Uranium Enrichment

http://www.world-nuclear.org/information-library/nuclear-fuel-cycle/conversion-enrichment-and-fabrication/uranium-enrichment.aspx

(Updated February 2019)

Most of the 500 commercial nuclear power reactors operating or under construction in the world today require uranium 'enriched' in the U-235 isotope for their fuel.
The commercial process employed for this enrichment involves gaseous uranium in centrifuges. An Australian process based on laser excitation is under development.
Prior to enrichment, uranium oxide must be converted to a fluoride so that it can be processed as a gas, at low temperature.
From a non-proliferation standpoint, uranium enrichment is a sensitive technology needing to be subject to tight international control.
There is a significant surplus of world enrichment capacity.

Uranium found in nature consists largely of two isotopes, U-235 and U-238. The production of energy in nuclear reactors is from the 'fission' or splitting of the U-235 atoms, a process which releases energy in the form of heat. U-235 is the main fissile isotope of uranium.

Natural uranium contains 0.7% of the U-235 isotope. The remaining 99.3% is mostly the U-238 isotope which does not contribute directly to the fission process (though it does so indirectly by the formation of fissile isotopes of plutonium). Isotope separation is a physical process to concentrate (‘enrich’) one isotope relative to others. Most reactors are light water reactors (of two types – PWR and BWR) and require uranium to be enriched from 0.7% to 3-5% U-235 in their fuel. This is normal low-enriched uranium (LEU). There is some interest in taking enrichment levels to about 7%, and even close to 20% for certain special power reactor fuels, as high-assay LEU (HALEU).

Uranium-235 and U-238 are chemically identical, but differ in their physical properties, notably their mass. The nucleus of the U-235 atom contains 92 protons and 143 neutrons, giving an atomic mass of 235 units. The U-238 nucleus also has 92 protons but has 146 neutrons – three more than U-235 – and therefore has a mass of 238 units.

The difference in mass between U-235 and U-238 allows the isotopes to be separated and makes it possible to increase or "enrich" the percentage of U-235. All present and historic enrichment processes, directly or indirectly, make use of this small mass difference.

Some reactors, for example the Canadian-designed Candu and the British Magnox reactors, use natural uranium as their fuel. (For comparison, uranium used for nuclear weapons would have to be enriched in plants specially designed to produce at least 90% U-235.)

Enrichment processes require uranium to be in a gaseous form at relatively low temperature, hence uranium oxide from the mine is converted to uranium hexafluoride in a preliminary process, at a separate conversion plant.

There is significant over-supply of enrichment capacity worldwide, much of which has been used to diminish uranium demand or supplement uranium supply. The ability of enrichment to substitute for uranium (see description of underfeeding below) has become more significant as centrifuge technology has taken over, since this means both lower SWU costs and the need to keep the centrifuges running, so capacity remains on line even as demand drops away.

Although 13 countries have enrichment production capability or near-capability, about 90% of world enrichment capacity is in the five nuclear weapons states. These plus Germany, Netherlands and Japan provide toll enrichment services to the commercial market.

International Enrichment Centres, Multilateral approaches

Following proposals from the International Atomic Energy Agency (IAEA) and Russia, and in connection with the US-led Global Nuclear Energy Partnership (GNEP), there are moves to establish international uranium enrichment centres. These are one kind of multilateral nuclear approaches (MNA) called for by IAEA. Part of the motivation for international centres is to bring all new enrichment capacity, and perhaps eventually all enrichment, under international control as a non-proliferation measure. Precisely what "control" means remains to be defined, and will not be uniform in all situations. But having ownership and operation shared among a number of countries at least means that there is a level of international scrutiny which is unlikely in a strictly government-owned and -operated national facility.

The first of these international centres is the International Uranium Enrichment Centre (IUEC) at Angarsk in Siberia, with Kazakh, Armenian and Ukrainian equity so far. The centre is to provide assured supplies of low-enriched uranium for power reactors to new nuclear power states and those with small nuclear programs, giving them equity in the project, but without allowing them access to the enrichment technology. Russia will maintain majority ownership, and in February 2007 the IUEC was entered into the list of Russian nuclear facilities eligible for implementation of IAEA safeguards. The USA has expressed support for the IUEC at Angarsk. IUEC will sell both enrichment services (SWU) and enriched uranium product.

In some respects this is very similar to what pertained with the Eurodif set-up, where a single large enrichment plant in France with five owners (France – 60%, Italy, Spain, Belgium and Iran) was operated under IAEA safeguards by the host country without giving participants any access to the technology – simply some entitlement to share of the product, and even that was constrained in the case of Iran. The French Atomic Energy Commission proposed that the new Georges Besse II plant which replaced Eurodif should be open to international partnerships on a similar basis, and minor shares in the Areva subsidiary operating company Societe d'Enrichissement du Tricastin (SET) have so far been sold to GDF Suez (now Engie), a Japanese partnership, and Korea Hydro and Nuclear Power (KHNP) – total 12%.

The three-nation Urenco set-up is also similar though with more plants in different countries – UK, Netherlands and Germany – and here the technology is not available to host countries or accessible to other equity holders. Like Russia with IUEC, Urenco (owned by the UK and Netherlands host governments plus E.On and RWE in Germany) has made it plain that if its technology is used in international centres it will not be accessible. Its new plant is in the USA, where the host government has regulatory but not management control.

A planned new Areva plant in the USA has no ownership diversity beyond that of Areva itself, so will be essentially a French plant on US territory. The only other major enrichment plant in the Western world is USEC's very old one, in the USA.

The Global Laser Enrichment project which may proceed to build a commercial plant in the USA has shareholding from companies based in three countries: USA (51%), Canada (24%) and Japan (25%).

Enrichment Processes

A number of enrichment processes have been demonstrated historically or in the laboratory but only two, the gaseous diffusion process and the centrifuge process, have operated on a commercial scale. In both of these, UF₆ gas is used as the feed material. Molecules of UF₆ with U-235 atoms are about one percent lighter than the rest, and this difference in mass is the basis of both processes. Isotope separation is a physical process.*

*One chemical process has been demonstrated to pilot plant stage but not used. The French Chemex process exploited a very slight difference in the two isotopes' propensity to change valency in oxidation/reduction, utilising aqueous (III valency) and organic (IV) phases.

Large commercial enrichment plants are in operation in France, Germany, Netherlands, UK, USA, and Russia, with smaller plants elsewhere. New centrifuge plants are being built in France and USA. Several plants are adding capacity. China’s capacity is expanding considerably, in line with domestic requirements. With surplus capacity, Russian plants operate at low tails assays (underfeeding) to produce low-enriched uranium for sale.

World enrichment capacity – operational and planned (thousand SWU/yr)

Country	Company and plant	2013	2015	2020
France	Areva, Georges Besse I & II	5500	7000	7500
Germany-Netherlands-UK	Urenco: Gronau, Germany; Almelo, Netherlands; Capenhurst, UK.	14,200	14,400	14,900
Japan	JNFL, Rokkaasho	75	75	75
USA	USEC, Piketon	0*	0	0
USA	Urenco, New Mexico	3500	4700	4700
USA	Areva, Idaho Falls	0	0	0
USA	Global Laser Enrichment, Paducah	0	0	0
Russia	Tenex: Angarsk, Novouralsk, Zelenogorsk, Seversk	26,000	26,578	28,663
China	CNNC, Hanzhun & Lanzhou	2200	5760	10,700+
Other	Various: Argentina, Brazil, India, Pakistan, Iran	75	100	170
	Total SWU/yr approx	51,550	58,600	66,700
	Requirements (WNA reference scenario)	49,154	47,285	57,456

Source: World Nuclear Association Nuclear Fuel Report 2013 & 2105, information paper on China's Nuclear Fuel Cycle, Areva 2014 Reference Document for most 2013 figures.

* Diffusion, closed mid-2013, US centrifuge proposed.
'Other' includes Resende in Brazil, Rattehallib in India and Natanz in Iran. At end of 2012 Iran had about 9000 SWU/yr capacity operating, according to ISIS and other estimates.
The Euratom Supply Agency Annual Report 2014 estimated world nameplate capacity at 56 million SWU, Russia 28 million SWU, Urenco 18.1 million SWU and Areva 7.5 million SWU.

The feedstock for enrichment consists of uranium hexafluoride (UF₆) from the conversion plant. Following enrichment two streams of UF₆ are formed: the enriched ‘product’ containing a higher concentration of U-235 which will be used to make nuclear fuel, and the ‘tails’ containing a lower concentration of U-235, and known as depleted uranium (DU). The tails assay (concentration of U-235) is an important quantity since it indirectly determines the amount of work that needs to be done on a particular quantity of uranium in order to produce a given product assay. Feedstock may have a varying concentration of U-235, depending on the source. Natural uranium will have a U-235 concentration of approximately 0.7%, while recycled uranium will be around 1% and tails for re-enrichment often around 0.25-0.30%.

The capacity of enrichment plants is measured in terms of 'separative work units' or SWU. The SWU is a complex unit which indicates the energy input relative to the amount of uranium processed, the degree to which it is enriched (i.e. the extent of increase in the concentration of the U-235 isotope relative to the remainder) and the level of depletion of the remainder – called the ‘tails’. The unit is strictly: kilogram separative work unit, and it measures the quantity of separative work performed to enrich a given amount of uranium a certain amount when feed and product quantities are expressed in kilograms. The unit 'tonnes SWU' is also used.

Percentage Variation in Uranium Requirements and Energy Input to Enrichment with Different Tails Assay, from a Base of 0.22%25 U-235 line graph

For instance, to produce one kilogram of uranium enriched to 5% U-235 requires 7.9 SWU if the plant is operated at a tails assay 0.25%, or 8.9 SWU if the tails assay is 0.20% (thereby requiring only 9.4 kg instead of 10.4 kg of natural U feed). There is always a trade-off between the cost of enrichment SWU and the cost of uranium.

Today, 5% U-235 is the maximum level of enrichment for fuel used in normal power reactors. However, especially in relation to new small reactor designs, there is increasing interest in higher enrichment levels. Some small demand already exists for research reactors. High-assay LEU (HALEU) may be 10-20% U-235 for certain special power reactor fuels. However such HALEU is best produced onsite where it is converted from fluoride and made into fuel, to avoid the need for special transport packages for HALEU UF₆.

The first graph shows enrichment effort (SWU) per unit of product. The second shows how one tonne of natural uranium feed might end up: as 120-130 kg of uranium for power reactor fuel, as 26 kg of typical research reactor fuel, or conceivably as 5.6 kg of weapons-grade material. The curve flattens out so much because the mass of material being enriched progressively diminishes to these amounts, from the original one tonne, so requires less effort relative to what has already been applied to progress a lot further in percentage enrichment. The relatively small increment of effort needed to achieve the increase from normal levels is the reason why enrichment plants are considered a sensitive technology in relation to preventing weapons proliferation, and are very tightly supervised under international agreements. Where this safeguards supervision is compromised or obstructed, as in Iran, concerns arise.About 140,000 SWU is required to enrich the annual fuel loading for a typical 1000 MWe light water reactor at today's higher enrichment levels. Enrichment costs are substantially related to electrical energy used. The gaseous diffusion process consumes about 2500 kWh (9000 MJ) per SWU, while modern gas centrifuge plants require only about 50 kWh (180 MJ) per SWU.

Enrichment accounts for almost half of the cost of nuclear fuel and about 5% of the total cost of the electricity generated. In the past it has also accounted for the main greenhouse gas impact from the nuclear fuel cycle where the electricity used for enrichment is generated from coal. However, it still only amounts to 0.1% of the carbon dioxide from equivalent coal-fired electricity generation if modern gas centrifuge plants are used.

The utilities which buy uranium from the mines need a fixed quantity of enriched uranium in order to fabricate the fuel to be loaded into their reactors. The quantity of uranium they must supply to the enrichment company is determined by the enrichment level required (% U-235) and the tails assay (also % U-235). This is the contracted or transactional tails assay, and determines how much natural uranium must be supplied to create a quantity of Enriched Uranium Product (EUP) – a lower tails assay means that more enrichment services (notably energy) are to be applied. The enricher, however, has some flexibility in respect to the operational tails assay at the plant. If the operational tails assay is lower than the contracted/transactional assay, the enricher can set aside some surplus natural uranium, which it is free to sell (either as natural uranium or as EUP) on its own account.

This is known as underfeeding.* Utilities are increasingly seeking to control this flexibility of operation in contracts, and themselves get some of the benefit from underfeeding.

* The opposite situation, where the operational tails assay is higher, requires the enricher to supplement the natural uranium supplied by the utility with some of its own – this is called overfeeding.

In respect to underfeeding (or overfeeding), the enricher will base its decision on the plant economics together with uranium and energy prices. The World Nuclear Association's 2015 Nuclear Fuel Report estimates that by underfeeding the enrichers have the potential to contribute 5700 to 8000 tU/yr to world markets to 2025 on the basis of typical Western 0.22% tails assay, much of this potential in Russia, where tails assays are normally 0.10% U-235.

With reduced demand for enriched uranium following the Fukushima accident, enrichment plants have continued running, since it is costly to shut down and re-start centrifuges. The surplus SWU output can be sold, or the plants can be underfed so that the enricher ends up with excess uranium for sale, or with enriched product for its own inventory and later sale. The inertia of the enrichment process thus exacerbates over-supply in the uranium market and depresses SWU prices (from $160/SWU in 2010, the spot price in March 2016 was $60). With forecast overcapacity, it is likely that some older cascades will be retired.

Obsolete diffusion plants have been retired, the last being some belated activity at Paducah in 2013.

Natural uranium is usually shipped to enrichment plants in type 48Y cylinders, each holding about 12.5 tonnes of uranium hexafluoride (8.4 tU). These cylinders are then used for long-term storage of DU, typically at the enrichment site. Enriched uranium is shipped in type 30B cylinders, each holding 2.27 t UF6 (1.54 tU).

Supply source:	2000	2010	2015	projected 2020
Diffusion	50%	25%	0	0
Centrifuge	40%	65%	100%	93%
Laser	0	0	0	3%
HEU ex weapons	10%	10%	0	4%

The three enrichment processes described below have different characteristics. Diffusion is flexible in response to demand variations and power costs but is very energy-intensive. With centrifuge technology it is easy to add capacity with modular expansion, but it is inflexible and best run at full capacity with low operating cost. Laser technology can strip down to very low level tails assay, and is also capable of modular plant expansion.

Centrifuge process

The gas centrifuge process was first demonstrated in the 1940s but was shelved in favour of the simpler diffusion process. It was then developed and brought on stream in the 1960s as the second-generation enrichment technology. It is economic on a smaller scale, e.g. under 2 million SWU/yr, which enables staged development of larger plants. It is much more energy efficient than diffusion, requiring only about 40-50 kWh per SWU.

The centrifuge process has been deployed at a commercial level in Russia, and in Europe by Urenco, an industrial group formed by British, German and Dutch governments. Russia's four plants at Seversk, Zelenogorsk, Angarsk and Novouralsk account for some 40% of world capacity*. Urenco operates enrichment plants in the UK, Netherlands and Germany and one in the USA.

* In 2012 Russia was commissioning 8th generation centrifuges with service life of up to 30 years. The last 6th & 7th generation ones were installed in 2005, and 8th generation equipment has been supplied since 2004 to replace 5th generation models with a service life of only 15 years.

In Japan, JNC and JNFL operate small centrifuge plants, the capacity of JNFL's at Rokkasho was planned to be 1.5 million SWU/yr. China has two small centrifuge plants imported from Russia. China has several centrifuge plants, the first at Hanzhun with 6th generation centrifuges imported from Russia. The Lanzhou plant is operating at 3.5 million SWU/yr but expanding to 6.5 million SWU/yr by 2020, and Hanzhun is operating at 2.2 million SWU/yr. Others are under construction. Brazil has a small plant which is being developed to 0.2 million SWU/yr. Pakistan has developed centrifuge enrichment technology, and this appears to have been sold to North Korea. Iran has sophisticated centrifuge technology which is operational, with estimated 9000 SWU/yr capacity.

In both France and the USA plants with late-generation Urenco centrifuge technology have been built to replace ageing diffusion plants, not least because they are more economical to operate. As noted, a centrifuge plant requires as little as 40 kWh/SWU power (Urenco at Capenhurst, UK, input 62.3 kWh/SWU for the whole plant in 2001-02, including infrastructure and capital works).

Areva's new €3 billion French plant – Georges Besse II – started commercial operation in April 2011 and reached full capacity of 7.5 million SWU/yr in 2016. As noted above, customers hold more than 10% equity in Areva’s operating subsidiary SET.

Urenco's new $1.5 billion National Enrichment Facility in New Mexico, USA commenced production in June 2010. Full initial capacity of 3.7 million SWU/yr was reached in 2014, and capacity reached 4.7 million SWU/yr in 2015 – enough for 10% of US electricity needs.

Following this, Areva was planning to build the $2 billion, 3.3 million SWU/yr Eagle Rock plant at Idaho Falls, USA. In 2009 it applied for doubling in capacity to 6.6 million SWU/yr. It is now cancelled, and in 2018 Orano requested the NRC to terminate the licence.

USEC, now Centrus, was building its American Centrifuge Plant in Piketon, Ohio, on the same Portsmouth site where the DOE's experimental plant operated in the 1980s as the culmination of a very major R&D programme. Operation from 2012 was envisaged, at a cost of $3.5 billion then estimated. It was designed to have an initial annual capacity of 3.8 million SWU. Authorisation for enrichment up to 10% was sought – most enrichment plants operate up to 5% U-235 product, which is becoming a serious constraint as reactor fuel burnup increases. A demonstration cascade started up in September 2007 with about 20 prototype machines, and a lead cascade of commercial centrifuges started operation in March 2010. These are very large machines, 13 m tall, each with about 350 SWU/yr capacity and requiring regular maintenance. However the whole project was largely halted in July 2009 pending further finance, although a demonstration cascade became operational in October 2013 as "the centerpiece of the RD&D program with DOE." It was licensed for 7 million SWU/yr enrichment up to 10% U-235, but operations ceased in February 2016.

A bank of centrifuges at a Urenco plant

Like the diffusion process, the centrifuge process uses UF₆ gas as its feed and makes use of the slight difference in mass between U-235 and U-238. The gas is fed into a series of vacuum tubes, each containing a rotor 3 to 5 metres tall and 20 cm diameter.* European centrifuges produce 40-100 SWU/yr. When the rotors are spun rapidly, at 50,000 to 70,000 rpm, the heavier molecules with U-238 increase in concentration towards the cylinder's outer edge. There is a corresponding increase in concentration of U-235 molecules near the centre. The countercurrent flow set up by a thermal gradient enables enriched product to be drawn off axially, heavier molecules at one end and lighter ones at the other.

