and fission products in salt are then incorporated into a zeolite matrix which is hot pressed into a ceramic composite waste. The highly-enriched uranium recovered from the EBR-II driver fuel is down-blended to less than 20% enrichment and stored for possible future use.
The PYRO-A process, being developed at Argonne to follow the UREX process, is a pyrochemical process for the separation of transuranic elements and fission products contained in the oxide powder resulting from denitration of the UREX raffinate. The nitrates in the residual raffinate acid solution are converted to oxides, which are then reduced electrochemically in a LiCl-Li2O molten salt bath. The more chemically active fission products (eg Cs, Sr) are not reduced and remain in the salt. The metallic product is electrorefined in the same salt bath to separate the transuranic elements on a solid cathode from the rest of the fission products. The salt bearing the separated fission products is then mixed with a zeolite to immobilize the fission products in a ceramic composite waste form. The cathode deposit of transuranic elements is then processed to remove any adhering salt and is formed into ingots for subsequent fabrication of transmutation targets.
The PYRO-B process, has been developed for the processing and recycle of used fuel from a transmuter (fast) reactor. A typical transmuter fuel is free of 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.
The KAERI advanced spent fuel conditioning process (ACP) involves separating uranium, transuranics including plutonium, and fission products including lanthanides. It utilises a high-temperature lithium-potassium chloride bath from which uranium is recovered electrolytically to concentrate the actinides, which are then removed together (with some remaining fission products). The latter product is then fabricated into fast reactor fuel without further treatment. The process is intrinsically proliferation-resistant because it is so hot radiologically, and the curium provides a high level of spontaneous neutrons. It recycles about 95% of the used fuel. Development of this process is at the heart of US-South Korean nuclear cooperation, and is central to the renewal of the bilateral US-South Korean nuclear cooperation agreement in March 2014, so is already receiving considerable attention in negotiations.
With US assistance through the International Nuclear Energy Research Initiative (I-NERI) program KAERI built the Advanced Spent Fuel Conditioning Process Facility (ACPF) at KAERI. KAERI hopes the project will be expanded to engineering scale by 2012, leading to the first stage of a Korea Advanced Pyroprocessing Facility (KAPF) starting in 2016 and becoming a commercial-scale demonstration plant in 2025.
South Korea has declined an approach from China to cooperate on electrolytic reprocessing, and it has been rebuffed by Japan's CRIEPI due to government policy.
Russian pyroprocessing consists of three main stages: dissolution of the used nuclear fuel in molten salts, precipitation of plutonium dioxide or electrolytic deposition of uranium and plutonium dioxides from the melt, then processing the material deposited on the cathode or precipitated at the bottom of the melt for granulated fuel production. The process recovers the cathode deposits without changing their chemical composition or redistributing the plutonium. All used fuel is reprocessed with the goal of having a complete recycle of plutonium, neptunium, americium, and curium as well as the uranium. This process, combined with vibropacking* in fuel fabrication will be used to produce fuel for the BN-800 fast reactor. The technologies complement one another well and involve high levels of radioactivity throughout, making them self-protecting against diversion or misuse.
Vibropacked MOX fuel (VMOX) is seen as the way forward. This is made by agitating a mechanical mixture of (U, Pu)O2 granulate and uranium powder, which binds up excess oxygen and some other gases (that is, operates as a getter) and is added to the fuel mixture in proportion during agitation. The getter resolves problems arising from fuel-cladding chemical interactions. The granules are crushed UPuO2 cathode deposits from pyroprocessing. VMOX needs to be made in hot cells. It has been used in BOR-60 since 1981 (with 20-28% Pu), and tested in BN-350 and BN-600.
A pilot scale pyroprocessing demonstration facility for fast reactor fuel has been developed by the Russian Institute of Atomic Reactors (RIAR) at Dimitrovgrad.
GE Hitachi is designing an Advanced Recycling Centre (ARC) which integrates electrometallurgical processing with its PRISM fast reactors. The main feed is used fuel from light water reactors, and the three products are fission products, uranium, and transuranics (Np, Pu, Am, Cm), which become fuel for the fast reactors (with some of the uranium). The uranium can be re-enriched or used as fuel for Candu reactors. As the cladding reaches its exposure limits, used PRISM fuel is recycled after removal of fission products. A full commercial-scale ARC would comprise an electrometallurgical plant and three power blocks of 622 MWe each (six 311 MWe reactor modules), but a "full-scale building block" of ARC is a 50 t/yr electrometallurgical plant coupled to one 311 MWe reactor module, with breeding ratio of 0.8.