* USEC's American Centrifuges are more than 12 m tall and 40-50 cm diameter. The Russian centrifuges are less than one metre tall. Chinese ones are larger, but shorter than Urenco's.

The enriched gas forms part of the feed for the next stages while the depleted UF6 gas goes back to the previous stage. Eventually enriched and depleted uranium are drawn from the cascade at the desired assays.

To obtain efficient separation of the two isotopes, centrifuges rotate at very high speeds, with the outer wall of the spinning cylinder moving at between 400 and 500 metres per second to give a million times the acceleration of gravity.

Although the volume capacity of a single centrifuge is much smaller than that of a single diffusion stage, its capability to separate isotopes is much greater. Centrifuge stages normally consist of a large number of centrifuges in parallel. Such stages are then arranged in cascade similarly to those for diffusion. In the centrifuge process, however, the number of stages may only be 10 to 20, instead of a thousand or more for diffusion. Centrifuges are designed to run for about 25 years continuously, and cannot simply be slowed or shut down and restarted according to demand. Western cascades are designed for 0.18 to 0.22% tails assay, Russian ones for 0.10%.

Laser processes

Laser enrichment processes have been the focus of interest for some time. They are a possible third-generation technology promising lower energy inputs, lower capital costs and lower tails assays, hence significant economic advantages. One of these processes is almost ready for commercial use. Laser processes are in two categories: atomic and molecular.

Development of the Atomic Vapour Laser Isotope Separation (AVLIS, and the French SILVA) began in the 1970s. In 1985 the US Government backed it as the new technology to replace its gaseous diffusion plants as they reached the end of their economic lives early in the 21st century. However, after some US$ 2 billion in R&D, it was abandoned in USA in favour of SILEX, a molecular process. French work on SILVA ceased following a 4-year program to 2003 to prove the scientific and technical feasibility of the process. Some 200kg of 2.5% enriched uranium was produced in this.

Atomic vapour processes work on the principle of photo-ionisation, whereby a powerful laser is used to ionise particular atoms present in a vapour of uranium metal. (An electron can be ejected from an atom by light of a certain frequency. The laser techniques for uranium use frequencies which are tuned to ionise a U-235 atom but not a U-238 atom.) The positively-charged U-235 ions are then attracted to a negatively-charged plate and collected. Atomic laser techniques may also separate plutonium isotopes.

Most molecular processes which have been researched work on a principle of photo-dissociation of UF₆ to solid UF₅⁺, using tuned laser radiation as above to break the molecular bond holding one of the six fluorine atoms to a U-235 atom. This then enables the ionized UF₅ to be separated from the unaffected UF₆ molecules containing U-238 atoms, hence achieving a separation of isotopes.* Any process using UF₆ fits more readily within the conventional fuel cycle than the atomic process.

* A similar principle can be used in enriching atomic lithium, with magnetic separation of the ionised atoms, leaving pure Li-7.

The main molecular laser process to enrich uranium is SILEX, which utilises UF₆ and is now known as Global Laser Enrichment (GLE). In 2006 GE Energy entered a partnership with Australia's Silex Systems to develop the third-generation SILEX process. It provided for GE (now GE-Hitachi) to construct in the USA an engineering-scale test loop, then a pilot plant or lead cascade, which could be operating in 2012, and expanded to a full commercial plant. Apart from US$ 20 million upfront and subsequent payments, the license agreement would yield 7-12% royalties, the precise amount depending on the cost of deploying the commercial technology. In mid-2008 Cameco bought into the GLE project, paying $124 million for 24% share, alongside GE (51%) and Hitachi (25%). (Earlier, in 1996 USEC had secured the rights to evaluate and develop SILEX for uranium but baled out of the project in 2003.)

In April 2016 GE and Hitachi notified their intention to exit GLE, and during subsequent negotiations Silex funded 76% of GLE’s R&D at Wilmington, North Carolina. GLE is well advanced in commercialising the SILEX process, and has an agreement with the US Department of Energy to enrich about 300,000 tonnes of depleted uranium tails at Paducah, Kentucky to natural-grade uranium.

In February 2019 Silex Systems and Cameco agreed to buy out the GEH 76% share in GLE for US$ 20 million on a deferred payment basis, so that Cameco holds 49% of GLE and Silex 51%. Cameco has an option to purchase an additional 26% of GLE. The agreement calls for Silex and Cameco to pay $300,000 per month to complete construction of the prototype enrichment facility, known as Wilmington Test Loop, which has been partially built by GEH. The agreement is contingent upon US government approvals. Silex said: "The Paducah commercial opportunity respresents an ideal path to market for our disruptive SILEX laser enrichment technology."

GE had earlier referred to SILEX, which it rebadged as GLE, as "game-changing technology" with a "very high likelihood" of success. GLE is completing the test loop program, the initial phase of which has already been successful in meeting performance criteria, and engineering design for a commercial facility has commenced. GLE is operating the test loop at Global Nuclear Fuel's Wilmington, North Carolina fuel fabrication facility – GNF is a partnership of GE, Toshiba, and Hitachi.

In October 2007 the two largest US nuclear utilities, Exelon and Entergy, signed letters of intent to contract for uranium enrichment services from GE Hitachi Global Laser Enrichment LLC (GLE). The utilities may also provide GLE with support if needed for development of a commercial-scale GLE plant. In August 2010 TVA agreed to buy $400 million of enrichment services from GLE if the project proceeds.

In mid-2009 GEH submitted the last part of its licence application for this GLE plant at Wilmington, and following review of provisions for the physical protection of special nuclear material and classified matter, material control and accounting, plus further review by the NRC Atomic Safety and Licensing Board, a full licence to construct and operate a plant of up to 6 million SWU/yr was issued in September 2012. GLE will now decide in the light of commercial considerations on whether to proceed with a full-scale enrichment facility at Wilmington. The project, enriching up to 8% U-235, could be operational in a few years, and ramp up to capacity fairly quickly.

In August 2013 GLE submitted a proposal to the DOE to establish a “$1 billion” laser enrichment plant at Paducah, Kentucky to enrich high-assay tails (above 0.34% U-235) owned by DOE to natural uranium level (0.7% U-235). There is about 115,000 tonnes of these at Paducah and Portsmouth (among a total of 550,000 t tails). In November 2013 the DOE announced that it would proceed with contract negotiations to this end. In January 2014 GLE told the NRC that though negotiations with the DOE continued, it expected to apply for a licence later that year to build and operate the Paducah Laser Enrichment Facility (PLEF) which would enrich the tails over about 40 years to natural grade, for sale. GLE expects licensing to take 2-3 years. Negotiations with the DOE continued into 2016, and in November an agreement was signed with the DOE for it to supply about 300,000 tonnes of high-assay tails, justifying construction by GLE of the plant in the early 2020s. PLEF would become a commercial uranium enrichment production facility under a US NRC licence, producing about 100,000 tonnes of natural-grade uranium over 40 years or more. The DOE would dispose of the reduced-assay balance. The estimated plant size is 0.5 to 1.0 million SWU/yr, since purchases of DU may not exceed 2000 t/yr natural uranium equivalent.

Applications to silicon and zirconium stable isotopes are also being developed by Silex Systems near Sydney.

CRISLA is another molecular laser isotope separation process which is the early stages of development. In this a gas is irradiated with a laser at a particular wavelength that would excite only the U-235 isotope. The entire gas is subjected to low temperatures sufficient to cause condensation on a cold surface or coagulation in the un-ionised gas. The excited molecules in the gas are not as likely to condense as the unexcited molecules. Hence in cold-wall condensation, gas drawn out of the system is enriched in the U-235 isotope that was laser-excited. NeuTrek, the development company, is aiming to build a pilot plant in USA.

Gaseous diffusion process

The energy-intensive gaseous diffusion process of uranium enrichment is no longer used in the nuclear industry. It involves forcing uranium hexafluoride gas under pressure through a series of porous membranes or diaphragms. As U-235 molecules are lighter than the U-238 molecules they move faster and have a slightly better chance of passing through the pores in the membrane. The UF₆ which diffuses through the membrane is thus slightly enriched, while the gas which did not pass through is depleted in U-235.

This process is repeated many times in a series of diffusion stages called a cascade. Each stage consists of a compressor, a diffuser and a heat exchanger to remove the heat of compression. The enriched UF₆ product is withdrawn from one end of the cascade and the depleted UF₆ is removed at the other end. The gas must be processed through some 1400 stages to obtain a product with a concentration of 3-4% U-235. Diffusion plants typically have a small amount of separation through one stage (hence the large number of stages) but are capable of handling large volumes of gas.

Commercial uranium enrichment was first carried out by the diffusion process in the USA, at Oak Ridge, Tennessee. The process was also used in Russia, UK, France, China and Argentina as well, but only on a significant scale in the USA and France in recent years. Russia phased out the process in 1992 and the last diffusion plant was USEC's Paducah facility, which shut down in mid-2013. It is very energy-intensive, requiring about 2400 kWh per SWU*. USEC said that electricity accounted for 70% of the production cost at Paducah, which was the last of three large plants in the USA originally developed for weapons programs and had a capacity of some 8 million SWU per year. It was used to enrich some high-assay tails before being finally shut down after 60 years' operation. At Tricastin, in southern France, a more modern diffusion plant with a capacity of 10.8 million kg SWU per year had been operating since 1979 (see photo below). This Georges Besse I plant could produce enough 3.7% enriched uranium a year to fuel some ninety 1000 MWe nuclear reactors. It was shut down in mid-2012, after 33 years' continuous operation. Its replacement (GB II, a centrifuge plant – see above) has commenced operation.

* It has been estimated that 7% of total US electricity demand was from enrichment plants at the height of the Cold War, when 90% U-235 was required, rather than the reactor grades of 3-4%for power generation.

In recent years the gaseous diffusion process had accounted for about 25% of world enrichment capacity. However, though they have proved durable and reliable, gaseous diffusion plants reached the end of their design life and the much more energy-efficient centrifuge enrichment technology has replaced them.

The large Georges Besse I enrichment plant at Tricastin in France (beyond cooling towers) was shut down in 2012.
Most of the output from the nuclear power plant (4x915MWe net) was used to power the enrichment facility.

Electromagnetic process

A very early endeavour was the electromagnetic isotope separation (EMIS) process using calutrons. This was developed in the early 1940s in the Manhattan Project to make the highly enriched uranium used in the Hiroshima bomb, but was abandoned soon afterwards. However, it reappeared as the main thrust of Iraq's clandestine uranium enrichment program for weapons discovered in 1992. EMIS uses the same principles as a mass spectrometer (albeit on a much larger scale). Ions of uranium-238 and uranium-235 are separated because they describe arcs of different radii when they move through a magnetic field. The process is very energy-intensive – about ten times that of diffusion.

Aerodynamic processes

Two aerodynamic processes were brought to demonstration stage around the 1970s. One is the jet nozzle process, with demonstration plant built in Brazil, and the other the Helikon vortex tube process developed in South Africa. Neither is in use now, though the latter is the forerunner of new R&D. They depend on a high-speed gas stream bearing the UF6 being made to turn through a very small radius, causing a pressure gradient similar to that in a centrifuge. The light fraction can be extracted towards the centre and the heavy fraction on the outside. Thousands of stages are required to produce enriched product for a reactor. Both processes are energy-intensive - over 3000 kWh/SWU. The Helikon Z-plant in the early 1980s was not commercially oriented and had less than 500,000 SWU/yr capacity. It required some 10,000 kWh/SWU.

The Aerodynamic Separation Process (ASP) being developed by Klydon in South Africa employs similar stationary-wall centrifuges with UF6 injected tangentially. It is based on Helikon but pending regulatory authorisation it has not yet been tested on UF6 - only light isotopes such as silicon. However, extrapolating from results there it is expected to have an enrichment factor in each unit of 1.10 (cf 1.03 in Helikon) with about 500 kWh/SWU and development of it is aiming for 1.15 enrichment factor and less than 500 kWh/SWU. Projections give an enrichment cost under $100/SWU, with this split evenly among capital, operation and energy input.

One chemical process has been demonstrated to pilot plant stage but not used. The French Chemex process exploited a very slight difference in the two isotopes' propensity to change valency in oxidation/reduction, utilising aqueous (III valency) and organic (IV) phases.

Enrichment of reprocessed uranium

In some countries used fuel is reprocessed to recover its uranium and plutonium, and to reduce the final volume of high-level wastes. The plutonium is normally recycled promptly into mixed-oxide (MOX) fuel, by mixing it with depleted uranium.

Where uranium recovered from reprocessing used nuclear fuel (RepU) is to be re-used, it needs to be converted and re-enriched. This is complicated by the presence of impurities and two new isotopes in particular: U-232 and U-236, which are formed by or following neutron capture in the reactor, and increase with higher burn-up levels. U-232 is largely a decay product of Pu-236, and increases with storage time in used fuel, peaking at about ten years. Both decay much more rapidly than U-235 and U-238, and one of the daughter products of U-232 emits very strong gamma radiation, which means that shielding is necessary in any plant handling material with more than very small traces of it. U-236 is a neutron absorber which impedes the chain reaction, and means that a higher level of U-235 enrichment is required in the product to compensate. 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. Being lighter, both isotopes tend to concentrate in the enriched (rather than depleted) output, so reprocessed uranium which is re-enriched for fuel must be segregated from enriched fresh uranium. The presence of U-236 in particular means that most reprocessed uranium can be recycled only once - the main exception being in the UK with AGR fuel made from recycled Magnox uranium being reprocessed. U-234 is also present in RepU, but as an alpha emitter it does not pose extra problems. Traces of some fission products such as Tc-99 may also carry over.

All these considerations mean that only RepU from low-enriched, low-burnup used fuel is normally recycled directly through an enrichment plant. For instance, some 16,000 tonnes of RepU from Magnox reactors* in UK has been used to make about 1650 tonnes of enriched AGR fuel, via two enrichment plants. Much smaller quantities have been used elsewhere, in France and Japan. Some re-enrichment, eg for Swiss, German and Russian fuel, is actually done by blending RepU with HEU.

* Since Magnox fuel was not enriched in the first place, this is actually known as Magnox depleted uranium (MDU). It assayed about 0.4% U-235 and 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 LEU. Recycling of MDU was discontinued in 1996 due to economic factors.

A laser process would theoretically be ideal for enriching RepU as it would ignore all but the desired U-235, but this remains to be demonstrated with reprocessed feed.

Tails from enriching reprocessed uranium remain the property of the enricher. Some recycled uranium has been enriched by Tenex at Seversk for Areva, under a 1991 ten-year contract covering about 500 tonnes UF₆. French media reports in 2009 alleging that wastes from French nuclear power plants were stored at Seversk evidently refer to tails from this.

Enrichment of depleted uranium tails

Early enrichment activities often left depleted uranium tails with about 0.30% U-235, and there were tens of thousands of tonnes of these sitting around as the property of the enrichment companies. With the wind-down of military enrichment, particularly in Russia, there was a lot of spare capacity unused. Consequently, since the mid 1990s some of the highest-assay tails have been sent to Russia by Areva and Urenco for re-enrichment by Tenex. These arrangements however cease in 2010, though Tenex may continue to re-enrich Russian tails. Tenex now owns all the tails from that secondary re-enrichment, and they are said to comprise only about 0.10% U-235.

After enrichment

The enriched UF₆ is converted to UO₂ and made into fuel pellets – ultimately a sintered ceramic, which are encased in metal tubes to form fuel rods, typically up to four metres long. A number of fuel rods make up a fuel assembly, which is ready to be loaded into the nuclear reactor. See Fuel Fabrication paper.

Environmental Issues

With the minor exception of reprocessed uranium, enrichment involves only natural, long-lived radioactive materials; there is no formation of fission products or irradiation of materials, as in a reactor. Feed, product, and depleted material are all in the form of UF₆, though the depleted uranium may be stored long-term as the more stable U₃O₈.

Uranium is only weakly radioactive, and its chemical toxicity – especially as UF₆ – is more significant than its radiological toxicity. The protective measures required for an enrichment plant are therefore similar to those taken by other chemical industries concerned with the production of fluorinated chemicals.

Uranium hexafluoride forms a very corrosive material (HF – hydrofluoric acid) when exposed to moisture, therefore any leakage is undesirable. Hence:

in almost all areas of a centrifuge plant the pressure of the UF₆ gas is maintained below atmospheric pressure and thus any leakage could only result in an inward flow;
double containment is provided for those few areas where higher pressures are required;
effluent and venting gases are collected and appropriately treated.

Notes & References

General sources

Heriot, I.D. (1988). Uranium Enrichment by Centrifuge, Report EUR 11486, Commission of the European Communities, Brussels.
Kehoe, R.B. (2002). The Enriching Troika, a History of Urenco to the Year 2000. Urenco, Marlow UK.
Wilson, P.D. (ed)(1996). The Nuclear Fuel Cycle – from ore to wastes. Oxford University Press, Oxford UK.
IAEA 2007, Management of Reprocessed Uranium – current status and future prospects, Tecdoc 1529.

Nuclear Fuel Cycle / Conversion Enrichment and Fabrication / Conversion and Deconversion

Conversion and Deconversion

(Updated January 2019)

http://www.world-nuclear.org/information-library/nuclear-fuel-cycle/conversion-enrichment-and-fabrication/conversion-and-deconversion.aspx

Uranium enrichment requires uranium as uranium hexafluoride, which is obtained from converting uranium oxide to UF₆.
Conversion plants are operating commercially in the USA, Canada, France, Russia and China.
Deconversion of depleted UF₆ to uranium oxide or UF₄ is undertaken for long-term storage of depleted uranium in more stable form.

Uranium leaves the mine as the concentrate of a stable oxide known as U₃O₈ or as a peroxide. It still contains some impurities and prior to enrichment has to be further refined before or after being converted to uranium hexafluoride (UF₆), commonly referred to as 'hex'. Both processes are normally included in the step between the mine and enrichment plant – referred to as 'conversion'.

Conversion plants are operating commercially in the USA, Canada, France, Russia and China. The main new plant is Areva’s Comurhex, operating between two sites in France. China’s capacity is expected to grow considerably through to 2025 and beyond to keep pace with domestic requirements.

World Primary Conversion Capacity

Company	Location	Nameplate capacity (tonnes U/yr as UF₆)	Approx capacity utilisation	Capacity utilisation tU/yr
Cameco	Port Hope, Canada	12,500	50%	6250
TVEL (Rosatom)	SGCE Seversk, Russia	18,000	100% assumed	18,000
Areva	Pierrelatte, France	15,000	70%	10,500
ConverDyn	Metropolis, USA	7000	100%	7000
CNNC	Lanzhou, China*	5000	80%	4000
IPEN	Brazil	100	70%	70
World Total		57,600		45,820

World Nuclear Association Nuclear Fuel Report 2017; World Nuclear Association information paper on China's Nuclear Fuel Cycle.

* Information on China's conversion capacity is uncertain. An additional 9000 t/yr plant is reported to be under construction at Lanzhou, as well as a 3000 tU/yr plant at Hengyang in Hunan.

Conversion process

The main, 'wet' process, is used by Cameco in Canada, by Areva in France, at Lanzhou in China and Seversk in Russia. For the wet process, the concentrate is first dissolved in nitric acid. The resulting clean solution of uranyl nitrate UO2(NO3)2.6H2O is fed into a countercurrent solvent extraction process, using tributyl phosphate dissolved in kerosene or dodecane. The uranium is collected by the organic extractant, from which it can be washed out by dilute nitric acid solution and then concentrated by evaporation. The solution is then calcined in a fluidised bed reactor to produce UO₃ (or UO₂ if heated sufficiently).

Alternatively, the uranyl nitrate may be concentrated and have ammonia injected to produce ammonium diuranate, which is then calcined to produce pure UO₃.

Crushed U₃O₈ from the dry process and purified uranium oxide UO₃ from the wet process are then reduced in a kiln by hydrogen to UO₂:

U₃O₈ + 2H₂ ===> 3UO₂ + 2H₂O ΔH = -109 kJ/mol

or UO₃ + H₂ ===> UO₂ + H₂O ΔH = -109 kJ/mol

This reduced oxide is then reacted in another kiln with gaseous hydrogen fluoride (HF) to form uranium tetrafluoride (UF4), though in some places this is made with aqueous HF by a wet process:

UO₂ + 4HF ===> UF₄ + 2H₂O ΔH = -176 kJ/mol

The tetrafluoride is then fed into a fluidised bed reactor or flame tower with gaseous fluorine to produce uranium hexafluoride, UF6. Hexafluoride ('hex') is condensed and stored.

UF₄ + F₂ ===> UF₆

Removal of impurities takes place at each step.

The alternative, 'dry' process is used in the USA. In the dry process, uranium oxide concentrates are first calcined (heated strongly) to drive off some impurities, then agglomerated and crushed. At Converdyn’s US conversion plant, U₃O₈ is first made into impure UF₆ and this is then refined in a two-stage distillation process.

UF₆, particularly if moist, is highly corrosive. When warm it is a gas, suitable for use in the enrichment process. At lower temperature and under moderate pressure, the UF₆ can be liquefied. The liquid is run into specially designed steel shipping cylinders which are thick walled and weigh over 15 tonnes when full. As it cools, the liquid UF₆ within the cylinder becomes a white crystalline solid and is shipped in this form.

The siting, environmental and security management of a conversion plant is subject to the regulations that are in effect for any chemical processing plant involving fluorine-based chemicals.

Secondary sources of conversion supply

Secondary supply of equivalent conversion services includes UF₆ material from commercial and government inventories, enricher underfeeding, and DU tails recovery. Uranium and plutonium recycle effectively adds to this. All these were estimated at 26,000 tU in 2013 but with the end of the Russian HEU supply to the USA, they are now much less – an estimated 10,000 tU in 2017. By 2030 they are predicted to be less than 9000 tU.

Depleted uranium and deconversion

Up to 90% of the original uranium feed ends up as depleted uranium (DU), which is stored long-term as UF₆ or preferably, after deconversion, as U₃O₈, allowing HF to be recycled. It may also be deconverted to UF₄, which is more stable, with much higher temperature of volatalisation. To early 2007, about one-quarter of the world's 1.5 million tonnes of DU had been deconverted.

The main deconversion plant is the 20,000 t/yr one run by Areva NC at Tricastin, France, and over 300,000 tonnes has been processed here. The technology has been sold to Russia. Two plants have been built by Uranium Disposition Services (UDS) at Portsmouth and Paducah, USA, with capacities of 13,500 and 18,000 t/yr respectively. A 6500 t/yr plant is being built at New Mexico in the USA by International Isotopes (INIS). In the UK, Urenco ChemPlants has built a 15,000 t/yr plant.

Uranium Deconversion Plants

Operator	Location	Capacity tU/yr
Areva	Tricastin, France	20,000
	Richland, Washington, USA	small
Urenco ChemPlants	Capenhurst, UK	15,000
Mid America Conversion Services	Portsmouth, Ohio, USA	13,500
	Paducah, Kentucky, USA	18,000
INIS Fluorine Products	Hobbs, New Mexico, USA	6500 (construction on hold)
Tenex	Zelenogorsk, Russian Federation	10,000

Russia’s W-ECP deconversion plant is at Zelenogorsk Electrochemical Plant (ECP) in Siberia. The 10,000 tU/yr deconversion (defluorination) plant was built by Tenex under a technology transfer agreement with Areva NC, so that depleted uranium can be stored long-term as uranium oxide, and HF is produced as a by-product. The W-ECP plant is similar to Areva’s W2 plant at Pierrelatte in France and has mainly west European equipment. It was commissioned in December 2009.

Russia is also building a plant at Angarsk to deconvert UF₆ to UF₄, recovering some HF in the process. Capacity of 2000 tU/yr was planned for 2012, with subsequent increase to 6000 tU/yr.

These use essentially a dry process, with no liquid effluent. It is the same as that used for the enriched portion, albeit at a scale of 20,000 tU/yr in the one plant.

The UF₆ is first vapourised in autoclaves with steam, then the uranyl fluoride (UO₂F₂) is reacted with hydrogen at 700°C to yield an HF product for sale to converters and U₃O₈ powder which is packed into 10-tonne containers for storage.

UF₆ + 2H₂O ==⇒ UO₂F₂ + 4HF

3UO₂F₂ + 2H₂O + H₂ ===> U₃O₈ + 6HF

The INIS plant in Idaho uses a slightly different deconversion followed by fluorine extraction process (FEP), on a toll basis. The deconversion plant had been used to produce DU metal for the military and was purchased by INIS. In this, the depleted UF6 is first vapourised in autoclaves and hydrogen is added to give depleted UO2 and anhydrous UF4 which is the main product for sale. The FEP then involves reacting some UF₄ with silica to give silicon fluoride (SiF₄) as a commercial co-product.

Ownership title is normally transferred to the enricher as part of the commercial deal. It is sometimes considered as a waste, though only for legal or regulatory reasons and in the USA, but usually it is understood as a long-term strategic resource which can be used in a future generation of fast neutron reactors. Any much more efficient enrichment process would also make it into an immediately usable resource to supply more U-235. Enrichment companies with ownership of large amounts of depleted uranium are quite clear that their stocks are a significant asset, though Urenco speaks of deconversion being for long-term storage “prior to geological disposal”.

Notes & references

General references

World Nuclear Association,The Nuclear Fuel Report 2017-2035 (September 2017)

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Enriched uranium

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Proportions of uranium-238 (blue) and uranium-235 (red) found naturally versus enriched grades

Enriched uranium is a type of uranium in which the percent composition of uranium-235 has been increased through the process of isotope separation. Natural uranium is 99.284% ²³⁸U isotope, with ²³⁵U only constituting about 0.711% of its mass. ²³⁵U is the only nuclide existing in nature (in any appreciable amount) that is fissile with thermal neutrons.^[1]
Enriched uranium is a critical component for both civil nuclear power generation and military nuclear weapons. The International Atomic Energy Agency attempts to monitor and control enriched uranium supplies and processes in its efforts to ensure nuclear power generation safety and curb nuclear weapons proliferation.
During the Manhattan Project enriched uranium was given the codename oralloy, a shortened version of Oak Ridge alloy, after the location of the plants where the uranium was enriched. The term oralloy is still occasionally used to refer to enriched uranium. There are about 2,000 tonnes (t, Mg) of highly enriched uranium in the world,^[2] produced mostly for nuclear power, nuclear weapons, naval propulsion, and smaller quantities for research reactors.
The ²³⁸U remaining after enrichment is known as depleted uranium (DU), and is considerably less radioactive than even natural uranium, though still very dense and extremely hazardous in granulated form – such granules are a natural by-product of the shearing action that makes it useful for armor-penetrating weapons and radiation shielding. At present, 95 percent of the world's stocks of depleted uranium remain in secure storage.

Grades

Uranium as it is taken directly from the Earth is not suitable as fuel for most nuclear reactors and requires additional processes to make it usable. Uranium is mined either underground or in an open pit depending on the depth at which it is found. After the uranium ore is mined, it must go through a milling process to extract the uranium from the ore. This is accomplished by a combination of chemical processes with the end product being concentrated uranium oxide, which is known as "yellowcake", contains roughly 60% uranium whereas the ore typically contains less than 1% uranium and as little as 0.1% uranium (Henderson 2000). After the milling process is complete, the uranium must next undergo a process of conversion, "to either uranium dioxide, which can be used as the fuel for those types of reactors that do not require enriched uranium, or into uranium hexafluoride, which can be enriched to produce fuel for the majority of types of reactors". Naturally-occurring uranium is made of a mixture of U-235 and U-238. The U-235 is fissile meaning it is easily split with neutrons while the remainder is U-238, but in nature, more than 99% of the extracted ore is U-238. Most nuclear reactors require enriched uranium, which is uranium with higher concentrations of U-235 ranging between 3.5% and 4.5%. There are two commercial enrichment processes: gaseous diffusion and gas centrifugation. Both enrichment processes involve the use of uranium hexafluoride and produce enriched uranium oxide.

A drum of yellowcake (a mixture of uranium precipitates)

Reprocessed uranium (RepU)

Reprocessed uranium (RepU) is a product of nuclear fuel cycles involving nuclear reprocessing of spent fuel. RepU recovered from light water reactor (LWR) spent fuel typically contains slightly more U-235 than natural uranium, and therefore could be used to fuel reactors that customarily use natural uranium as fuel, such as CANDU reactors. It also contains the undesirable isotope uranium-236, which undergoes neutron capture, wasting neutrons (and requiring higher U-235 enrichment) and creating neptunium-237, which would be one of the more mobile and troublesome radionuclides in deep geological repository disposal of nuclear waste.

Low enriched uranium (LEU)

Low enriched uranium (LEU) has a lower than 20% concentration of ²³⁵U; for instance, in commercial light water reactors (LWR), the most prevalent power reactors in the world, uranium is enriched to 3 to 5% ²³⁵U. Fresh LEU used in research reactors is usually enriched 12% to 19.75% U-235, the latter concentration being used to replace HEU fuels when converting to LEU.^[3]

Highly enriched uranium (HEU)

A billet of highly enriched uranium metal

Highly enriched uranium (HEU) has a 20% or higher concentration of ²³⁵U. The fissile uranium in nuclear weapon primaries usually contains 85% or more of ²³⁵U known as weapons-grade, though theoretically for an implosion design, a minimum of 20% could be sufficient (called weapon(s)-usable) although it would require hundreds of kilograms of material and "would not be practical to design";^[4]^[5] even lower enrichment is hypothetically possible, but as the enrichment percentage decreases the critical mass for unmoderated fast neutrons rapidly increases, with for example, an infinite mass of 5.4% ²³⁵U being required.^[4] For criticality experiments, enrichment of uranium to over 97% has been accomplished.^[6]
The very first uranium bomb, Little Boy, dropped by the United States on Hiroshima in 1945, used 64 kilograms of 80% enriched uranium. Wrapping the weapon's fissile core in a neutron reflector (which is standard on all nuclear explosives) can dramatically reduce the critical mass. Because the core was surrounded by a good neutron reflector, at explosion it comprised almost 2.5 critical masses. Neutron reflectors, compressing the fissile core via implosion, fusion boosting, and "tamping", which slows the expansion of the fissioning core with inertia, allow nuclear weapon designs that use less than what would be one bare-sphere critical mass at normal density. The presence of too much of the ²³⁸U isotope inhibits the runaway nuclear chain reaction that is responsible for the weapon's power. The critical mass for 85% highly enriched uranium is about 50 kilograms (110 lb), which at normal density would be a sphere about 17 centimetres (6.7 in) in diameter.
Later US nuclear weapons usually use plutonium-239 in the primary stage, but the jacket or tamper secondary stage, which is compressed by the primary nuclear explosion often uses HEU with enrichment between 40% and 80%^[7] along with the fusion fuel lithium deuteride. For the secondary of a large nuclear weapon, the higher critical mass of less-enriched uranium can be an advantage as it allows the core at explosion time to contain a larger amount of fuel. The ²³⁸U is not fissile but still fissionable by fusion neutrons.
HEU is also used in fast neutron reactors, whose cores require about 20% or more of fissile material, as well as in naval reactors, where it often contains at least 50% ²³⁵U, but typically does not exceed 90%. The Fermi-1 commercial fast reactor prototype used HEU with 26.5% ²³⁵U. Significant quantities of HEU are used in the production of medical isotopes, for example molybdenum-99 for technetium-99m generators.^[8]

Enrichment methods

Isotope separation is difficult because two isotopes of the same element have very nearly identical chemical properties, and can only be separated gradually using small mass differences. (²³⁵U is only 1.26% lighter than ²³⁸U.) This problem is compounded by the fact that uranium is rarely separated in its atomic form, but instead as a compound (²³⁵UF₆ is only 0.852% lighter than ²³⁸UF₆.) A cascade of identical stages produces successively higher concentrations of ²³⁵U. Each stage passes a slightly more concentrated product to the next stage and returns a slightly less concentrated residue to the previous stage.
There are currently two generic commercial methods employed internationally for enrichment: gaseous diffusion (referred to as first generation) and gas centrifuge (second generation), which consumes only 2% to 2.5%^[9] as much energy as gaseous diffusion (at least a "factor of 20" more efficient).^[10] Some work is being done that would use nuclear resonance; however there is no reliable evidence that any nuclear resonance processes have been scaled up to production.

Diffusion techniques

Gaseous diffusion

Gaseous diffusion is a technology used to produce enriched uranium by forcing gaseous uranium hexafluoride (hex) through semi-permeable membranes. This produces a slight separation between the molecules containing ²³⁵U and ²³⁸U. Throughout the Cold War, gaseous diffusion played a major role as a uranium enrichment technique, and as of 2008 accounted for about 33% of enriched uranium production,^[11] but in 2011 was deemed an obsolete technology that is steadily being replaced by the later generations of technology as the diffusion plants reach their ends-of-life.^[12] In 2013, the Paducah facility in the US ceased operating, it was the last commercial ²³⁵U gaseous diffusion plant in the world.^[13]

Thermal diffusion

Thermal diffusion utilizes the transfer of heat across a thin liquid or gas to accomplish isotope separation. The process exploits the fact that the lighter ²³⁵U gas molecules will diffuse toward a hot surface, and the heavier ²³⁸U gas molecules will diffuse toward a cold surface. The S-50 plant at Oak Ridge, Tennessee was used during World War II to prepare feed material for the EMIS process. It was abandoned in favor of gaseous diffusion.

Centrifuge techniques

Gas centrifuge

A cascade of gas centrifuges at a U.S. enrichment plant

The gas centrifuge process uses a large number of rotating cylinders in series and parallel formations. Each cylinder's rotation creates a strong centripetal force so that the heavier gas molecules containing ²³⁸U move tangentially toward the outside of the cylinder and the lighter gas molecules rich in ²³⁵U collect closer to the center. It requires much less energy to achieve the same separation than the older gaseous diffusion process, which it has largely replaced and so is the current method of choice and is termed second generation. It has a separation factor per stage of 1.3 relative to gaseous diffusion of 1.005,^[11] which translates to about one-fiftieth of the energy requirements. Gas centrifuge techniques produce close to 100% of the world's enriched uranium.

Zippe centrifuge

Diagram of the principles of a Zippe-type gas centrifuge with U-238 represented in dark blue and U-235 represented in light blue

The Zippe centrifuge is an improvement on the standard gas centrifuge, the primary difference being the use of heat. The bottom of the rotating cylinder is heated, producing convection currents that move the ²³⁵U up the cylinder, where it can be collected by scoops. This improved centrifuge design is used commercially by Urenco to produce nuclear fuel and was used by Pakistan in their nuclear weapons program.

Laser techniques

Laser processes promise lower energy inputs, lower capital costs and lower tails assays, hence significant economic advantages. Several laser processes have been investigated or are under development. Separation of Isotopes by Laser Excitation (SILEX) is well advanced and licensed for commercial operation in 2012.

Atomic vapor laser isotope separation (AVLIS)

Atomic vapor laser isotope separation employs specially tuned lasers^[14] to separate isotopes of uranium using selective ionization of hyperfine transitions. The technique uses lasers tuned to frequencies that ionize ²³⁵U atoms and no others. The positively charged ²³⁵U ions are then attracted to a negatively charged plate and collected.

Molecular laser isotope separation (MLIS)

Molecular laser isotope separation uses an infrared laser directed at UF₆, exciting molecules that contain a ²³⁵U atom. A second laser frees a fluorine atom, leaving uranium pentafluoride, which then precipitates out of the gas.

Separation of Isotopes by Laser Excitation (SILEX)

Separation of isotopes by laser excitation is an Australian development that also uses UF₆. After a protracted development process involving U.S. enrichment company USEC acquiring and then relinquishing commercialization rights to the technology, GE Hitachi Nuclear Energy (GEH) signed a commercialization agreement with Silex Systems in 2006.^[15] GEH has since built a demonstration test loop and announced plans to build an initial commercial facility.^[16] Details of the process are classified and restricted by intergovernmental agreements between United States, Australia, and the commercial entities. SILEX has been projected to be an order of magnitude more efficient than existing production techniques but again, the exact figure is classified.^[11] In August, 2011 Global Laser Enrichment, a subsidiary of GEH, applied to the U.S. Nuclear Regulatory Commission (NRC) for a permit to build a commercial plant.^[17] In September 2012, the NRC issued a license for GEH to build and operate a commercial SILEX enrichment plant, although the company had not yet decided whether the project would be profitable enough to begin construction, and despite concerns that the technology could contribute to nuclear proliferation.^[18]

Other techniques

Aerodynamic processes

Schematic diagram of an aerodynamic nozzle. Many thousands of these small foils would be combined in an enrichment unit.