Recycled LWR uranium and used fuel in PHWRs 

The established approach to using RepU is recycling it through conversion and enrichment, for light water reactors. Another approach to used nuclear fuel recycling is directing recycled uranium (referred to as RepU, reprocessed uranium), or actual used light water reactor (LWR) fuel, into pressurized heavy water reactors (PHWRs). This may be directly using RepU, or by blending RepU with depleted uranium to give natural uranium equivalent (NUE), or by direct use of used PWR fuel in CANDU reactors (DUPIC).
PHWRs (such as CANDU reactors) normally use as fuel natural uranium which has not undergone enrichment and so can operate fuelled by the uranium and plutonium that remains in used fuel from LWRs.  This might typically contain about 0.5 to 0.9% U-235 and 0.6% Pu-239 but with significant neutron absorbers.
In unit 1 of the Qinshan Phase III plant in China, there has been a demonstration using fuel bundles with RepU from PWRs blended with depleted uranium to give natural uranium equivalent (NUE) fuel with 0.71% U-2356.  It behaved the same as natural uranium fuel. Subject to supply from reprocessing plants, a full core of natural U equivalent (NUE) is envisaged. Following design, licensing, etc, full core implementation in both China's CANDU reactors is envisaged by the end of 2013.  (Recycled plutonium will be used in MOX fuel for fast reactors.)
AECL says that it is also possible to use the RepU directly in CANDUs, without blending it down, and Qinshan III envisages this possibility with RepU having 0.9% U-235.
With DUPIC, the direct use of used PWR fuel as such in CANDUs, used fuel assemblies from LWRs would be dismantled and refabricated into fuel assemblies the right shape for use in a CANDU reactor. This could be direct, involving only cutting the used LWR fuel rods to CANDU length (about 50 cm), resealing and re-engineering into cylindrical bundles suitable for CANDU geometry.
Alternatively, a "dry reprocessing" technology has been developed which removes only the volatile fission products from the used LWR fuel mix. After removal of the cladding, a thermal-mechanical process is used to reduce the used LWR fuel pellet to a powder. This could have more fresh natural uranium added, before being sintered and pressed into CANDU pellets.  It would contain all the actinides and most of the fission products from irradiation in LWR.
The DUPIC technique has certain advantages:
  • No materials are separated during the refabrication process. Uranium, plutonium, fission products and minor actinides are kept together in the fuel powder and bound together again in the DUPIC fuel bundles.
  • A high net destruction rate can be achieved of actinides and plutonium.
  • Up to 25% more energy can be realised compared to other PWR used fuel recycling techniques.
  • And a DUPIC fuel cycle could reduce a country's need for used PWR fuel disposal by 70% while reducing fresh uranium requirements by 30%.
However, as noted above, used nuclear fuel is highly radioactive and generates heat. This high activity means that the DUPIC manufacture process must be carried out remotely behind heavy shielding. While these restrictions make the diversion of fissile materials much more difficult and hence increase security, they also make the manufacture process more complex compared with that for the original PWR fuel, which is barely radioactive before use.  (NUE would be more radioactive than natural U, due to U-232 in the RepU.)
Canada, which developed the CANDU reactor, and South Korea, which hosts four CANDU units as well as many PWRs, have initiated a bilateral joint research program to develop DUPIC.  This also involves the USA.
The Korean Atomic Energy Research Institute (KAERI) has had a development program since 1992 to demonstrate the DUPIC fuel cycle concept. KAERI believes that although it is too early to commercialise the DUPIC fuel cycle, the key technologies are in place for a practical demonstration of the technique. Challenges which remain include the development of a technology to produce fuel pellets of the correct density, the development of remote fabrication equipment and the handling of the used PWR fuel. However, KAERI successfully manufactured DUPIC small fuel elements for irradiation tests inside the HANARO research reactor in April 2000 and fabricated full-size DUPIC elements in February 2001. AECL is also able to manufacture DUPIC fuel elements.
Research is also underway on the reactor physics of DUPIC fuel and the impacts on safety systems. 
A further complication is the loading of highly radioactive DUPIC fuel into the CANDU reactor. Normal fuel handling systems are designed for the fuel to be hot and highly radioactive only after use, but it is thought that the used fuel path from the reactor to cooling pond could be reversed in order to load DUPIC fuel, and studies of South Korea's Wolsong CANDU units indicate that both the front- and rear-loading techniques could be used with some plant modification. 