The X-ray based LIGA manufacturing process was originally developed at the Forschungszentrum Karlsruhe, Germany, to produce nozzles for isotope enrichment.^[19]

Aerodynamic enrichment processes include the Becker jet nozzle techniques developed by E. W. Becker and associates using the LIGA process and the vortex tube separation process. These aerodynamic separation processes depend upon diffusion driven by pressure gradients, as does the gas centrifuge. They in general have the disadvantage of requiring complex systems of cascading of individual separating elements to minimize energy consumption. In effect, aerodynamic processes can be considered as non-rotating centrifuges. Enhancement of the centrifugal forces is achieved by dilution of UF₆ with hydrogen or helium as a carrier gas achieving a much higher flow velocity for the gas than could be obtained using pure uranium hexafluoride. The Uranium Enrichment Corporation of South Africa (UCOR) developed and deployed the continuous Helikon vortex separation cascade for high production rate low enrichment and the substantially different semi-batch Pelsakon low production rate high enrichment cascade both using a particular vortex tube separator design, and both embodied in industrial plant.^[20] A demonstration plant was built in Brazil by NUCLEI, a consortium led by Industrias Nucleares do Brasil that used the separation nozzle process. However all methods have high energy consumption and substantial requirements for removal of waste heat; none are currently still in use.

Electromagnetic isotope separation

Main article: Calutron

Schematic diagram of uranium isotope separation in a calutron shows how a strong magnetic field is used to redirect a stream of uranium ions to a target, resulting in a higher concentration of uranium-235 (represented here in dark blue) in the inner fringes of the stream.

In the electromagnetic isotope separation process (EMIS), metallic uranium is first vaporized, and then ionized to positively charged ions. The cations are then accelerated and subsequently deflected by magnetic fields onto their respective collection targets. A production-scale mass spectrometer named the Calutron was developed during World War II that provided some of the ²³⁵U used for the Little Boy nuclear bomb, which was dropped over Hiroshima in 1945. Properly the term 'Calutron' applies to a multistage device arranged in a large oval around a powerful electromagnet. Electromagnetic isotope separation has been largely abandoned in favour of more effective methods.

Chemical methods

One chemical process has been demonstrated to pilot plant stage but not used for production. The French CHEMEX process exploited a very slight difference in the two isotopes' propensity to change valency in oxidation/reduction, utilising immiscible aqueous and organic phases. An ion-exchange process was developed by the Asahi Chemical Company in Japan that applies similar chemistry but effects separation on a proprietary resin ion-exchange column.

Plasma separation

Plasma separation process (PSP) describes a technique that makes use of superconducting magnets and plasma physics. In this process, the principle of ion cyclotron resonance is used to selectively energize the ²³⁵U isotope in a plasma containing a mix of ions. The French developed their own version of PSP, which they called RCI. Funding for RCI was drastically reduced in 1986, and the program was suspended around 1990, although RCI is still used for stable isotope separation.

Separative work unit

"Separative work" – the amount of separation done by an enrichment process – is a function of the concentrations of the feedstock, the enriched output, and the depleted tailings; and is expressed in units that are so calculated as to be proportional to the total input (energy / machine operation time) and to the mass processed. Separative work is not energy. The same amount of separative work will require different amounts of energy depending on the efficiency of the separation technology. Separative work is measured in Separative work units SWU, kg SW, or kg UTA (from the German Urantrennarbeit – literally uranium separation work)

1 SWU = 1 kg SW = 1 kg UTA
1 kSWU = 1 tSW = 1 t UTA
1 MSWU = 1 ktSW = 1 kt UTA

Cost issues

In addition to the separative work units provided by an enrichment facility, the other important parameter to be considered is the mass of natural uranium (NU) that is needed to yield a desired mass of enriched uranium. As with the number of SWUs, the amount of feed material required will also depend on the level of enrichment desired and upon the amount of ²³⁵U that ends up in the depleted uranium. However, unlike the number of SWUs required during enrichment, which increases with decreasing levels of ²³⁵U in the depleted stream, the amount of NU needed will decrease with decreasing levels of ²³⁵U that end up in the DU.
For example, in the enrichment of LEU for use in a light water reactor it is typical for the enriched stream to contain 3.6% ²³⁵U (as compared to 0.7% in NU) while the depleted stream contains 0.2% to 0.3% ²³⁵U. In order to produce one kilogram of this LEU it would require approximately 8 kilograms of NU and 4.5 SWU if the DU stream was allowed to have 0.3% ²³⁵U. On the other hand, if the depleted stream had only 0.2% ²³⁵U, then it would require just 6.7 kilograms of NU, but nearly 5.7 SWU of enrichment. Because the amount of NU required and the number of SWUs required during enrichment change in opposite directions, if NU is cheap and enrichment services are more expensive, then the operators will typically choose to allow more ²³⁵U to be left in the DU stream whereas if NU is more expensive and enrichment is less so, then they would choose the opposite.

Uranium enrichment calculator designed by the WISE Uranium Project

Downblending

The opposite of enriching is downblending; surplus HEU can be downblended to LEU to make it suitable for use in commercial nuclear fuel.
The HEU feedstock can contain unwanted uranium isotopes: ²³⁴U is a minor isotope contained in natural uranium; during the enrichment process, its concentration increases but remains well below 1%. High concentrations of ²³⁶U are a byproduct from irradiation in a reactor and may be contained in the HEU, depending on its manufacturing history. HEU reprocessed from nuclear weapons material production reactors (with an ²³⁵U assay of approx. 50%) may contain ²³⁶U concentrations as high as 25%, resulting in concentrations of approximately 1.5% in the blended LEU product. ²³⁶U is a neutron poison; therefore the actual ²³⁵U concentration in the LEU product must be raised accordingly to compensate for the presence of ²³⁶U.
The blendstock can be NU, or DU, however depending on feedstock quality, SEU at typically 1.5 wt% ²³⁵U may used as a blendstock to dilute the unwanted byproducts that may be contained in the HEU feed. Concentrations of these isotopes in the LEU product in some cases could exceed ASTM specifications for nuclear fuel, if NU, or DU were used. So, the HEU downblending generally cannot contribute to the waste management problem posed by the existing large stockpiles of depleted uranium.
A major downblending undertaking called the Megatons to Megawatts Program converts ex-Soviet weapons-grade HEU to fuel for U.S. commercial power reactors. From 1995 through mid-2005, 250 tonnes of high-enriched uranium (enough for 10,000 warheads) was recycled into low-enriched-uranium. The goal is to recycle 500 tonnes by 2013. The decommissioning programme of Russian nuclear warheads accounted for about 13% of total world requirement for enriched uranium leading up to 2008.^[11]
The United States Enrichment Corporation has been involved in the disposition of a portion of the 174.3 tonnes of highly enriched uranium (HEU) that the U.S. government declared as surplus military material in 1996. Through the U.S. HEU Downblending Program, this HEU material, taken primarily from dismantled U.S. nuclear warheads, was recycled into low-enriched uranium (LEU) fuel, used by nuclear power plants to generate electricity.^[21]

A uranium downblending calculator designed by the WISE Uranium Project

Global enrichment facilities

The following countries are known to operate enrichment facilities: Argentina, Brazil, China, France, Germany, India, Iran, Japan, the Netherlands, North Korea, Pakistan, Russia, the United Kingdom, and the United States.^[22]^[23] Belgium, Iran, Italy, and Spain hold an investment interest in the French Eurodif enrichment plant, with Iran's holding entitling it to 10% of the enriched uranium output. Countries that had enrichment programs in the past include Libya and South Africa, although Libya's facility was never operational.^[24] Australia has developed a laser enrichment process known as SILEX, which it intends to pursue through financial investment in a U.S. commercial venture by General Electric.^[25] It has also been claimed that Israel has a uranium enrichment program housed at the Negev Nuclear Research Center site near Dimona.^[26]

References

OECD Nuclear Energy Agency (2003). Nuclear Energy Today. OECD Publishing. p. 25. ISBN 9789264103283.

Thomas B. Cochran (Natural Resources Defense Council) (12 June 1997). "Safeguarding Nuclear Weapon-Usable Materials in Russia" (PDF). Proceedings of international forum on illegal nuclear traffic. Archived from the original (PDF) on 22 July 2012.

Alexander Glaser (6 November 2005). "About the Enrichment Limit for Research Reactor Conversion : Why 20%?" (PDF). Princeton University. Retrieved 18 April 2014.

Forsberg, C. W.; Hopper, C. M.; Richter, J. L.; Vantine, H. C. (March 1998). "Definition of Weapons-Usable Uranium-233" (PDF). ORNL/TM-13517. Oak Ridge National Laboratories. Archived from the original (PDF) on 2 November 2013. Retrieved 30 October 2013.

Sublette, Carey (4 October 1996). "Nuclear Weapons FAQ, Section 4.1.7.1: Nuclear Design Principles – Highly Enriched Uranium". Nuclear Weapons FAQ. Retrieved 2 October 2010.

Mosteller, R.D. (1994). "Detailed Reanalysis of a Benchmark Critical Experiment: Water-Reflected Enriched-Uranium Sphere" (PDF). Los Alamos technical paper (LA–UR–93–4097): 2. Retrieved 19 December 2007. The enrichment of the pin and of one of the hemispheres was 97.67 w/o, while the enrichment of the other hemisphere was 97.68 w/o.

"Nuclear Weapons FAQ". Nuclearweaponarchive.org. Retrieved 26 January 2013.

Frank N. Von Hippel; Laura H. Kahn (December 2006). "Feasibility of Eliminating the Use of Highly Enriched Uranium in the Production of Medical Radioisotopes". Science & Global Security. 14 (2 & 3): 151–162. doi:10.1080/08929880600993071. Retrieved 26 March 2010.

"Uranium Enrichment". world-nuclear.org.

Economic Perspective for Uranium Enrichment (PDF), The throughput per centrifuge unit is very small compared to that of a diffusion unit so small, in fact, that it is not compensated by the higher enrichment per unit. To produce the same amount of reactor-grade fuel requires a considerably larger number (approximately 50,000 to 500,000] of centrifuge units than diffusion units. This disadvantage, however, is outweighed by the considerably lower (by a factor of 20) energy consumption per SWU for the gas centrifuge

"Lodge Partners Mid-Cap Conference 11 April 2008" (PDF). Silex Ltd. 11 April 2008.

Rod Adams (24 May 2011). "McConnell asks DOE to keep using 60 year old enrichment plant to save jobs". Atomic Insights. Archived from the original on 28 January 2013. Retrieved 26 January 2013.

"Paducah enrichment plant to be closed. The 1950s facility is the last remaining gaseous diffusion uranium enrichment plant in the world.".

F. J. Duarte and L.W. Hillman (Eds.), Dye Laser Principles (Academic, New York, 1990) Chapter 9.

[1] Archived 23 July 2015 at the Wayback Machine

"GE Hitachi Nuclear Energy Selects Wilmington, N.C. as Site for Potential Commercial Uranium Enrichment Facility". Business Wire. 30 April 2008. Retrieved 30 September 2012.

Broad, William J. (20 August 2011). "Laser Advances in Nuclear Fuel Stir Terror Fear". The New York Times. Retrieved 21 August 2011.

"Uranium Plant Using Laser Technology Wins U.S. Approval". New York Times. September 2012.

Becker, E. W.; Ehrfeld, W.; Münchmeyer, D.; Betz, H.; Heuberger, A.; Pongratz, S.; Glashauser, W.; Michel, H. J.; Siemens, R. (1982). "Production of Separation-Nozzle Systems for Uranium Enrichment by a Combination of X-Ray Lithography and Galvanoplastics". Naturwissenschaften. 69 (11): 520–523. Bibcode:1982NW.....69..520B. doi:10.1007/BF00463495.

Smith, Michael; Jackson A G M (2000). "Dr". South African Institution of Chemical Engineers – Conference 2000: 280–289.

[2] Archived 6 April 2001 at the Wayback Machine

Arjun Makhijani; Lois Chalmers; Brice Smith (15 October 2004). Uranium enrichment (PDF). Institute for Energy and Environmental Research. Retrieved 21 November 2009.

Australia's uranium - Greenhouse friendly fuel for an energy hungry world (PDF). Standing Committee on Industry and Resources (Report). The Parliament of the Commonwealth of Australia. November 2006. p. 730. Retrieved 3 April 2015.

BBC (1 September 2006). "Q&A: Uranium enrichment". BBC News. Retrieved 3 January 2010.

"Laser enrichment could cut cost of nuclear power". The Sydney Morning Herald. 26 May 2006.

"Israel's Nuclear Weapons Program". Nuclear Weapon Archive. 10 December 1997. Retrieved 7 October 2007.

External links

Look up enriched uranium in Wiktionary, the free dictionary.

Annotated bibliography on enriched uranium from the Alsos Digital Library for Nuclear Issues
Silex Systems Ltd
Uranium Enrichment, World Nuclear Association
Overview and history of U.S. HEU production
News Resource on Uranium Enrichment
Nuclear Chemistry-Uranium Enrichment
A busy year for SWU (a 2008 review of the commercial enrichment marketplace), Nuclear Engineering International, 1 September 2008
Uranium Enrichment and Nuclear Weapon Proliferation, by Allan S. Krass, Peter Boskma, Boelie Elzen and Wim A. Smit, 296 pp., published for SIPRI by Taylor and Francis Ltd, London, 1983
Poliakoff, Martyn (2009). "How do you enrich Uranium?". The Periodic Table of Videos. University of Nottingham.
Gilinsky, V.; Hoehn, W. (December 1969). "The Military Significance of Small Uranium Enrichment Facilities Fed with Low-Enrichment Uranium (Redacted)". Defense Technical Information Center. RAND Corporation.

Categories:

China's Nuclear Fuel Cycle

(Updated January 2019)

http://www.world-nuclear.org/information-library/country-profiles/countries-a-f/china-nuclear-fuel-cycle.aspx

China has become self-sufficient in most aspects of the fuel cycle.
China aims to produce one-third of its uranium domestically, obtain one-third through foreign equity in mines and joint ventures overseas, and to purchase one-third on the open market.
China's two major enrichment plants were built under agreements with Russia but much current capacity is indigenous.
China’s R&D investment in nuclear technologies is very significant, particularly in high-temperature gas-cooled and molten salt-cooled reactors.

China has stated it intends to become self-sufficient not just in nuclear power plant capacity, but also in the production of fuel for those plants. However, the country still relies to some extent on foreign suppliers for all stages of the fuel cycle, from uranium mining through fabrication and reprocessing, but mostly for uranium supply. As China rapidly increases the number of new reactors, it has also initiated a number of domestic projects, often in cooperation with foreign suppliers, to meet its nuclear fuel needs.

The national policy is to obtain about one-third of uranium supply domestically, one-third from Chinese equity in foreign mines, and one-third on the open market. Increasingly, other stages of the fuel cycle will be indigenous. Uranium demand in 2020 is expected to be over 11,000 tU (with 58 reactors operating), in 2025 about 18,500 tU (for 100 reactors) and in 2030 about 24,000 tU (for 130 reactors). UxC reports that China imported over 115,000 tU over 2009-14, notably 25,000 tU in 2014 and 10,400 tU to July in 2015. With annual consumption currently about 9000 tU, much of this will be stockpiled.

China National Nuclear Corporation (CNNC) maintains a strong monopoly on the nuclear fuel cycle in China, notably the front end, forcing China General Nuclear Power (CGN) to work around this, principally with international ventures, some involving large capital outlays. With the merger of SNPTC and CPI to form SPI in 2015, so that SNPTC took over all the nuclear-related business of CPI to function as an active subsidiary of SPI, SNPTC said it intended to get into both uranium mining and fuel fabrication.

CNNC is also the main operator in the fuel cycle back end, evidenced by a series of agreements with Areva for a reprocessing plant. That in November 2015 was part of a wider agreement in relation to all aspects of the fuel cycle, and foreshadowing an intention to take equity in Areva NC (now Orano), in connection with evolving agreements to build a reprocessing plant based on Areva technology.

Following Areva’s restructuring, a new framework agreement between what is now Orano and CNNC was signed in February 2017, covering “the whole industrial chain of the nuclear fuel cycle.” In particular it supports plans for construction of a reprocessing plant in China. The commitmment was reaffirmed in January 2018 through the signing of an MoU.

As well as a long-standing close relationship with France, China has a bilateral nuclear cooperation agreement (‘123 agreement’) with the USA from 1985 which was renewed in 2015. This is a prerequisite for nuclear trade in plant and materials that involves the USA.

Domestic uranium resources and mining

CNNC is the only current supplier of domestic uranium. CGN has responded energetically to this situation through its subsidiary China Guangdong Nuclear Uranium Resources Co Ltd (CGN-URC) as described below.

China now claims to be “a uranium-rich country” on the basis of some two million tonnes of uranium, though published known in situ uranium resources were 366,000 tU to $130/kg at 1/1/15, of which 173,000 tU were reasonably assured, and in situ inferred resources were 193,000 tU in the 2016 edition of the 'Red Book', which are modest in relation to the country's needs. New discoveries in the north and northwest in sandstones, and deep hydrothermal ones in southeast China have raised expectations. There is also potential in lignite, black shale and phosphates. Over 2013-14 about 71,000 tU was added to known resources in northern China – in the Yili, Erlian, Erdos, Songliao and Bayingebi basins as well as Longshoushan – and 29,000 tU in southern China in the Rouoergai and Dazhou uranium fields. The 2016 Red Book tabulates 366,000 tU in 21 deposits in 13 provinces, 39% of the total in Inner Mongolia, 21% in Jiangxi, 14% in Xinjiang and 12% in Guangdong.

As of 2012, 35% of resources were in sandstone deposits mainly in the north and northwest, 28% in vein/granite deposits in central and southeast China, 21% in volcanic deposits in the southeast, and 10% in black shale in the southeast. Most known resources are at less than 500m depth.

Domestic production was 1616 tU/yr in 2015, enough for about 7000 MWe, apart from new cores. This was approximately 530 t from sandstone by ISL, 620 t from granite-related ore and 450 t from volcanic-related ore. All production is acid-leached. By international standards, China's ores are low-grade and production has been inefficient. The nuclear power companies are not depending on the national goal of sourcing one-third of uranium domestically, and are ramping up international arrangements to obtain fuel.

Operating uranium mines in China

Mine^a	Province	Type	Nominal capacity tonnes U/yr (planned)	Started
Yining	Xinjiang	In-situ leach (ISL)	480 (800)	1993
Lantian	Shaanxi	Underground, heap leach	100	1993
Benxi	Liaoning	Underground, block leach	120	1996
Qinglong	Liaoning	Underground, heap leach	100 (200)	2007
Fuzhou	Jiangxi	Underground, mill	350 (500)	1966
Chongyi	Jiangxi	Underground, heap leach	200 (300)	1979
Shaoguan	Guangdong	Underground, heap leach	200 (300)	2008
Total			1550 (2320)

Xinjiang's Yili basin in the far west of China, in which the Yining (or Kujiltai) ISL mine sits, is contiguous with the Ili uranium province in Kazakhstan, though the geology is apparently different. The Fuzhou mine in the southeastern Jiangxi province is in a volcanic hydrothermal deposit, as is Qinglong in Liaoning. The other mines are in granitic deposits. Source: Red Book 2016.

China National Uranium Corporation (CNUC or CUC), a subsidiary of CNNC, operates these mines. Pilot testing is under way on the Shihongtan deposit in the Turpan-Hami basin of Xinjiang, and the western portion appears suitable for ISL. A uranium-molybdenum mine is being developed at Guyuan, Hebei province, in granites. Other uranium deposits with abundant reserves but with complex mining and milling technologies are the subject of pilot tests and feasibility studies, such as the Dongsheng and Erlian sandstone deposits in Inner Mongolia. The former, in the Ordos/Erdos Basin, has an estimated 30,000 tonnes of uranium in a palaeochannel system, the latter is unsuitable for ISL due to low permeability.

An underground uranium mine at Hengyang in Hunan is on stand-by. The mine, which started up in 1963, has a nominal production capacity of 500-1000 tU/yr.

CGN subsidiary China Guangdong Nuclear Uranium Resources Co Ltd (CGN-URC) was set up in 2006 to be responsible for CGN's fuel supply, and in particular to undertake uranium exploration and mining, uranium trade, and management of fuel processing for CGN. It is pursuing the second stage of a planned three-stage development, with diversification of supplies and integration of front-end services. A third stage will involve new technology as well as consolidation of its role as viable supplier. It aims to free up international trade and bring about better logistics.

CGN-URC has been undertaking uranium exploration in Xinjiang Uygur autonomous region, and also in Guangdong, via CGN-URC Guangdong Uranium Ltd. In May 2011 CGN-URC announced that it was developing two 500 tU/yr mines on these deposits, to operate from 2013, but this venture appears to have stalled.

Mineral exploration

CNNC's Geological Survey Bureau and the Beijing Research Institute of Uranium Geology are the key organisations involved with a massive increase in exploration effort since 2000, focused on sandstone deposits amenable to ISL in the Xinjiang and Inner Mongolia regions, and the granite and volcanic metallogenic belts in southern China, including the Xiangshen uranium orefield.

In northern China, the exploration is focused on previously discovered mineralisation spanning the Yili, Turpan-Hami, Junggar and Tarim basins of Xinjiang Autonomous Region, and the Erdos/Ordos, Erlian, Songliao, Badanjili and Bayingebi basins of Inner Mongolia. The Ordos basin itself covers over 250,000 sq. km of Shaanxui, Shanxi, Gansu and Inner Mongolia and contains major coal units as well as commercial gas reservoirs and some oil. It starts just north on Xi’an in Shaanxi province and extends nearly to Baotou near the Mongolian border. By 2012 this had become the premier uranium region of China, right across its north. In 2008 significant deposits were discovered in the Yili basin of Xinjiang, including J3, and then in the Ordos basin Nalinggou, Darong and (in 2012) Daying were discovered. Daying is expected to become China’s largest uranium resource and in late 2014 was being described by the Geological Survey Bureau as ‘world class’. Also in the Erlian basin the Bayanwula deposit, a roll front deposit with biogenic origins, was identified. In the Songliao basin in the east of Inner Mongolia the Qianjiadian deposit was identified, and in 2017 CNNC announced "a breakthrough in sandstone-type uranium ore exploration," and expects a new orebody – with an overall length of more than 10 km – to develop into a large uranium deposit.

CNNC Inner Mongolia Mining Industry LLC based in Baotou is responsible for overseeing natural uranium geological prospecting, scientific research and project management in the middle and western parts of Inner Mongolia. Its Mining Business Division is focused mainly evaluating the Nalinggou and Bayanwula projects by the end of 2015. The Division is also setting up regional headquarters in Inner Mongolia, Jiangxi, Guangdong and Xinjiang.

Some northern uranium mineralization is interbedded with coal deposits, giving rise to concerns about mining efficiently, and about the amount of radioactivity in coal as burned in some northern power stations. The Daying uranium deposit in Inner Mongolia is evidently in this category, with separate layers of coal and uranium ore in sandstone palaeochannels extending over many kilometres. The coal resource is major.

In March 2013 CNNC signed an agreement with China Petroleum & Chemical Corp. (SINOPEC) to set up the joint venture of CNNC and SINOPEC Uranium Resources Co. Ltd to accelerate the exploration for uranium resources, starting with the Chaideng area of Inner Mongolia. The Chaideng prospecting region of Dongsheng Coal Field is in the northeast of the Ordos Basin.

In August 2014 CNNC signed an agreement with Shenhua Group to recover uranium from a mine near Ordos city. In March 2016 it signed a broader strategic agreement. Shenhua is the largest coal mining company in China.

The Dongsheng group of uranium deposits is located in south-central Inner Mongolia, about 100 km south of Baotou and on the northern edge of the Ordos Basin. Uranium ore bodies are mostly in area of 200 sq. km hosted by fluvial sandstones in the Zhiluo Formation as a regional redox front, and to a lesser degree within the Yan'an Formation, which has coal-bearing strata. Individual tabular and roll-front ore bodies are several tens to one hundred metres long, up to 20 m thick, and have average ore grades of 0.02 to 0.05%U. They plunge from 75 to 185 m deep, following the dip of the Zhiluo formation.

International uranium sources

Increasingly, uranium is imported from Kazakhstan, Uzbekistan, Canada, Namibia, Niger and Australia. In 2012 imports were 12,908 tU, and in 2013 China imported 18,968 tonnes of uranium for $2.37 billion from five countries (Kazakhstan, Uzbekistan, Australia, Namibia and Canada) according to China’s General Administration of Customs. Anticipated need in 2015, including new cores, was 8160 tU. International sources are both from Chinese equity in mines and uranium bought on the open market.

Chinese equity in uranium mines in other countries

Company	Country	Mine	Equity %	Start production with China equity
CNUC	Niger	Azelik	37.2 + 24.8 ZXJOY	2010 but now closed
	Niger	Imouraren	25+, more pending	On hold
	Namibia	Langer Heinrich	25+, more pending	2014
	Namibia	Rössing	68.6	2019
	Kazakhstan	Zhalpak	49?	2017?
CGN-URC	Namibia	Husab	90	2016
	Kazakhstan	Irkol & Semizbai	49	2008, 2009
	Uzbekistan	Boztau black shales	50	Uncertain
	Canada	Patterson Lake	19.99	2023

CNNC initiatives abroad

With the prospective need to import much more uranium, China Nuclear International Uranium Corporation (SinoU) was set up by CNNC to acquire equity in uranium resources internationally. It set up the Azelik mine in Niger and has agreed to buy a 10% share of Areva’s Imouraren project there for €200 million. It is now consolidated into CNUC.

In January 2014 it bought a 25% stake in Paladin’s Langer Heinrich mine in Namibia for $190 million, entitling it to that share of output. In November 2018, it bought Rio Tinto’s majority stake (68.6%) in Namibia’s Rössing mine.

It has investigated prospects in Kazakhstan, Uzbekistan, Mongolia, Namibia, Algeria and Zimbabwe. Canada and South Africa are also seen as potential suppliers for SinoU. About 2007 it bought a share (49%?) in the Zhalpak mine in Kazakhstan, and a joint venture with Kazatomprom was set up to develop it.

Sinosteel Corporation holds minor equity in explorer PepinNini Minerals Ltd in Australia and has 60% of a joint venture with PepinNini to develop a uranium deposit in South Australia. Sinosteel is also involved with exploration on Quebec and Krygystan.

In 2010 CNNC contracted with Cameco for 8865 tU through to 2025.

In March 2009, CNNC International, a 70% subsidiary of CNNC Overseas Uranium Holding Ltd and through it, of CNUC, agreed on a $25 million takeover of Western Prospector Group Ltd which controlled the Gurvanbulag deposit in Mongolia, very close to the Chinese border. Western Prospector and its Mongolian subsidiary, Emeelt Mines, undertook a definitive feasibility study which showed that the project was barely economic, on the basis of 6900 tU reserves averaging 0.137% U. With radiometric sorting the head grade would be 0.152%U and the mine could produce 700 tU/yr for nine years. Mine development cost would be about $280 million. In June 2012 CNNC Mongolia Project Co announced an agreement with the Mongolian government's Nuclear Energy Agency (NEA) to develop Gurvanbulag, following three years of feasibility studies and preparation. In August 2014 CNNC said that the government had approved its recent feasibility study, and negotiations towards a joint venture company with NEA continued. In November 2015 CNNC International reported that all exploration work was complete and that it awaited a mining licence. It then expected to form a project JV with the government, which would hold 51% leaving CNNC International Ltd. with 49% of the project.

CNNC has been searching for uranium in Jordan.

CGN initiatives abroad

CGN subsidiary China Guangdong Nuclear Uranium Resources Co Ltd (CGN-URC) set up in 2006 has uranium imports and investment in overseas sources of supply as part of its remit. It has been active in securing foreign supplies of uranium.

In September 2007, two agreements were signed in Beijing between Kazatomprom and CGN on Chinese participation in Kazakh uranium mining joint ventures and on reciprocal Kazatomprom investment in China's nuclear power industry. These came in the context of an earlier strategic cooperation agreement and one on uranium supply and fuel fabrication. This is a major strategic arrangement for both companies, with Kazatomprom to become a major uranium and nuclear fuel supplier to CGN. A framework strategic cooperation agreement was then signed with CNNC. A CGN subsidiary, Sino-Kazakhstan Uranium Resources Investment Co, has invested in two Kazakh uranium mines: Irkol and Semizbai, while CNNC is investing in another: Zhalpak. In April 2015 CGN Mining Co Ltd purchased the Sino-Kazakh shares, so it now holds 49% of the Semizbai-U JV.

In November 2010 CGN signed a long-term contract with Kazatomprom for 24,200 tonnes of uranium through to 2020. In May 2014 CGN contracted with Uzbekistan’s Navoi Mining & Metallurgy for $800 million worth of uranium to 2021. In 2013 Uzbekistan exported 1,663 tonnes of uranium (U₃O₈?) to China.

In November 2007 CGN signed an agreement with Areva to take a 24.5% equity stake in its UraMin subsidiary (now Areva Resources Southern Africa), and for China to take half the output, but this did not proceed.*

* Uramin was proposing mines in Namibia, South Africa and Central African Republic. In October 2008, Areva announced that a further 24.5% would be taken up by other 'Chinese sovereign funds', though it would remain the operator. China also agreed to buy more than half of the uranium from UraMin over the lifetime of the three deposits – the total quantity involved was to be over 40,000 tU to 2022. However, production from those mines, Trekkopje, Ryst Kuil and Bakouma respectively has not yet materialized, and at the end of 2009 Areva’s reported 100% equity in the company, with no Chinese equity.

CGN-URC has embarked upon a 50-50 joint venture with Uzbekistan's Goskomgeo focused on black shales the Sino-Uz Uranium Resources Co Ltd (or Uz-China Uran LLC), in particular the Boztau uranium exploration project in the Central Kyzylkum desert of the Navoi region of Uzbekistan. Over 2011-13 CGN-URC was to develop technology for the separate production of uranium and vanadium from these black shale deposits with a view to commencing production from 2014. In May 2014 Goskomgeo said CGN-URC planned to start mining in 2014, with production being sold to China.

In 2012 CGN-URC, through a Hong Kong subsidiary Taurus Minerals (60% CGN, 40% China-Africa Development Fund), took over Kalahari Minerals PLC and then Extract Resources Ltd, giving it ownership of the massive Husab project in Namibia, with 137,700 tU measured and indicated resources and a further 50,000 tU inferred resources at Rossing South. The cost was about $2.2 billion. Swakop Uranium is the development company owned by Taurus, except for a 10% share held by the government’s Epangelo Minerals. Mine development commenced in April 2013, production commenced at the end of 2016, ramping up to 5500 tU/yr. In July 2014 CGN Global Uranium Ltd (CGU) was incorporated in the UK to sell Husab uranium on the world market, though most production will be for CGN.

In 2015 CGN paid C$82.2 million for a nearly 20% stake in Fission Uranium Corp, making it the first direct Chinese investment in a Canadian uranium developer. An offtake agreement will entitle CGN to up to 35% of Patterson Lake South production at a 5% discount on prevailing spot market prices.

In mid-2010, CGN signed a framework agreement with Cameco under which the two companies would negotiate long-term uranium purchase agreements and potential joint development of uranium resources. In November the Cameco sale of 11,200 tonnes of uranium through to 2025 was confirmed. Then, in November 2010, CGNPC signed a $3.5 billion, ten-year contract with Areva for supply of 20,000 tonnes of uranium.

Alternative sources of uranium

In 2007 CNNC commissioned Sparton Resources of Canada with the Beijing No.5 Testing Institute to undertake advanced trials on leaching uranium from coal ash out of the Xiaolongtang power station in Yunnan province, in the southwest. The Lincang ash contains 160-180 ppm U – above the cut-off level for some uranium mines. The power station ash heap contains over 1700 tU, with annual arisings of 106 tU. Two other nearby power stations burn lignite from the same mine. A joint venture company Yunnan Sparton New Environ Tech Consulting Co. Ltd. (SNET), 60% owned by Sparton, has been set up to operate the secondary recovery programs. No results were evident by mid-2011, or since.

Nuclear fuel industrial parks

Two industrial parks for nuclear fuel are planned – a northern one in Hebei near Beijing, and one in the south, probably Guangdong province. They will each include uranium conversion, enrichment, and fabrication facilities to support China's goal to become the centre of Asian nuclear fuel preparation and manufacturing. Also, sales of Hualong reactors are envisaged as being with a supply of fuel. About CNY 80 billion is being invested in the two parks.

In May 2013 CGN and CNNC announced that their new China Nuclear Fuel Element Co (CNFEC) joint venture would build a CNY 45 billion ($7.33 billion) complex in Daying Industrial Park at Zishan town in Heshan and Jiangmen city, Guangdong province. This nuclear fuel industrial park was to be established during the 12th Five-Year Plan and be fully operational by 2020. However, in July 2013 the plan was abruptly stalled. The 200 ha park was to involve 1000 tU/yr fuel fabrication as well as a conversion plant (14,000 t/yr) and an enrichment plant, close to CGN’s Taishan power plant.

The plan is being implemented at Cangzhou in Hebei province – the North Park – and a new site being sought in Guangdong – the South Park.

Conversion

Information on China’s conversion capacity is uncertain. The World Nuclear Association's (WNA's) 2017 Nuclear Fuel Report has 8800 tU/yr as reference case requirements for 2018, rising to 9660 tU/yr in 2020 and 16,000 tU/yr in 2025.

Conversion requirements are of the same order. A conversion plant at Lanzhou in Gansu province of about 1000 tU/yr started operation in 1980 but may now be closed. A 5000 t/yr plant is reported there, operating at about 80% capacity, and a 9000 t/yr one is reported as under construction and due on line in 2017 or 2018.

Another conversion plant at Diwopu, Jiuquan, near Yumen in northwest Gansu province, is run by CNNC 404 company and is about 500 tU/yr.

China Nuclear Fuel Corp is building a plant at Hengyang in Hunan province. UxC quotes this as 3000 tU/yr, with construction permit issued in October 2014 and operation expected in 2018. The head of Kazatomprom visited this in mid-2016.

New conversion capacity was proposed with the new China Nuclear Fuel Element Co (CNFEC) plant at Daying Industrial Park in Heshan and Jiangmen city, Guangdong province. It was quoted at 14,000 t/yr by 2020 but plans for this location were cancelled in July 2013. The new location of the complex is Cangzhou in Hebei province in the north, due to commence production in 2018 and ramp up to full capacity after 2020. However, the Guangdong government wants to revive the original project and CNNC is looking for a southern site for part of the capacity.

Ux Consulting comments that if all these plans are realized by 2030, China will have a total conversion capacity about 31,000 tU/year, which can match an enrichment capacity about 23 million SWU/year, sufficient to feed about 180 GWe of PWR capacity.

Enrichment and enriched uranium imports

In 2010 China needed 3600 tU and 2.5 million SWU of enrichment. The WNA Nuclear Fuel Report has demand in 2020 at 15,000 tU (natural) and about 8 million SWU. Enrichment requirements rise to about 13 million SWU in 2025 and 19.6 million SWU in 2030. All enrichment capacity is inland, in Shaanxi and Gansu provinces. China aims for a fully independent enrichment capability including R&D, engineering, manufacturing and operating.

A Russian centrifuge enrichment plant at Hanzhun/Hanzhong, SE Shaanxi province, was set up under 1992, 1993 and 1996 agreements between Minatom/Tenex and CNEIC covering a total 1.5 million SWU/yr capacity in China at two sites. The first two modules at Hanzhun came into operation in 1997-2000, giving 0.5 million SWU/yr as phases 1 & 2 of the agreements. In November 2007, Tenex undertook to build a further 0.5 million SWU/yr of capacity at Hanzhun, completing the 1990s agreements in relation to the Hanzhun plant. This was commissioned ahead of schedule in mid-2011 and has operated reliably since.

A north expansion project at Hanzhun was then built over 2012-14, with 1.2 million SWU/yr capacity using indigenous technology.

The full agreement for the main $1 billion Hanzhun plant was signed in May 2008 between Tenex (Techsnabexport) and China Nuclear Energy Industry Corporation. The site, or at least two phases of it, is under IAEA safeguards. Up to 2001 China was a major customer for Russian 6th generation centrifuges, and more of these were supplied in 2009-10 for Hanzhun, under phase 4 of the agreement.

The Lanzhou enrichment plant in Gansu province to the west started in 1964 for military use and operated commercially 1980 to 1997 using Soviet-era diffusion technology. A Russian centrifuge plant of 500,000 SWU/yr started operation there in 2001 as phase 3 of the above agreements and it replaced the diffusion capacity. Subsequent expansion is based on indigenous centrifuge technology, about 2.5 million SWU of which was operating in mid-2015. Two 0.5 million SWU units (CEP 2&3) and one 1.2 million SWU unit (CEP 4) comprise the indigenous additional capacity. CEP 4 is due to start full commercial operation in 2016.

Another and larger diffusion enrichment plant, Plant 814, operated at Heping, Sichuan province, from 1975 to 1987 for military purposes. It was indigenously built, about 200-250,000 SWU/yr capacity, but its continued operational status and purpose is uncertain, possibly including fuel for naval reactors. It appears to have been upgraded about 2006. A new 0.8 million SWU/yr centrifuge plant was then built at Emeishan nearby, operating from about 2013. A second 0.8 million SWU/yr plant is under construction there.