Transmutation    

The objective of transumutation is to change (long-lived) actinides into fission products and long-lived fission products into significantly shorter-lived nuclides. The goal is to have wastes which become radiologically innocuous in only a few hundred years. The need for a waste repository is certainly not eliminated, but it can be smaller and simpler and the hazard posed by the disposed waste materials is greatly reduced.
Transmutation of one radionuclide into another is achieved by neutron bombardment in a nuclear reactor or accelerator-driven device. In the latter, a high-energy proton beam hitting a heavy metal target produces a shower of neutrons by spallationh. The neutrons can cause fission in a subcritical fuel assembly, but unlike a conventional reactor, fission ceases when the accelerator is turned off. The fuel may be uranium, plutonium or thorium, possibly mixed with long-lived wastes from conventional reactors. See also page on Accelerator-Driven Nuclear Energy.
Transmutation is mainly initiated by fast neutrons. Since these are more abundant in fast neutron reactors, such reactors are preferred for transmutation. Some radiotoxic nuclides, such as Pu-239 and the long-lived fission products Tc-99 and I-129, can be transmuted (fissioned, in the case of Pu-239) with thermal (slow) neutrons. However, a 2001 BNFL-Cogema study found that full transmutation in a light water reactor would take at least several decades, and recent research has focused on use of fast reactors. The minor actinides Np, Am and Cm (as well as the higher isotopes of plutonium), all highly radiotoxic, are much more readily destroyed by fissioning in a fast neutron energy spectrum, where they can also contribute to the generation of power.
One of the main functions of France's Phenix fast neutron reactor in its last two years of operation was test burning fuel assemblies containing high concentrations of minor actinides. From mid-2007 it irradiated four fuel pins containing actinides from the US Department of Energy, two from the CEA, and two from the European Commission's Institute for Transuranics.


Further Information 

Notes 

a. Used fuel from light water reactors (at normal US burn-up levels) contains approximately:
  • 95.6% uranium, over 98.5% of which is U-238 (the remainder consists of: trace amounts of U-232 and U-233; less than 0.02% U-234; 0.5-1.0% U-235; around 0.5% U-236; and around 0.001% U-237 – which accounts for nearly all of the activity)
  • 2.9% stable fission products
  • 0.9% plutonium
  • 0.3% caesium & strontium (fission products)
  • 0.1% iodine and technetium (fission products)
  • 0.1% other long-lived fission products
  • 0.1% minor actinides (americium, curium, neptunium)
[Back]
b. For the Dutch Borssele reactor which normally uses 4.4% enriched fuel, compensated enriched reprocessed uranium (c-ERU) is 4.6% enriched to compensate for U-236. [Back]
c. Since Magnox fuel was not enriched in the first place, this is actually known as Magnox depleted uranium (MDU), which assayed about 0.4% U-235. The MDU was converted to UF6, enriched to 0.7% at BNFL's Capenhurst diffusion plant and then to 2.6% to 3.4% at Urenco's centrifuge plant. Until the mid 1990s some 60% of all AGR fuel was made from MDU and it amounted to about 1650 tonnes of low enriched uranium. Although used Magnox fuel continues to be reprocessed, recycling of MDU was discontinued in 1996 due to economic factors. [Back]
d. At anything over about 20 GWday/t burn-up the plutonium is considered to be 'reactor grade' and significantly different from weapons grade material. Some figures for the Oskarshamn 3 nuclear unit: with 30 GWd/t burn-up, 69% Pu is fissile; 40 GWd/t, 61% fissile; 50 GWd/t, 55% fissile; and 60 GWD/t, 50% fissile. [Back]
e. See Note c above. [Back]
f. Minor actinides are americium and curium (95 & 96 in periodic table), sometimes also neptunium (93). The major actinides are plutonium (94) and uranium (92). [Back]
g. Atelier Alpha et Laboratoire pour les Analyses de Transuraniens et Etudes de retraitement, Alpha shop and laboratory for the analysis of transuranics and reprocessing studies. [Back]
h. Spallation is the process where nucleons are ejected from a heavy nucleus being hit by a high energy particle. In this case, a high-enery proton beam directed at a heavy target expels a number of spallation particles, including neutrons. [Back]

References 

1. Sources: Nuclear Engineering International Handbook 2007 [Back]
2. Nuclear Energy Data 2007, OECD Nuclear Energy Agency (ISBN 9789264034532) [Back]
3. Nuclear Energy Data 2007, OECD Nuclear Energy Agency (ISBN 9789264034532) [Back]
4. Fast Reactor Technology: A Path to Long-Term Energy Sustainability – Position Statement, November 2005, American Nuclear Society [Back]
5. Economic Assessment of Used Nuclear Fuel Management in the United States, Prepared by the Boston Consulting Group for AREVA (July 2006) [Back]
6. Chinese Candu reactor trials uranium reuse, World Nuclear News (24 March 2010) [Back]