China Uranium Enrichment Capacity

Plant	Province	Annual capacity (million SWU) 2015	Projected capacity 2020
Hanzhun	Shaanxi	2.2	2.2
Lanzhou	Gansu	3.56	6.52
Heping 814	Sichuan	0.4 (uncertain)	0.4
Emeishan	Sichuan	0.8 (part built)	1.6 - 2.4
Estimated total		5.7 - 7.0	10.7 - 12.0

Sources: World Nuclear Association Nuclear Fuel Report, September 2015; Harvard Kennedy School study, August 2015 .
UxC estimates 2015 capacity at 4.5 million SWU.

China has developed its own centrifuge technology at Lanzhou, and the first domestically-produced centrifuge was commissioned there in February 2013. An estimated 4.1 million SWU capacity has been built using indigenous Chinese centrifuge technology.

Further enrichment capacity was planned with the new China Nuclear Fuel Element Co (CNFEC) plant at Daying Industrial Park in Heshan city, Guangdong province. It was quoted at 7 million SWU/yr by 2020. However plans for this location were cancelled in July 2013. The new location of the complex is Cangzhou in Hebei province in the north. It is due to commence production in 2018 and ramp up to full capacity after 2020. The Guangdong government wants to revive the original project, and a southern site is being sought.

CGN-URC contracts fuel fabrication services from CNEIC on behalf of its operational power generation companies. There has been some minor export of enrichment services, and in April 2014 a new initiative was reported, and export delivery of 1 million SWU was estimated for 2014 (unconfirmed).

Enriched uranium

Much of the enriched uranium for China's reactors has come from outside the country, and some still does so in connection with early fuel loads for foreign-sourced reactors.

A contract with Urenco supplied 30% of the enrichment for Daya Bay from Europe.

Under the May 2008 enrichment agreement Tenex is to supply (from Russia) 6 million SWU as low-enriched uranium product from 2010 to 2021 for the first four AP1000 reactors, this apparently being related to completion of the Hanzhun enrichment plant. It is expected to involve $5 to 7 billion of LEU and possibly more. Enriched uranium for the first four AP1000 reactors is being supplied by Tenex from Russia, under the 2008 agreement.

Fuel fabrication

CNNC is responsible for fuel fabrication in China, utilising some technology transferred from Areva, Westinghouse and TVEL. Fuel fabrication plants are inland, in Sichuan and Inner Mongolia. Demand in 2013 was about 1300 tU in fabricated fuel, and by 2020 this will rise to about 1800 tU – though precise levels fluctuate due to demand for initial core loads in new reactors.

Fuel Fabrication in and for China

Location	Company	Type of fuel	Capacity
Yibin, Sichuan	Jianzhong Nuclear Fuel, China Nuclear Fuel South	PWR VVER	800 t/yr 100 t/yr
Baotou, Inner Mongolia	China Nuclear Fuel North	PHWR PWR	200 t/yr 200 t/yr total 800 t/yr by 2020
Baotou, Inner Mongolia	CNNC Baotou Nuclear Fuel Company Ltd	AP1000, CAP etc	800 t/yr
Baotou, Inner Mongolia	INET?	HTR	300,000 fuel pebbles
Oskomen, Kazakhstan	Ulba-FA (CGN & Kazatomprom)	PWR	200 t/yr

CNNC's main PWR fuel fabrication plant at Yibin, Sichuan province, was set up in 1982 (though based on a 1965 military plant) to supply Qinshan 1. It is operated by CNNC subsidiary China Jianzhong Nuclear Fuel (JNF), with its subsidiary China Nuclear Fuel South, and by October 2008 was producing fuel assemblies with 400 tU/yr. It reached 800 tU/yr of PWR fuel and 100 tU/yr VVER fuel* by the end of 2013 and plans indicate at least 1000 tU/yr by 2020. It supplies Qinshan, Tianwan, Fuqing, Ningde, Hongyanhe and Yangjiang and is contracted to produce Hualong One fuel for Fuqing 5&6. It has started producing the locally-designed CF3 fuel assemblies, with the first loaded in Qinshan II-2 CNP-600 in mid-2014. It has certified Kazakhstan's Ulba Metallurgical Plant as a source of pellets.

* VVER fuel fabrication at Yibin began in 2009, using technology transferred from TVEL under the fuel supply contract for Tianwan. (First core and three reloads for Tianwan 1&2 were from Novosibirsk Chemical Concentrate Plant in Russia – 638 fuel assemblies, under the main contract.) By August 2010, Yibin had produced 54 VVER-1000 fuel assemblies which were being loaded into the Tianwan 1 & 2 units. In November 2010, TVEL contracted with Jiangsu Nuclear Power Corporation (JNPC) and the China Nuclear Energy Industry Corporation (CNEIC) to supply six fuel reloads for Tianwan 1, and the technology for fuel to be produced at Yibin thereafter, for about US$ 500 million. TVEL certified the plant to manufacture the new TVS-2M fuel for Tianwan in April 2014. The two units run on 18-month refueling cycles.

CNNC set up a second civil fuel fabrication plant run by China North Nuclear Fuel Co Ltd (CNNFC) at Baotou, Inner Mongolia, in 1998, based on a military plant there dating from 1956. This has become a major R&D base, producing the most types of fuel. It fabricates fuel assemblies for Qinshan's CANDU PHWRs (200 tU/yr) and 200 tU/yr for older PWRs. Areva has assisted the plant to qualify for production of modern fuel. In 2012 the plant became the Northern Branch of China Nuclear Fuel Element Co Ltd., or simply China Nuclear Northern, though the original company name continued being used. The plant is ramping up from 400 tU/yr to 800 tU/yr production over 2013 to 2020 for PWR and PHWR fuel.

In 2008 SNPTC agreed with both fuel companies (Jianzhong and Northern) to set up CNNC Baotou Nuclear Fuel Company Ltd to make fuel assemblies for China's AP1000 reactors (first cores and some re-loads of the initial units are supplied by Westinghouse). In January 2011 a $35 million contract was signed with Westinghouse "to design, manufacture and install fuel fabrication equipment that will enable China to manufacture fuel" for AP1000 units. This production line was commissioned in 2015, at 400 tU/yr for phase 1. Phase 2, also 400 tU/yr, was completed in October 2016 and commissioned in June 2017. CNNC signed a contract with Sanmen Nuclear Power Company in January 2017 to provide refuelling components for the second, third and fourth operating cycles of units 1&2 of the Sanmen plant. Early in 2016 a prototype fuel assembly for the CAP1400 was produced. In 2015, now as a subsidiary of SPI, SNPTC declared a strong interest in pellet and fuel manufacturing for its AP and CAP reactors, with associated R&D.

A new fuel production line at Baotou to make the 9% enriched fuel spheres for the Shidaowan HTR-PM high temperature reactors in Shandong province was completed in May 2015 at a cost of almost CNY 300 million. It started production in March 2016, ramping up to a capacity of 300,000 fuel pebbles per year. By July 2017 it had produced 200,000, transitioning to full commercial operation. NNSA licensed the project in February 2013. It is based on a trial production line developed by INET at Tsinghua University to produce 100,000 spherical fuel elements per year, and INET is involved in the new plant. (In March 2011 a contract was signed with SGL Group in Germany for supply of 500,000 machined graphite spheres for HTR-PM fuel load by the end of 2013.) Qualification irradiation tests of fuel elements were completed in December 2014 at the High Flux Reactor at Petten in the Netherlands.

In May 2013 CNNC and CGN announced that they would build a new China Nuclear Fuel Element Co (CNFEC) plant at Daying Industrial Park in Heshan and Jiangmen city, Guangdong province, and it would have 1000 tU/yr capacity by 2020. Construction was due to start in 2013, but in July the plan for this location was abruptly cancelled. Some of this capacity is likely to be built at Cangzhou in Hebei province. The CNY 45 billion industrial park was also to involve a conversion plant and an enrichment plant.

CGN in joint venture with Kazatomprom is building the 200 t/yr Ulba-FA fuel fabrication plant in Kazakhstan to produce AFA 3G fuel assemblies for its French-designed reactors. CGN-URC holds 49% of the Ulba-FA JV.

In order meet its goal of being self-sufficient in nuclear fuel supply, additional fuel production capacity will be required. However, the EPR fuel for Taishan being supplied to CGN by Framatome, comprising the two first cores and 17 reloads, will be fabricated in France.

Also the VVER fuel for Tianwan 3&4 is being supplied by Russia’s TVEL until 2025, and it will help to equip the Yibin plant to produce from then, under a $1 billion contract with Jiangsu Nuclear Power Corporation (JNPC) and the China Nuclear Energy Industry Corporation (CNEIC).* In 2018 TVEL and CNEIC were considering joint construction of a VVER fuel fabrication plant in Ukraine.

* TVEL's TVS-2M fuel offers the possibility of an extended 18-month operating cycle and is used in Russia's Balakovo and Rostov power plants. After pilot operations using six TVS-2M assemblies at Tianwan 1, the design was licensed in China, and Tianwan 1&2 were converted to 18-month operating cycles from 2014. Units 3&4 are to run on the fuel from their first core loadings onwards.

CGN-URC contracts fuel fabrication services from CNFSC and CNFNC, and retails these to its operational power generation companies.

CNNC and Areva have set up a 50-50 joint venture to produce and market zirconium alloy tubes for nuclear fuel assemblies. The joint venture, CNNC Areva Shanghai Tubing Co. (CAST), started production at the end of 2012, and was expected to ramp up from 300 km of tubes per year to 1500 km in 2015, supplying both Yibin and Baotou fuel fabrication plants. A further agreement in 2013 may extend this JV to producing the zircaloy itself, at 600 t/yr by 2017.

Fuel cycles

A standard 18-month fuel cycle is the normal routine for Daya Bay, Ling Ao, and early M310 to CPR-1000 reactors. This has average burn-up of 43 GWd/t, with maximum of 50 GWd/t. An Advanced Fuel Management cycle using fuel with gadolinium burnable poison is implemented at Ling Ao phase II, Hongyanhe, Ningde, and Yangjiang, giving average 50 GWd/t and maximum 57 GWd/t through to CPR-1000+.

Moving to recycling fuel in PWRs is the next step, though with limited advantage compared with longer-term goal of using fast neutron reactors with MOX and advanced reprocessing. This will be electrometallurgical reprocessing (pyroprocessing) coupled with some sort of partitioning.

Reprocessing, recycling

Establishing significant reprocessing capacity is seen as vital both for management of domestic used fuel and as a service export in connection with selling reactors overseas.

Late in 2015 CNNC said it estimated that 23,500 tonnes of used fuel will have been discharged from reactors by 2030 and 15,000 tonnes of that would be in dry cask storage. When 50 GWe is operating, up to 2000 t used fuel will be discharged each year.

Most of the civil back-end facilities are currently in Gansu province.

CNNC Ruineng Technology Co Ltd was set up by CNNC in November 2011 to industrialise used fuel reprocessing technology and mixed oxide (MOX) fuel production to close the fuel cycle. This will involve both local initiatives and the planned Areva plant. It will also be responsible for storage and management of used fuel.

A pilot reprocessing plant using the Purex process was constructed from 2006 at Lanzhou Nuclear Fuel Complex in Gansu province. It completed hot commissioning in 2010 to reprocess about 50 tonnes of used fuel over 2013-15. A demonstration used fuel treatment plant, with capacity of 200 tonnes of used fuel per year, is being built in Gansu Nuclear Technology Industrial Park in Gansu province by CNNC Longrui Technology Company, which was set up in March 2015.

A large (800-1000 t/yr) commercial reprocessing plant based on indigenous advanced technology was planned to follow and begin operation about 2020, but the 800 t/yr Areva project will apparently displace it. A second 800 t/yr plant was to follow.

The planned Areva (now Orano) reprocessing plant was first mooted in November 2007, when Areva and CNNC signed an agreement to assess the feasibility of a reprocessing plant for used fuel and a mixed oxide (MOX) fuel fabrication plant in China, representing an investment estimated then of €15 billion. The 800 t/yr reprocessing plant was then envisaged to be in Jinta county, north of Jiayuguan in Gansu province, employing proven French technology and operated by Areva. Design, construction and commissioning was expected to take ten years from 2010. In November 2010, an industrial agreement on this was signed, which Areva said was "the final step towards a commercial contract" for the project. In April 2013 a further agreement was signed with Areva, setting out the technical specifications for the 800 t/yr plant. Then in March 2014 another agreement on the matter was signed, to continue planning the project and completing a business case for it.

In June 2015 a further agreement "formalizes the end of technical discussions, defines the schedule for commercial negotiations and confirms the willingness of both groups to finalize the negotiations in the shortest possible timeframe," according to Areva. In September CNNC said that it was selecting a site and that construction was expected to start in 2020 to be completed in 2030. China Nuclear Power Engineering Corporation (CNPE) was seeking seismic studies of coastal sites for the plant on behalf of CNNC Ruineng Technology Co Ltd. In November 2015 another Areva-CNNC agreement was signed for the 800 t/yr plant, referencing Orano’s La Hague plant and envisaging a cost of CNY 100 billion ($15.7 billion). CNNC would be responsible for building the plant but with Areva NC (now Orano) responsible for technical aspects.

A coastal site in Jiangsu province was suggested, so that used fuel could be transported on ships (rather than a 3000 km road or rail trip inland to Gansu, though public and local government support there is strong). In July 2016 Lianyungang city in Jiangsu was mentioned as likely for the 3km² site, close to the Tianwan nuclear power plant, but in August public protests caused local government to back away from the proposal. A final contract had been envisaged in 2017 with construction from 2020, possibly including a MOX plant.

In addition to the reprocessing plant, the site will also house a used fuel storage facility with the capacity to hold 3000 tonnes of fuel initally, and possibly 6000 tonnes eventually. An associated high-level liquid waste vitrification facility is also planned.

It is estimated that another 800 t/yr plant will be required every ten years to match nuclear growth.

The China Institute of Atomic Energy (CIAE) envisaged an industrial reprocessing plant of about 1000 t/yr in operation from about 2021.

Technology for recycling uranium recovered from used nuclear fuel from Chinese PWRs for use in the Qinshan Phase III Candu units is being developed (see section below on Recycled uranium in PHWRs; thorium in PHWRs).

Mixed oxide fuel

A small experimental mixed oxide (MOX) fuel plant was built in 2008, giving experience at 500 kg/yr.

In October 2010, GDF Suez Belgian subsidiary Tractabel, with Belgonucleaire and the nuclear research centre SCK-CEN signed an agreement with CNNC to build a pilot mixed oxide (MOX) fuel fabrication plant in China. Belgium has experience in MOX fuel development and production dating back to 1960, including 20 years of industrial MOX production at Belgonucleaire's 35 tonne per year Dessel plant from 1986 to 2006 (see section on Fuel cycle in the information page on Nuclear Power in Belgium). MOX has been in use in Belgium's nuclear power plants since 1995.

Fuel for the BN-800 reactors (referred to as Chinese Demonstration Fast Reactors – see section below on Fast neutron reactors) earlier planned to be built at Sanming would be MOX pellets, initially made in Russia.

CIAE shows two 40 t/yr MOX fabrication plants in operation from about 2018. A 50 t/yr MOX reprocessing plant is under consideration for operation by 2030. This may be part of the Areva NC (Orano) reprocessing plant contract. Ux Consulting suggests a MOX fuel plant in Gansu province by about 2020.

Waste management

When China started to develop nuclear power, a closed fuel cycle strategy was also formulated and declared at an International Atomic Energy Agency conference in 1987. The used fuel activities involve: at-reactor storage; away-from-reactor storage; and reprocessing. CNNC has drafted a state regulation on civil used fuel treatment as the basis for a long-term government programme. There is a levy of CNY 2.6 cents/kWh from the fifth year of commercial operation of each reactor, to pay for used fuel management, reprocessing, and the eventual disposal of separated HLW.

Based on expected installed capacity of 50 GWe by 2020, the annual used fuel arisings will amount to about 1,300 tonnes at that stage, the cumulative total being about 14,000 tonnes then. The two Qinshan Phase III CANDU units, with lower burn-up, will discharge 176 tonnes of used fuel annually.

Storage of used fuel and disposal of HLW

A centralised used fuel storage facility has been built at Lanzhou Nuclear Fuel Complex, 25 km northeast of Lanzhou in central Gansu province. The initial stage of that project has a storage capacity of 550 tons and could be doubled. However, most used fuel is stored at reactor sites, in ponds. It or an intermediate-level waste repository there is 10-20 m underground. The only dry storage operating is at Qinshan, and this is being expanded.

CNNC Everclean Co Ltd is responsible for used fuel transport from nuclear power plant sites to Lanzhou Nuclear Fuel Complex, and storage there. Some used fuel – over 100 fuel assemblies per year – is transported 3700 km by road from Daya Bay to Gansu province for storage. According to the State Administration for Science, Technology and National Defense Industry (SASTIND), this quantity needs to increase to 600 assemblies per year. In June 2018 CNNC Everclean contracted with Holtec International to supply its HI-STAR 100MB casks by 2020. In 2016 it had bought four NAC-STC casks for high-burnup fuel, and in January 2018 Spain's ENSA also supplied a cask.

Following reprocessing, separated high-level waste will be vitrified, encapsulated and put into a geological repository some 500 metres deep. Site selection and evaluation has been under way since 1986 and is focused on three candidate locations in the Beishan area of Gansu province and will be completed by 2020. All are in granite. An underground research laboratory was to be built 2015-20 and operate for 20 years. The third step is to construct the final repository from 2040 and to carry out demonstration disposal. Acceptance of high-level waste into a national repository is anticipated from 2050. All this is taking place under the 2006 R&D Guidelines for Geological Disposal jointly published by China Atomic Energy Authority, Ministry of Science &Technology, and Ministry of Environmental Protection.

In mid-2014 construction started on a vitrification plant for HLW in Sichuan, where 800 m^₃ of liquid waste was reported to be stored already. It will use German technology and plant from Karlsruhe Institute of Technology. It is entirely for military waste, but the technology may be usable later for civil waste.

The regulatory authorities of high-level radioactive waste disposal projects are Ministry of Environmental Protection (MEP) and the National Nuclear Safety Administration (NNSA). The China Atomic Energy Agency (CAEA) is in charge of the project control and financial management. CNNC deals with implementation, and four CNNC subsidiaries are key players: Beijing Research Institute of Uranium Geology (BRIUG) handles site investigation and evaluation, engineered barrier study and performance analyses, with the China Institute of Atomic Energy (CIAE) undertaking radionuclide migration studies. The China Institute for Radiation Protection (CIRP) is responsible for safety assessment, and the China Nuclear Power Engineering Company (CNPE) works on engineering design.

Low- and intermediate-level waste

Industrial scale disposal of low- and intermediate-level waste (LILW) is at three sites: near Yumen, northwest Gansu province; at the Beilong repository in Gunagdong province near the Daya Bay nuclear plant; and at Feifengshan, Sichuan province. Two of these are run by CNNC Everclean Co, the other by a CGN subsidiary. These are the first three of five planned regional LILW disposal facilities.

China LILW disposal sites

Site name	Location	Operator	Storage capacity planned	Storage capacity actual	Functions
Xibei Center, Northwest	Gansu province (CNNC 404 plant)	CNNC Everclean Co.	200,000 m^₃	20,000 m³	Includes waste from local military facilities.
Beilong Center	Guangdong province	Guangdong Daya Bay Nuclear Power Environmental Protection Co.	80,000 m³	8800 m³	5 km away from Daya Bay. Dedicated to waste from Daya Bay and Ling Ao units.
Feifeng Mountain	Sichuan province	CNNC Everclean Co.	180,000 m³	20,000 m³	As a testing facility.

Decommissioning

The China Institute of Atomic Energy (CIEA) 15 MWt HWRR-II heavy water research reactor that started up in 1958 was shut down at the end of 2007 and decommissioned.

Heavy manufacturing industrial parks

Two significant industrial parks focused on nuclear manufacturing were announced in 2010 and are being set up.

The first is a nuclear technology base near Nanjing in Jiangsu province, known as the Nanjing Jiangning district Binjiang Development Zone, and part of the China Nuclear Binjiang Production Base inaugurated in 2003 which includes a research facility for nuclear-grade concrete. China Huaxing Nuclear Construction Company (HXCC) committed to build this on the banks of the Yangtze River about 300 km west of Shanghai, in three phases to 2015. Nanjing is a transport hub, and the overall 51 square kilometre development zone will be served by a new river port including a bulk cargo terminal and 12 deep-water piers.

The zone will feature as its centrepiece a $146 million factory for pre-assembled structural and equipment modules for CPR-1000 and Westinghouse AP1000 reactors. The modules, weighing up to nearly 1000 tonnes each in the case of AP1000, can then be taken by barge to construction sites. Currently AP1000 modules are made by Shandong Nuclear Power Equipment Manufacturing Co. which has the capacity to support construction of two reactors per year. HXCC is the main civil engineering contractor for China Guangdong Group.

The second is the China Haiyan Nuclear Power City, launched by CNNC at Haiyan, Zhejiang province, on the Yangtze delta about 120 km southwest of Shanghai and close to the cities of Hangzhou, Suzhou and Ningbo. As well as having the nuclear power plants in the Qinshan complex nearby, Haiyan hosts the headquarters of 18 leading Chinese nuclear equipment suppliers and branch offices of all the major Chinese nuclear design institutes and construction companies. The new China Haiyan Nuclear Power City will cover 130 square kilometers and has a 10-year budget of $175 billion, according to reports. It is expected to have four main areas of work: development of the nuclear power equipment manufacturing industry; nuclear training and education; applied nuclear science industries (medical, agricultural, radiation detection and tracing); and promotion of the nuclear industry.

The Haiyan Nuclear Power City is entitled to all the preferential benefits granted to national economic and technological zones and national hi-tech industrial zones. Enterprises in the industrial park will enjoy priority for bidding quota, bidding training, qualification guidance and specific purchasing with CNNC. The concept is based on the French equivalent in the Burgundy area, and French suppliers will be involved at Haiyan, as will CGNPC.

As well as these major industry centres there is a factory for AP1000 modules set up at Haiyang, on the coast, and another in central Hubei province to support inland AP1000 projects and later the CAP-1400 derivatives.

A further centre, the Taishan Clean Energy (Nuclear Power) Equipment Industrial Park, opened in February 2010 in the Pearl River Delta region of Guangdong province, and is expected to become a centre for nuclear power equipment manufacturing, initially supplying hardware and services to nearby nuclear power projects. The planned development will eventually cover about 45 sq km and include design, R&D and technical services. The initial 3.1 sq km phase of the park costing CNY 2 billion was followed by a second 2.4 sq km phase. Targets call for manufacturers at the park to have 45% of the nuclear equipment market in Guangdong and produce goods worth CNY 22 billion by 2020 while playing a leading role in R&D and maintenance of nuclear power equipment. The park also plans to produce CNY 20 billion in goods not related to the power industry by 2020.

Research & development

Initial Chinese nuclear R&D was military. A water-cooled graphite-moderated production reactor for military plutonium started operating in 1966, located at the Jiuquan Atomic Energy Complex some 100 km northwest of the city of Jiuquan in Gansu province, north-central China. The area is mainly desert and very remote. In the early 1980s it was decided to convert it to dual-use, and plutonium production evidently ceased in 1984. Reprocessing was onsite. Another, larger, plutonium production reactor with associated facilities was in a steep valley at Guangyuan in Sichuan province, about 1000 km south. It started up about 1975 and produced the major part of China's military plutonium through to 1991.

In November 2013 China National Nuclear Power Company, Ltd. (CNNP) joined two of the nuclear-related research programs run by the Electric Power Research Institute (EPRI) in the USA. These are the Nuclear Maintenance Application Center (NAMC), which develops technologies, systems, and guides to drive improvements in nuclear plant maintenance activities; and the Nondestructive Evaluation (NDE) program, which develops technologies and procedures to quickly, accurately, and cost-effectively inspect and characterize nuclear component condition and inform strategic decisions on whether and when to replace, repair, or continue operation. CNNP said that it “will expand its engagement with EPRI soon to become a full member in all of its nuclear research programs.” Earlier in 2013 EPRI had signed agreements with CGN and SNERDI.

Research Reactors

Apart from military facilities, China has about 19 operational research reactors, and a report by the Ministry of Environmental Protection (MEP) in June 2013 asserted their good condition and safety, along with that of the country’s power reactors.

The 125 MW light water High-Flux Engineering Test Reactor (HFETR) has been run by the (Southwest) Leshan Nuclear Power Institute of China at Jiajiang, Sichuan province, since 1979. Early in 2007, this was converted to use low-enriched uranium, with the help of the US National Nuclear Security Administration (NNSA). At least one of the five research reactors in Sichuan province was near the epicentre of the May 2008 earthquake.

The China Institute of Atomic Energy (CIEA) near Beijing undertakes fundamental research on nuclear science and technology and is the leading body in relation to fast neutron reactors, as well as other research reactors. Its 15 MWt HWRR-II heavy water research reactor started up in 1958 and was shut down at the end of 2007. An updated version of this was supplied to Algeria and has operated since 1992.

CIEA built the new 60 MWt China Advanced Research Reactor (CARR), a sophisticated and versatile light water tank type unit with heavy water reflector which started up in May 2010, reaching full power in March 2011, and it also built the 65 MW China Experimental Fast Reactor (CEFR) which started up in July 2010. (see subsection below on Fast neutron reactors).

In October 2010, the Belgian nuclear research centre SCK-CEN signed an agreement with the China Academy of Sciences to collaborate on the Belgian MYRRHA project^b, which China sees as a way forward in treating nuclear wastes.

Reactor and fuel cycle development

In 2008, SNPTC and Tsinghua University set up the State Research Centre for Nuclear Power Technology, focused on large-scale advanced PWR technology and to accelerate China's independent development of third-generation nuclear power.

A 200 MWt NHR-200 integral PWR design for heat and desalination has been developed by Tsinghua University's Institute of Nuclear Energy Technology (INET) near Beijing. It is developed from the 5 MW NHR-5 prototype which started up in 1989.

The NDRC is strongly supporting R&D on advanced fuel cycles, which will more effectively utilise uranium, and possibly also use thorium. The main research organisations are INET at Tsinghua University, China Institute of Atomic Energy (CIEA), also near Beijng, and the Nuclear Power Institute of China (NPIC) at Chengdu, which is the main body focused on the PHWR technology and fuel cycles. INET has been looking at a wide range of fuel cycle options including thorium, especially for the Qinshan Phase III PHWR units. NPIC has been looking at use of reprocessed uranium in Qinshan's PHWR reactors. CIAE is mainly involved with fast reactor R&D. China's R&D on fast neutron reactors started in 1964.

In November 2018 CNNC opened the new Research Centre for Nuclear Fuels and Materials in Beijing. CNNC said it marked “an important milestone in the development and production of high-performance nuclear fuels and materials, as well as high-performance nuclear reactor cores to realize the effective and efficient development of nuclear energy,” in an advanced nuclear science industrial system.

A report from NEA says that in Jiangxi a CNEC affiliate, Nuclear Construction Clean Energy Co. Ltd, has signed an agreement with Ruijin government to set up Jiangxi Ruijin Nuclear Power Preparatory Office. According to the agreement, Nuclear Construction Clean Energy Co would look for a site to construct a high temperature gas cooled reactor.

Thorium molten salt reactor programme

The China Academy of Sciences (CAS) in January 2011 launched a programme of R&D on thorium-breeding molten salt reactors (Th-MSR or TMSR), otherwise known as liquid fluoride thorium reactors (LFTRs), claiming to have the world's largest national effort on these and hoping to obtain full intellectual property rights on the technology. The unit they are building is said to be similar to the 7 MWt Oak Ridge test MSR which ran 1965-69 in the USA with U-235 then U-233 fuels. The timeline for full commercialisation of TMSR technology was originally 25 years, but is reported to have been dramatically shortened, which may be reflected in increased funding.

The TMSR Research Centre has a 5 MWe solid-fuel MSR prototype under construction at Shanghai Institute of Nuclear Applied Physics (SINAP, under the Academy) originally with 2015 target for operation, now 2020. This is also known as the fluoride salt-cooled high-temperature reactor (FHR) in Generation IV parlance, or Advanced HTR (AHTR). A 2 MWe accelerator-driven sub-critical liquid fuel prototype is also being developed at SINAP to demonstrate the thorium cycle.

In March 2016 a strategic cooperation agreement to develop accelerator-driven advanced nuclear energy systems (ADANES) was signed between China General Nuclear (CGN) and the Chinese Academy of Sciences (CAS).

SINAP has about 600 staff and 200 graduate students undertaking basic R&D on MSRs including that on molten salt manufacture and loop technology, R&D of the front end and back end of the Th-U fuel cycle, R&D of high-temperature durable materials, and R&D of safety standards and licensing. It is also establishing specifications for nuclear-grade ThF₄ and ThO₂ for MSRs. It has subcontracted some work on molten salt coolants to the Fangda Group. However, reports in late 2016 suggest that the availability of pure Li-7 is a constraint on progress. (See also information paper on Lithium.)

China plans for the TMSR-SF to be an energy solution for the northwest half of the country, with lower population density and little water. The application of water-free cooling in arid regions is envisaged from about 2025.

SINAP has two streams of MSR development – solid fuel (TRISO in pebbles or prisms/blocks) with once-through fuel cycle, and liquid fuel (dissolved in fluoride coolant) with reprocessing and recycle. A third stream of fast reactors to consume actinides from LWRs is planned. The aim is to develop both the thorium fuel cycle and non-electrical applications in a 20-30 year time frame.

The TMSR-SF stream has only partial utilization of thorium, relying on some breeding as with U-238, and needing fissile uranium input as well. It is optimized for high-temperature based hybrid nuclear energy applications. SINAP aimed at a 2 MW pilot plant (to be TMSR-SF1) initially, though this has been superseded by a simulator (TMSR-SF0) to be followed by a 10 MWt prototype (TMSR-SF1) before 2025. A 100 MWt demonstration pebble bed plant (TMSR-SF2) with open fuel cycle would follow, then a 1 GW demonstration plant (TMSR-SF3). TRISO particles will be with both low-enriched uranium and thorium, separately.
The TMSR-LF stream is claimed to use a fully closed Th-U fuel cycle with breeding of U-233 and much better sustainability with thorium but greater technical difficulty. It is optimized for utilization of thorium with electrometallurgical pyroprocessing. SINAP aims for a 2 MWt pilot plant (TMSR-LF1) initially, then a 10 MWt experimental reactor (TMSR-LF2) by 2025, and a 100 MWt demonstration plant (TMSR-LF3) with full electrometallurgical reprocessing by about 2035, followed by 1 a GW demonstration plant. The TMSR-LF timeline is about ten years behind the SF one.
A TMSFR-LF fast reactor optimized for burning minor actinides is to follow.

SINAP sees molten salt fuel being superior to the TRISO fuel in effectively unlimited burnup, less waste, and lower fabricating cost, but achieving lower temperatures (600°C+) than the TRISO fuel reactors (1200°C+). Near-term goals include preparing nuclear-grade ThF₄ and ThO₂ and testing them in a MSR.

The TMSR-SF program is proceeding with preliminary engineering design in cooperation with the Nuclear Power Institute of China (NPIC) and Shanghai Nuclear Engineering Research & Design Institute (SNERDI). Nickel-based alloys are being developed for structures, along with very fine-grained graphite.

Two methods of tritium stripping are being evaluated, and also tritium storage.

The 10 MWt TMSR-SF1 will have TRISO fuel in 60mm pebbles, similar to HTR-PM fuel, and deliver coolant at 650°C and low pressure. Primary coolant is FLiBe (with 99.99% Li-7) and secondary coolant is FLiNaK. Core height is 3 m, diameter 2.85 m, in a 7.8 m high and 3 m diameter pressure vessel. Residual heat removal is passive, by cavity cooling. A 20-year operating life is envisaged. The TMSR-SF0 simulator is one-third scale, with FLiNaK cooling and a 400 kW electric heater.

The 2 MWt TMSR-LF1 is only at the conceptual design stage, but it will use fuel enriched to under 20% U-235, have a thorium inventory of about 50 kg and conversion ratio of about 0.1. FLiBe with 99.95% Li-7 would be used, and fuel as UF₄. Residual heat removal is passive, by air cooling. The project would start on a batch basis with some online refuelling and removal of gaseous fission products, but discharging all fuel salt after 5-8 years for reprocessing and separation of fission products and minor actinides for storage. It would proceed to a continuous process of recycling salt, uranium and thorium, with online separation of fission products and minor actinides. It would work up from about 20% thorium fission to about 80%.

Beyond these, a 400 MWt/168 MWe liquid-fuel MSR small modular reactor is planned, with supercritical CO₂ cycle in a tertiary loop at 23 MPa using Brayton cycle, after a radioactive isolation secondary loop. Various applications as well as electricity generation are envisaged. It would be loaded with 15.7 tonnes of of thorium and 2.1 tonnes of uranium (19.75% enriched), with one kilogram of uranium added daily, and have 330 GWd/t burn-up with 30% of energy from thorium. Online refuelling would enable eight years of operation before shutdown, with the graphite moderator needing attention.

The US Department of Energy (especially Oak Ridge NL) is collaborating with the Academy on the program, which had a start-up budget of $350 million. Australia’s Nuclear Science & Technology Organisation (ANSTO) is also involved, along with the American Nuclear Society (ANS) on safety standards for the solid fuel TMSR, and with the American Society of Mechanical Engineers (ASME) on material processing standards.

Structural materials for MSRs must demonstrate strength at high temperatures, be radiation resistant and also withstand corrosion. Following joint research on alloys, ANSTO announced in February 2017 that NiMo-SiC alloys – prepared from nickel molybdenum metal powders with added silicon carbide particles – have superior corrosion resistance and radiation damage resistance. A number of NiMo-SiC alloy specimens containing varying amounts of silicon carbide were prepared in SINAP laboratories before being characterised at ANSTO. They possess superior mechanical properties owing to the precipitation, dispersion and solid-solution strengthening of the NiMo matrix. The strength of these alloys stems from the combination of dispersion strengthening by SiC particles, precipitation strengthening by Ni₃Si and solid-solution strengthening by Mo as NiMo.

The primary reason that American researchers and the China Academy of Sciences/ SINAP are working on solid fuel, salt-cooled reactor technology is that it is a realistic first step. The technical difficulty of using molten salts is significantly lower when they do not have the very high activity levels associated with them bearing the dissolved fuels and wastes. The experience gained with component design, operation, and maintenance with clean salts makes it much easier then to move on and consider the use of liquid fuels, while gaining several key advantages from the ability to operate reactors at low pressure and deliver higher temperatures.

Recycled uranium in PHWRs; thorium in PHWRs

Early in 2008, CNNC subsidiary the Nuclear Power Institute of China (NPIC) signed an agreement with Atomic Energy of Canada Ltd (AECL) to undertake research on advanced fuel cycle technologies such as recycling recovered uranium from used PWR fuel and Generation IV nuclear energy systems. The initial agrement developed into a strategic agreement among AECL (now Candu Energy), the Third Qinshan Nuclear Power Company (TQNPC), China North Nuclear Fuel Corporation and NPIC in November 2008, and subsequently one in July 2014 between Candu Energy’s parent company, SNC-Lavalin, and TQNPC and China North’s parent company, CNNC.

(Initially the project seemed to include DUPIC (direct use of used PWR fuel in Candu reactors), the main work on which so far has been in South Korea, but it differs in that here, plutonium is removed at the reprocessing plant for use in fast reactors – see Figure below.)

The partners jointly developed technology for recycling uranium recovered from used nuclear fuel from other Chinese reactors (PWRs) with up to 1.6% fissile content (but typically 0.9%) for use in the Qinshan Phase III Candu units. The first commercial demonstration of this was in unit 1 of Qinshan Phase III, using 12 fuel bundles with recycled uranium (RU/RepU) blended with depleted uranium (DU) to give natural uranium equivalent (NUE), similar to normal Candu fuel (0.7% U-235). It behaved the same as natural uranium fuel. Subject to supply from reprocessing plants, a full core of NUE was envisaged from 2014. Purchase of RU and DU, design and safety analysis, modification of fuel fabrication line, and licence application were planned in 2013. Full core implementation in both Candu reactors is expected in 2018.

In August 2012 a follow-on agreement among the parties (Candu Energy having taken over from AECL) focused on undertaking a detailed conceptual design of the Advanced Fuel Candu Reactor (AFCR), which is described as "a further evolution of the successful Candu 6 and Generation III Enhanced Candu 6 (EC6), optimized for use of recycled uranium and thorium fuel," or as “the fuel-flexible adaptation” of the EC6. This was directed towards "a pre-project agreement for two AFCR units in China, including site allocation and the definition of the licensing basis." Initially the two Qinshan units could be modified to become AFCRs (see above), and beyond that the first AFCR new build project is envisaged in China. One 700 MWe AFCR can be fully fuelled by the recycled uranium from four LWRs’ used fuel. Hence deployment of AFCRs will greatly reduce the task of managing used fuel and disposing of high-level wastes, and will reduce China’s fresh uranium requirements.