General sources

Charles Madic, Overview of the Hydrometallurgical and Pyro-metallurgical Processes Studied Worldwide for the Partitioning of High Active Nuclear Wastes, NEA/OECD 6th Information Exchange Meeting on Actinide and Fission Product Partitioning and Transmutation, Madrid, Spain (11-13 December 2000)
James Laidler, Pyrochemical Separations Technologies Envisioned for the U.S. Accelerator Transmutation of Waste System, NEA/OECD Workshop on Pyrochemical Separations, Avignon, France (14-15 March 2000)
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Jang Jin Park et al, Technology and Implementation of the DUPIC Concept for Spent Nuclear Fuel in the ROK, LLNL Nuclear Cooperation Meeting on Spent Fuel and High-Level Waste Storage and Disposal, Las Vegas, Nevada, USA (7-9 March 2000)
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Janet Wood, Should USA Reprocess?, Nuclear Engineering International (September 2006)
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Management of Reprocessed Uranium – Current Status and Future Prospects, IAEA TECDOC 1529 (2007), International Atomic Energy Agency (ISBN: 920114606X)
Pyrochemical Separations in Nuclear Applications – A Status Report, NEA #5427 (2004), Nuclear Energy Agency (ISBN: 9264020713)
Reprocessing of spent oxide fuel from nuclear power reactors. P.Netter, Areva, in Nuclear Fuel Science and Engineering, Woodhead Publishing, 2012

Military Warheads as a Source of Nuclear Fuel

(updated August 2013)
http://www.world-nuclear.org/info/Nuclear-Fuel-Cycle/Uranium-Resources/Military-Warheads-as-a-Source-of-Nuclear-Fuel/#.UhXWd3-N6So
  • Weapons-grade uranium and plutonium surplus to military requirements in the USA and Russia is being made available for use as civil fuel.
  • Weapons-grade uranium is highly enriched, to over 90% U-235 (the fissile isotope). Weapons-grade plutonium has over 93% Pu-239 and can be used, like reactor-grade plutonium, in fuel for electricity production.
  • Highly-enriched uranium from weapons stockpiles has been displacing some 9720 tonnes of U3O8 production from mines each year, and meets about 13% of world reactor requirements through to 2013.
For more than four decades concern has centred on the possibility that uranium intended for commercial nuclear power might be diverted for use in weapons. Today, however, attention is focused on the role of military uranium as a major source of fuel for commercial nuclear power.
Since 1987 the United States and countries of the former USSR have signed a series of disarmament treaties to reduce the nuclear arsenals by about 80%.
Nuclear materials declared surplus to military requirements by the USA and Russia are now being converted into fuel for commercial nuclear reactors. With the disintegration of the Soviet Union a unique opportunity arose to deploy military weapons material for making electricity. A 1993 agreement covered essentially the enrichment component of this material, but left unresolved the question of feed from mines, and a 1999 agreement dealt with what happened to the feed material.
The main weapons material is highly enriched uranium (HEU), containing at least 20% uranium-235 (U-235) and usually about 90% U-235. HEU can be blended down with uranium containing low levels of U-235 to produce low enriched uranium (LEU), typically less than 5% U-235, fuel for power reactors. It is blended with depleted uranium (mostly U-238), natural uranium (0.7% U-235), or partially-enriched uranium.
Highly-enriched uranium in US and Russian weapons and other military stockpiles amounts to about 2000 tonnes, equivalent to about twelve times annual world mine production.
World stockpiles of weapons-grade plutonium are reported to be some 260 tonnes, which if used in mixed oxide fuel in conventional reactors would be equivalent to a little over one year's world uranium production. Military plutonium can blended with uranium oxide to form mixed oxide (MOX) fuel.
After LEU or MOX is burned in power reactors, the spent fuel is not suitable for weapons manufacture.

Megatons to Megawatts

Commitments by the USA and Russia to convert nuclear weapons into fuel for electricity production is known as the Megatons to Megawatts program.
Surplus weapons-grade HEU resulting from the various disarmament agreements led in 1993 to an agreement between the US and Russian governments. Under this Russia would convert 500 tonnes of HEU from warheads and military stockpiles (equivalent to around 20,000 bombs) to LEU to be bought by the USA for use in civil nuclear reactors.
In 1994, a US$12 billion implementing contract was signed between the US Enrichment Corporation (now USEC Inc) and Russia's Technabexport (Tenex) as executive agents for the US and Russian governments. USEC purchased a minimum of 500 tonnes of weapons-grade HEU over 20 years to 2013, at a rate of up to 30 tonnes/year from 1999. The HEU is progressively blended down to 15,259 t of LEU at 4.4% U-235 in Russia, using 1.5% U-235 (re-enriched depleted uranium tails), to restrict levels of U-234 in the final product.* USEC can then sell the LEU to its utility customers as fuel. The LEU is equivalent to about 140,000 tonnes of natural uranium from mines (depending on assumptions about enrichment), or 9720 t/yr of U3O8.
* HEU metal is first removed from a warhead, machined into shavings, oxidized and fluorinated. The resulting highly enriched uranium hexafluoride is then mixed in a gaseous stream with slightly enriched uranium to form LEU suitable for commercial nuclear reactors. The LEU is transferred to shipping cylinders and sent to a collection point in St. Petersburg, Russia. USEC takes possession of the material there and ships it to the USA where it is included in USEC’s inventory for delivery to customers.
In 1999 a supplementary agreement addressed what should happen to the uranium feed from mines, against which the blended down LEU was supplied to customers. See section below.
By September 2009 a total of 375 tonnes HEU had produced nearly 10,868 tonnes of low-enriched fuel under a market-based pricing formula. By August 2011 the total had risen to 425 tonnes HEU, equivalent to 17,000 nuclear warheads according to USEC, which had paid over $7.2 billion to the Russian Federation. By June 2013 the total was 475 tonnes HEU and 19,000 warheads, and the 500 tonnes completion is expected in November, equivalent to 20,000 warheads and 89 million SWU, with about $8 billion paid to Russia for this.
For its part, the US Government has declared just over 174 tonnes of HEU (of various enrichments) to be surplus from military stockpiles. Of this, USEC has taken delivery of 14.2 tonnes in the form of uranium hexafluoride (UF6) containing around 75% U-235, and 50 tonnes as uranium oxide or metal containing around 40% U-235. Downblending of the UF6 was completed in 1998, to produce 387 tonnes of LEU. Some 13.5 tonnes of the HEU oxide or metal had been processed by September 2001 to produce 140.3 tonnes of LEU. In 2004 the Nuclear Regulatory Commission issued a licence for downblending 33 tonnes HEU by Nuclear Fuel Services in Tennessee and in 2005 the first delivery was made to a TVA power plant.
DOE's National Nuclear Security Administration (NNSA)* in 2005 announced that it was committing about 40 tonnes of off-specification HEU (with elevated levels of U-236) to the Blended Low-Enriched Uranium (BLEU) program. This material would be used by TVA. In 2008 NNSA was negotiating with TVA to release a further 21 tonnes of HEU under the program, which would yield about 250 tonnes of LEU, some of which might be sold to other utilities.
* Established by Congress in 2000, NNSA is a separately organized agency within the U.S. Department of Energy responsible for enhancing national security through the military application of nuclear science. NNSA maintains and enhances the safety, security, reliability and performance of the U.S. nuclear weapons stockpile without nuclear testing; works to reduce global danger from weapons of mass destruction; provides the U.S. Navy with safe and effective nuclear propulsion; and responds to nuclear and radiological emergencies in the United States and abroad.
In mid 2007 the NNSA awarded contracts to Nuclear Fuel Services and Wesdyne International to downblend 17.4 tonnes of HEU from dismantled warheads to be part of a new international fuel reserve. NFS is downblending the material in Tennessee to yield some 290 tonnes of LEU (4.95% U-235) by early 2012. Wesdyne, the prime contractor, will then store 230 tonnes of the LEU at the Westinghouse fuel fabrication plant in South Carolina to be available for the American Assured Fuel Supply (AFS) program. It will take about 60 t as payment in kind, to be sold on the market over a three to four year period. This first batch of LEU will be available for use in civilian reactors by nations in good standing with the International Atomic Energy Agency that have good nonproliferation credentials and are not pursuing uranium enrichment and reprocessing technologies. It will also now be available to domestic utilities. The fuel – worth some $500 million – will be sold at the current market price.
In June 2009 NNSA awarded a further contract ($209 million) to NFS and Wesdyne for 12.1 tonnes of HEU which will yield some 220 tonnes of LEU by 2012. This second batch of LEU is to provide fuel supply assurance for utilities which participate in DOE's mixed-oxide fuel program utilising surplus plutonium from US weapons. (The AFS scheme is consistent with international concerns to limit the spread of enrichment technology to countries without well-established nuclear fuel cycles. Russia has agreed to join the initiative.)
In the short term most US military HEU is likely to be blended down to 20% U-235, then stored. In this form it is not useable for weapons.

Market Impact

Overall, the blending down of 500 tonnes of Russian weapons HEU will result in about 15,000 tonnes of LEU over 20 years. This is equivalent to about 152,000 tonnes of natural U, or just over twice annual world demand.
From 2000 to 2013 the dilution of 30 tonnes of military HEU is displacing about 10,600 tonnes of uranium oxide mine production per year, which represents some 13% of world reactor requirements.
Under the 1994 Agreement, USEC recognised the need to release the diluted military uranium to nuclear utilities in such a way as not to impact negatively on the US uranium market.

How the Market Works 

Normally, a utility buys natural uranium from a mining company as "yellowcake" (U3O8) and has it converted to UF6. It then supplies this feed to USEC, paying them for the enrichment component. USEC runs its energy-intensive enrichment plant to separate an appropriate amount of enriched uranium (eg at 3.5 - 5.0% U-235, leaving a lot of depleted uranium). USEC then returns the enriched uranium to the utility for its reactor.

Enrichment market

A different, and somewhat complicated, system is used for the Russian material. The utility supplies the feed component of natural uranium as before and pays USEC for the enrichment component. But instead of running their plant, USEC pays the Russians for some blended-down weapons uranium and passes this on to the customer utility as "enriched" uranium fuel. The customers receive the blended-down Russian material, paid for as if it were their own uranium which had been enriched.

Blending down

USEC pays Russia for the enrichment services component (basically energy) of the low-enriched product it receives. This amounts to about 5.5 million SWU per year. Russia takes ownership of the corresponding amount of natural uranium "feed" provided to USEC by its utility customers for toll enrichment services. Under the 1999 agreement (below) at least 72% of the feed is sold to Cameco, Areva and Nukem in the proportion 45/45/10, and the remainder is sent to Russia for domestic use there. In 2009 Rosatom said that its portion of the natural uranium feed to date – worth US$ 2.7 billion – had been received in Russia.

1999 Market Agreement re natural uranium feed to USEC

After years of stalled negotiations on this matter, a major agreement was approved early in 1999 by the US and Russian governments. It involved 163,000 tonnes of natural U3O8 feed to be supplied over the remaining 15 years of the US-Russian HEU agreement.
Cameco, Cogema (now Areva), and Nukem signed the commercial agreement with Tenex of Russia, giving them "exclusive options to purchase" 118,000 tonnes of this (nominally 70%), leaving the remainder "available to Tenex". One important stipulation was that stockpiles, each of some 26,000 tonnes U3O8, would be held by both Russian and US governments for ten years, to 2009. The US stockpile already existed, Russia's was built up over the next few years from all feed not purchased by Tenex or an associate, and Russia was free to sell only what exceeded this.
The new agreement did not change the overall supply and demand situation, but it removed some major uncertainties over how the material would be released to the market.
The 1993 agreement significantly depressed uranium exploration activities and the uranium price, which took until about 2003 to recover.

Plutonium and MOX

Disarmament will also give rise to some 150-200 tonnes of weapons-grade plutonium (Pu). Weapons-grade plutonium has over 93% of the fissile isotope, Pu-239, and can be used, like reactor-grade Pu, in fuel for electricity production. Options considered for it included:
  • Immobilisation with high-level waste - treating plutonium as waste,
  • Fabrication with uranium oxide as a MOX fuel for burning in existing reactors,
  • Fabrication with thorium as a fuel for existing Russian reactors,
  • Fuelling fast-neutron reactors.
In 1994 the USA announced that about half of its military plutonium stockpile was surplus to military requirements. This included non-pit material, and about 20 tonnes of it was of such quality that it might not be possible to utilise it for MOX. In 2012 the US DOE plutonium inventory was announced as being 95.4 tonnes as of 2009, after some transuranic wastes containing 4.8 t of plutonium had been disposed of at the Waste Isolation Pilot Plant (WIPP) in New Mexico. The 95.4 t was composed of 81.3 t weapons grade, 12.7 t fuel grade, and 1.4 t reactor grade plutonium. Since 1944, 103.4 tonnes has been produced, and a further 8.0 t had come from industry, foreign countries and research reactors. Some 3.4 tonnes had been used in tests, 7.8 t discarded as waste and 2.8 t otherwise removed. Russia’s inventory of weapons-grade plutonium is unknown, but is assumed to be of the same order as USA’s.
In June 2000, the USA and Russia agreed to dispose of 34 tonnes each of weapons-grade plutonium by 2014. The US undertook to pursue a dual track program (immobilisation and MOX), self-funded, while the G-7 nations were to provide some US$ 2.5 billion to set up Russia's program. The latter was initially MOX-oriented for VVER reactors, the high cost being because this was not part of Russia's fuel cycle policy. A revised agreement signed in April 2010 allows the Russian plutonium to be used in BN-800 fast neutron reactors, and stretches the timeline to 2018. However, the G7 funding is not available on this basis and Russia will fund most of the program, with the USA contributing $400 million. The 68 tonnes of plutonium in both countries is equivalent to about 12,000 tonnes of natural uranium.
Weapons-grade plutonium entering the civil fuel cycle needs to be kept under very tight security, and there are some technical measures needed to achieve this. MOX fuel made from it should degrade it so that Pu-239 cannot be extracted. As it became clear that this could be achieved, the USA dropped its immobilisation plans for most military plutonium,* and this is reflected in the April 2010 agreement with Russia.
* Some detail on immobilisation is in the Synroc paper.

USA plans

After environmental and safety reviews, the US Nuclear Regulatory Commission authorised construction of a MOX fuel fabrication plant (MFFF) at the DOE Savannah River site in South Carolina by Duke, Cogema, Stone & Webster. Construction started in August 2007, by Shaw Areva MOX Services. It will make about 1700 civil MOX fuel assemblies from depleted uranium and at least 34 tonnes of weapons-grade plutonium, unlike other MOX plants which use fresh reactor-grade plutonium having around one third non-fissile plutonium isotopes. US reactors using MOX fuel will need to licensed for it. The MFFF is designed to turn 3.5 t/yr of weapons-grade plutonium into about 150 MOX fuel assemblies, both PWR and BWR.
Shaw Areva MOX Services is under contract to the National Nuclear Security Administration (NNSA), which will own the plant, expected to be in operation from 2016. The high cost of the plant – $3.5 billion plus $1.3 billion contingency and $183 million per year to operate - is justified on non-proliferation grounds. Annual cost will be offset by revenue. The following is a comment on this US situation from Dr C. Wolfe, former chairman of the Technical Advisory Panel to the Department of Energy's Plutonium Focus Area, whose task had been to advise on technology to enable the disposition of the excess plutonium: In discussion with Russia "the USA often emphasized elaborate technology schemes to immobilize the plutonium in a proliferation-resistant state. These included grouts, synthetic rock, glass and co-disposal with spent nuclear fuel. The Russians were astounded. They couldn't believe that we were willing to take this material, which we had spent billions of dollars producing, and just throw it away. Not only throw it away, but spend a lot of additional money to get rid of it. The Russians saw it for what it was: a tremendous energy resource. The US eventually came to the same conclusion and opted for converting 34 metric tons of weapons-grade plutonium into MOX fuel to provide electrical energy for the US economy." (Aiken Standard 10/8/09)
In June 2005 the first four fuel assemblies with mixed oxide fuel made from US military plutonium (plus depleted uranium) started generating electricity in Duke Power's Catawba-1 nuclear power plant in South Carolina, on a trial basis. They incorporated 140 kg of weapons-grade plutonium. The plutonium was made into 2 tonnes of pellets at the Cadrache plant and then fabricated into fuel assemblies at the Melox plant in France. This trial was concluded satisfactorily.
DOE was moving all its surplus non-pit weapons plutonium – reported to be 12.8 tonnes – to Savannah River by 2010. Once the material is consolidated there, the Department's plans for disposing of it involve the use of three Savannah River site facilities: the MOX Fuel Fabrication Facility (MFFF, under construction) for 7.8 tonnes, the existing H-Canyon processing plant followed by a proposed new small-scale plutonium vitrification plant for the balance of 5.0 tonnes. The H-Canyon facility* is the last such US plant able to treat used HEU fuel and similar materials still operational.
* H-Canyon dates from 1955 and originally recovered uranium, neptunium and plutonium from used military and research reactor HEU fuel. Since 1998 it has recovered HEU from degraded materials and spent fuel, to recycle it as LEU. This program will continue to 2019.
Following DOE's September 2007 addition of 9 tonnes of plutonium from dismantled weapons to the MOX program, making 43.4 tonnes surplus to defence needs and designated for MOX, NNSA decided that the Savannah River plant might also produce starter fuel for advanced fast reactors, part of the advanced fuel cycle initiative (AFCI) program.
Meanwhile the US has developed a "spent fuel standard". This specifies that plutonium should never be more accessible than if it were incorporated in spent fuel and thus protected from interference by strong gamma radiation. The plutonium immobilisation plant, if and when it is eventually built, would thus incorporate the plutonium in a version of Synroc ( artificial rock), and encase small discs of this in canisters of vitrified high-level radioactive waste. Alternatively, plutonium would be mixed with fission products and vitrified at the small plant proposed for Savannah River.
Europe's well-developed MOX capacity suggests that weapons plutonium could be disposed of relatively quickly. Input weapons-grade plutonium might need to be mixed with reactor grade material or blended with Pu-238, but using such MOX as 30% of the fuel in one third of the world's reactor capacity would remove about 15 tonnes of warhead plutonium per year. This would amount to burning 3000 warheads per year to produce 110 billion kWh of electricity.
Over 35 reactors in Europe are licensed to use MOX fuel, and 22 French reactors are licensed to use it as 30% of their fuel.

Russian plans

Russia intends to use its plutonium to fuel fast neutron reactors such as its BN-600 and BN-800, and later BREST at Beloyarsk. The USA earlier insisted that it duplicate US plans to make it into MOX fuel for late-model conventional reactors, and for this Russia insisted that the USA pay all costs. But after announcement of the Global Nuclear Energy Partnership in 2006 with its proposals for use of fast reactors, US objection to Russian plans disappeared. The 34 tonnes of plutonium initially available for MOX would have been enough for 1350 fuel assemblies for light-water reactors, but will now go into MOX fuel for BN-600 and BN-800 fast reactors – the former with one third MOX core and the latter with full MOX core, and accounting for most of the usage. The USA has agreed to contribute US$ 400 million towards the cost of this – much less than for the MOX option in VVER reactors.
A 60 t/yr commercial MOX Fuel Fabrication Facility (MFFF) is scheduled to start up at Zheleznogorsk by 2014, operated by the Mining & Chemical Combine (MCC). It will make MOX granules and pelletised MOX for 400 fuel assemblies per year for the BN-800 and future fast reactors.
Burning the plutonium in the BN-600 reactor was to commence in 2012, with the breeding blanket of depleted uranium removed and replaced by stainless steel reflector assemblies. The BN-800 reactor now under construction will have a uranium blanket but will operated as a net plutonium consumer for the life of the disposition project. Jointly they are expected to burn 1.5 tonnes of this weapons plutonium per year. The USA and Russia intend to continue cooperative development of a gas-cooled high-temperature reactor (GT-MHR) in Russia "which may create additional possibilities for speeding up plutonium disposition" from about 2015.
The 2000 US-Russian agreement precludes the reprocessing of MOX fuel using military plutonium if the plutonium is separated out, so such reprocessing will be either to give plutonium plus uranium or plus actinides. Russia is said to have 40 tonnes of separated reactor-grade plutonium already from reprocessed fuel. 

Thorium-plutonium fuel

Since the early 1990s Russia has had a program to develop a thorium-uranium fuel, which more recently has moved to have a particular emphasis on utilisation of weapons-grade plutonium in a thorium-plutonium fuel.
The program is based at Moscow's Kurchatov Institute and involves the US company Lightbridge Corporation (formerly Thorium Power) and US government funding to design fuel for Russian VVER-1000 reactors. Whereas normal fuel uses enriched uranium oxide throughout the fuel assembly, the new design has a demountable centre portion and blanket arrangement, with the plutonium in the centre and the thorium (with uranium) around it*. The Th-232 becomes U-233, which is fissile – as is the core Pu-239. Blanket material remains in the reactor for 9 years but the centre portion is burned for only three years (as in a normal VVER). Design of the seed fuel rods in the centre portion draws on extensive experience of Russian navy reactors.
* More precisely: A normal VVER-1000 fuel assembly has 331 rods each 9 mm diameter forming a hexagonal assembly 235 mm wide. Here, the centre portion of each assembly is 155 mm across and holds the seed material consisting of metallic Pu-Zr alloy (Pu is about 10% of alloy, and isotopically over 90% Pu-239) as 108 twisted tricorn-section rods 12.75 mm across with Zr-1%Nb cladding. The sub-critical blanket consists of U-Th oxide fuel pellets (1:9 U:Th, the U enriched up to almost 20%) in 228 Zr-1%Nb cladding tubes 8.4 mm diameter - four layers around the centre portion. The blanket material achieves 100 GWd/t burn-up. Together as one fuel assembly the seed and blanket have the same geometry as a normal VVER-100 fuel assembly.
The thorium-plutonium fuel claims four advantages over MOX: proliferation resistance, compatibility with existing reactors – which will need minimal modification to be able to burn it, and the fuel can be made in existing plants in Russia – hence it could be used from 2006. In addition, a lot more plutonium can be put into a single fuel assembly than with MOX, so that three times as much can be disposed of as when using MOX. The spent fuel amounts to about half the volume of MOX and is even less likely to allow recovery of weapons-useable material than spent MOX fuel, since less fissile plutonium remains in it. With an estimated 150 tonnes of weapons Pu in Russia, the thorium-plutonium project would not necessarily cut across existing plans to make MOX fuel.
See also Information Papers: Plutonium, Mixed Oxide Fuel and Synroc.
Sources:
Ivanov, 2000, paper in Proceedings of 25th UI Symposium.
NATO ASI series, 1994, Managing the Plutonium Surplus: Applications and Technical Options.
Underhill, D H, 1998, paper to U'98 Conference, Adelaide.
USEC, Megatons to Megawatts Program, Status Report 2001.
Thorium Power 2003, Weapons-grade Plutonium Burning Fuel for Russian VVER-1000 Nuclear Power Plants.
Morozov et al 2005, Thorium fuel as a superior approach to disposing of excess weapons-grade plutonium in Russian VVER-1000 reactors. Nuclear Future?

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