In November 2014 an expert panel hosted by the China Nuclear Energy Association (CNEA) made a positive recommendation on the AFCR, praising the reactor's safety characteristics and saying that AFCR technology forms a synergy with China’s existing PWRs and that it is positioned to “promote the development of closed fuel cycle technologies and industrial development,” which is consistent with the overall strategy of nuclear power development in China. Immediately following this, a framework joint venture agreement between CNNC and Candu Energy was signed to build AFCR projects domestically and develop opportunities for them internationally. In September 2016 an agreement among SNC-Lavalin, CNNC and Shanghai Electric Group (SEC) was to set up a joint venture in mid-2017 to develop, market and build the AFCR, with NUE fuel. Two design centres are envisaged, in China and Canada, to complete the AFCR technology. This could lead to the construction of two AFCR units in China.

The July 2014 agreement also provides a framework for collaboration between SNC-Lavalin and CNNC on uranium mining projects in China, “and the pursuit of international project opportunities in various high-growth sectors and markets.” This was supplemented by a November 2014 MoU between Natural Resources Canada and China’s NEA.

Phase one of the earlier AECL agreement was a joint feasibility study to examine the economic feasibility of utilizing thorium in the Qinshan Phase III PHWRs. (Geologically, China is better endowed with thorium than uranium.) This involved demonstration use of eight thorium oxide fuel pins in the middle of a Canflex fuel bundle with low-enriched uranium.

In July 2009, a second phase agreement was signed among these four parties to jointly develop and demonstrate the use of thorium fuel and to study the commercial and technical feasibility of its full-scale use in CANDU units. This was supported in December 2009 by an expert panel appointed by CNNC and comprising representatives from China’s leading nuclear academic, government, industry and R&D organizations. That panel also unanimously recommended that China consider building two new CANDU units to take advantage of the design's unique capabilities in utilizing alternative fuels. Like its 2014 successor, the expert panel comprised representatives from China’s leading nuclear academic, government, industry and R&D organizations. In particular it confirmed that thorium use in the Enhanced Candu 6 reactor design is “technically practical and feasible”, and cited the design’s “enhanced safety and good economics” as reasons it could be deployed in China in the near term.

In 2017, a thorium-optimised AFCR is envisaged for deployment about 2030.

Geologically, China is better endowed with thorium than uranium.

HTR demonstration: HTR-10

A 10 MWt high-temperature gas-cooled demonstration reactor (HTR-10) was commissioned in 2000 by the Institute of Nuclear Energy Technology (INET) at Tsinghua University near Beijing. It reached full power in 2003. It has TRISO fuel particles compacted with graphite moderator into 27,000 spherical fuel elements each 60mm diameter (as a pebble bed). Each fuel element contains 5g UO₂ enriched to 17%, and burn-up is 80 GWd/t. It has ten control rods in the graphite side reflector, with seven absorber ball units as secondary reactivity control, and passive heat removal. It reaches an outlet temperature of 700°C for the helium coolant at 3 MPa and may be used as a source of process heat for heavy oil recovery or coal gasification. It is similar to the South African PBMR (Pebble Bed Modular Reactor) intended for electricity generation.

In 2004, the reactor was subject to an extreme test of its safety when the helium circulator was deliberately shut off without the reactor being shut down. The temperature increased steadily, but the physics of the fuel meant that the reaction progressively diminished and eventually died away over three hours. At this stage a balance between decay heat in the core and heat dissipation through the steel reactor wall was achieved, the temperature never exceeded 1600°C, and there was no fuel failure. This was one of six safety demonstration tests conducted then.

Initially the HTR-10 has been equipped with a steam generator producing steam at 435°C coupled to a 2.5 MWe steam turbine power generation unit. However, second phase plans are for it to operate at 950°C and drive a gas turbine, as well as enabling R&D in heat application technologies. This phase will involve an international partnership with Korea Atomic Energy Research Institute (KAERI), focused particularly on hydrogen production.

Commercial HTRs: Shidaowan HTR-PM, Ruijin or Wan'an

A key R&D project is the demonstration Shidaowan HTR-PM of 210 MWe (two reactor modules, each of 250 MWt) which is being built at Shidaowan in Shandong province, driving a single steam turbine at about 40% thermal efficiency. The size was reduced to 250 MWt from earlier 458 MWt modules in order to retain the same core configuration as the prototype HTR-10 and avoid moving to an annular design like South Africa's PBMR.

Each reactor has a single steam generator with 19 elements (665 tubes) producing steam at 566°C. The fuel is 8.5% enriched (520,000 elements) giving 90 GWd/t discharge burn-up. Core outlet temperature is 750ºC for the helium, and steam temperature is 566°C. Core height is 11 metres, diameter 3 m. There are two independent reactivity control systems: the primary one is 24 control rods in the side graphite reflector, the secondary one six channels for small absorber spheres falling by gravity, also in the side reflector.

China Huaneng Group, one of China's major generators, is the lead organization in the consortium with China Nuclear Engineering & Construction Group (CNEC) and Tsinghua University's INET, which is the R&D leader. Chinergy Co. is the main contractor for the nuclear island. Projected cost is US$ 430 million, with the aim for later units being US$ 1500/kWe. The EPC contract was let in October 2008 and construction started in December 2012, with completion expected in 2017. The engineering of the key structures, systems, and components is based on Chinese capabilities, though they include completely new technical features. CNEC is the lead organisation regarding HTR technology.

The HTR-PM will pave the way for larger units based on the same module. The 600 MWe Ruijin units will effectively be three HTR-PMs. INET is in charge of R&D, and is aiming to increase the size of the 250 MWt module and also utilise thorium in the fuel. The HTR programme aims at exploring co-generation options in the near-term and producing hydrogen longer-term. Eventually it is intended that a series of HTRs, possibly using Brayton cycle with helium directly driving the gas turbines, will be factory-built and widely installed throughout China. Following the agreement on HTR industrialization, cooperation between CNEC and Tsinghua University in 2003, in March 2014 a new agreement between Tsinghua University and CNEC was described by CNEC as an important milestone in HTR commercialisation. CNEC is responsible for the HTR technical implementation, and becomes the main investor of HTR commercial promotion at home and abroad. In July 2016 CNEC signed an agreement with CGN to promote HTRs.

In January 2016 CNEC signed an agreement with Saudi Arabia’s King Abdullah City for Nuclear and Renewable Energy (KA-CARE) to build a high-temperature reactor in that country, based on the HTR-PM. In August 2016 CNEC signed an agreement with BATAN to develop HTRs in Indonesia.

At the end of 2014 the Nuclear Research and Consultancy Group (NRG), which operates the High Flux Reactor (HFR) at Petten in the Netherlands, completed a "multi-year qualification irradiation project" for the fuel elements produced by INET for the reactor, focused on fission gas release.

In March 2005, there was an agreement between PBMR of South Africa and Chinergy Co. of Beijing. PBMR Pty Ltd had been taking forward the HTR concept (based on earlier German work) since 1993 and was planning to build a 125 MWe demonstration plant. Chinergy Co. was drawing on the small operating HTR-10 research reactor at Tsinghua University which is the basis of the HTR-PM demonstration module which also derives from the earlier German development. The 2005 agreement was for cooperation on the demonstration projects and subsequent commercialisation, since both parties believed that the inherently safe pebble bed technology built in relatively small units would eventually displace the more complex light water reactors. In March 2009, a new agreement was signed between PBMR, Chinergy and INET, but PBMR then ran out of funds.

Russia is pursuing its interest in HTR development through collaboration with China, OKBM being responsible on their side.

Fast neutron reactors

China's R&D on fast neutron reactors started in 1964.

A 65 MWt sodium-cooled fast neutron reactor – the Chinese Experimental Fast Reactor (CEFR) – at the China Institute of Atomic Energy (CIAE) near Beijing, started up in July 2010.¹ It was built by Russia's OKBM Afrikantov in collaboration with OKB Gidropress, NIKIET and Kurchatov Institute. It was grid connected at 40% power (8 MWe net) in July 2011, and ramped up to full 20 MWe power in December, then passed 'official' checks in October 2012. However, it operated only 682 hours to October 2015. It has negative temperature, power reactivity and sodium void coefficients. Its first fuel loading was UO₂, reported to be high-enriched (65%), but ongoing fuel is MOX (25% Pu) with burnup of 60 GWd/t. Core outlet temperature is 530°C. It uses 260t of sodium in the two-loop primary circuit, and 48 t in secondary circuits. Steam temperature in tertiary circuits is 480°C.

The CFR600 demonstration fast reactor (CDFR) is the next step in CIAE's programme, with construction start having been planned for December 2017 at Xiapu in Fujian and operation envisaged from about 2023. This will be 1500 MWt, 600 MWe, with 41% thermal efficiency, using MOX fuel with 100 GWd/t burn-up, and with two sodium coolant loops producing steam at 480°C. Later fuel will be metal with burn-up 100-120 GWd/t. Breeding ratio is about 1.1, design operating lifetime 40 years. This was CIAE's 'project one' Chinese Demonstration Fast Reactor (CDFR).* It is to have active and passive shutdown systems and passive decay heat removal. In June 2018 an intergovernmental agreement with Russia provided for Rosatom to help with construction of the CFR600 and to supply equipment and services, including MOX fuel production. A fuel supply contract to cover seven years' operation was signed with TVEL Electrostal in January 2019.

* In October 2009, an agreement was signed by CIAE and China Nuclear Energy Industry Corporation (CNEIC) with Russia's Atomstroyexport to start pre-project and design works for a commercial nuclear power plant with two BN-800 reactors^c (see section on Sanming in the information page on Nuclear Power in China). These reactors are referred to by CIAE as 'project 2' Chinese Demonstration Fast Reactors (CDFRs), with construction originally to start in 2013 and commissioning 2018-19. In contrast to the intention in Russia, these would use ceramic MOX fuel pellets. The project was expected to lead to bilateral cooperation of fuel cycles for fast reactors. However, according to the Beloyarsk plant Director late in 2014, "The main objective of the BN-800 is [to provide] operating experience and technological solutions that will be applied to the BN-1200," and no further Russian BN-800 units are planned. The China BN-800 project shows no signs of proceeding.

The CFR1000 will be a commercial unit (CCFR, Chinese Commercial Fast Reactor) of 1000-1200 MWe. Subject to a 2020 decision to proceed, construction start could be in December 2028 and operation from about 2034, with metal U-Pu-Zr fuel and 120-150 GWd/t burnup. An earlier design of this, the CDFR1000, was to be a three-loop 2500 MWt pool-type, use MOX fuel with average 66 GWd/t burn-up, run at 544°C, have breeding ratio 1.2, with 316 core fuel assemblies and 255 blanket ones, and a 40-year life. It is to have active and passive shutdown systems and passive decay heat removal. Some of these features may be in the CFR1000.

MOX is seen only as an interim fuel, the target arrangement is metal fuel in closed cycle. The MOX fuel will come from conventionally reprocessed fuel, both PWR and FNR. Metal fuel will enable breeding ratio over 1.2 and will use electrometallurgical reprocessing (pyroprocessing). U-Pu-Zr fuel is also envisaged with 120 GWd/t burn-up and breeding ratio of 1.5, or less with minor actinide and long-lived fission product recycle.

Broadly, 40 GWe of FNR capacity (with conversion ratio of 1) is envisaged by 2050, increasing markedly thereafter and displacing PWR capacity. PWR capacity in China is expected to level off at 200 GWe about 2045, and fast reactors progressively increase from 2030. One scenario with MOX fuel has 200 GWe of fast reactors by 2075, another has 300 GWe with metal fuel by 2055 and 1400 GWe by 2100. A CIAE projection has FNR capacity overtaking PWRs by 2055.

Another fast reactor technology being pursued is the travelling-wave reactor (TWR). CGN and Xiamen University were reported to be cooperating on R&D for this. The Ministry of Science & Technology, with CNNC and SNPTC, were initially skeptical of it.* In January 2013 a prototype TWR-P was being discussed as a TerraPower-SNERDI joint project, and in December 2013 a US Federal Register notice said that the USA had negotiated an agreement with China “that would facilitate the joint development of TWR technology,” including standing wave versions of it. In September 2015 CNNC and TerraPower signed an agreement to work towards building a prototype 600 MWe TWR-P unit at Xiapu in Fujian province, over 2018 to 2025. Phase 3 of the TWR project is over 2019-32 with design and construction of a commercial TWR-C of about 1150 MWe.

* The original TWR design was a fast reactor using natural or depleted uranium packed inside hundreds of hexagonal pillars. In a 'wave' that moves through the core at only one centimetre per year, the U-238 is bred progressively into Pu-239, which is the actual fuel. However, this design was radically changed to become a standing wave reactor with the fuel shuffled in the core.

In September 2017 CNNC subsidiary China National Nuclear Power Company Ltd (CNNP) signed an investment agreement with Shenhua Group, Zhejiang Zheneng Electric Power Co., Huadian Fuxin Energy Limited Company, and Jianntou Energy Investment Co. to develop travelling wave reactors. The agreement sees the establishment of a new joint venture called CNNC Hebei Nuclear Power with a registered capital of 1 billion yuan ($150 million). The new entity is 35% owned by CNNP, with Shenhua Group holding 30%, Huadian Fuxin Energy holding 15%, and Zhejiang Zheneng and Jianntou holding 10% each.

In October 2017 Terrapower announced that it had signed a 50:50 joint venture agreement with CNNC to set up the Global Innovation Nuclear Energy Technology Co Ltd. This involves R&D by CNNC's China Tianjin TWR Investment Company.

China is part of the Generation IV International Forum (GIF) and in 2008 signed the system arrangement for sodium-cooled fast reactors, and has proceeded to the system integration & assessment (SIA) project arrangement for this design, with CIAE taking the lead.

As well as close Russian involvement in China’s fast reactor program, there is a 2008 cooperation protocol with France, involving CIAE and CEA at Cadrache and focused on France’s Astrid and China’s CFR600 plans.

In March 2018 CGN signed an agreement with Ansaldo Nucleare of Italy to cooperate in designing the initial lead-cooled fast reactor of the China Lead-based Reactor (CLEAR) series. The CLEAR development plan will include three phases, the first being a 10 MWt lead-bismuth eutectic-cooled research reactor (CLEAR-I), with both critical and subcritical modes of operation. Ansaldo Nucleare previously led the development of a Generation IV lead-cooled fast reactor design, the 600 MWe ELSY (European Lead-cooled System) or European Lead Fast Reactor (ELFR).

In November 2018 a new international group was launched by the China Academy of Sciences: the Cooperative Alliance for Small Lead-based Fast Reactor (CASLER). South Korea is working on lead-bismuth cooled designs of various sizes which would run on pyroprocessed fuel, and provided the first chairman of CASLER. The lead-cooled reactors are expected to be the first Generation IV types commissioned, before 2030.

Light water reactors

CNNC has been developing an ACP1000, which has led to an ACP100 modular small reactor for electricity, heating and desalination. An ACP600 is also being developed.

CGN has been upgrading its CPR-1000 to the Generation III ACPR1000 with Chinese intellectual property rights.

The NEA then ordered these later 1000 MWe designs to be rationalised, the result of which was the Hualong 1000 or ACC1000. In this the ACP1000 core design prevailed, though it was less mature. Some features of the ACPR1000 are incorporated, at least in the CGN version. The CNNC and CGN versions will be very similar but not identical, and each organisation will maintain much of its own supply chain.

Fuller details of these numerous LWR designs are in the information paper on Nuclear Power in China.

CNNC’s Nuclear Power Institute of China based in Chengdu is working on supercritical water-cooled reactor (SCWR) designs, both pressure vessel and pressure tube types. Two conceptual designs with thermal and mixed neutron spectrum cores have been established. It is reported to be working on a demonstration unit which could be operating in 2022.

Accelerator-driven systems

In connection with the Generation IV International Forum (GIF) collaboration, the China Academy of Sciences (CAS) started in 2011 a new effort to develop an ADS. The China LEAd-based Reactor (CLEAR) was selected as the reference reactor.

Cobalt-60 production

China has started production of the medical and industrial radioisotope cobalt-60 using CNNC's Candu 6 power reactors at Qinshan. This will be China's first domestic production of the isotope. Candu reactors are also used to produce cobalt-60 at Wolsong in South Korea, Bruce in Canada and Embalse in Argentina. The core of a Candu 6 has stainless steel adjusters that help to shape neutron flux to optimise power output and ensure efficient burn up of uranium fuel. The normal cobalt in these can be replaced with cobalt-59, which absorbs neutrons to become Co-60. After about 15 months the stainless steel 'targets' with Co-59 are withdrawn for processing. The development is part of China’s 11th Five Year Plan, and should lead to the production of 220 petabecquerels (PBq) of Co-60 per year – enough to satisfy 80% of Chinese needs. The addition will boost global production by around 10%.

Wastes

Early in 2012 it was reported that Changsha Boiler Plant Co Ltd in Hunan province in collaboration with Shenzhen-based China Nuclear Power Technology Research Institute (CNPRI) was starting to build a plasma furnace or reactor for "transmuting nuclear wastes". No details were supplied.

Non-proliferation

China is a nuclear weapons state, party to the Nuclear Non-Proliferation Treaty (NPT) under which a safeguards agreement with the International Atomic Energy Agency (IAEA) has been in force since 1989, with the Additional Protocol in force since 2002. China undertook nuclear weapons tests in 1964-96. Since then it has signed the Comprehensive Nuclear Test Ban Treaty, although it has not yet ratified it. In May 2004, it joined the Nuclear Suppliers Group (NSG).

The NSG membership gives rise to questions about China's supply of two small power reactors to Pakistan, Chashma 3&4. Contracts for Chashma units 1&2 were signed in 1990 and 2000, before 2004 when China joined the NSG, which maintains an embargo on sales of nuclear equipment to Pakistan. The agreement for units 3&4 was announced in 2007, and signed in October 2008. China argues that units 3&4 are 'grandfathered', under arrangements which are consistent with those for units 1&2. In 2013 contracts were signed for two Hualong One reactors to be built near Karachi.

China has a bilateral safeguards agreement with Australia, and peaceful use agreements for nuclear materials with Canada, USA, Germany and France. The Canadian one is very similar to Australian bilateral safeguards agreements.

China uses Australian-obligated nuclear material only at nuclear facilities covered by its safeguards agreement with the IAEA. However, uranium conversion facilities are before the 'starting point' for IAEA safeguards procedures and are not included in IAEA safeguards agreements with nuclear weapons states. In accordance with long-standing international principles of accounting for nuclear material, on receipt of Australian natural uranium oxide concentrate in China an equivalent quantity of converted natural uranium in the form of uranium hexafluoride will be added to the inventory of a facility designated for safeguards – e.g. an enrichment plant. This will have exactly the same effect as if the natural uranium oxide had moved through the conversion plant, and will ensure that after receipt in China, such material remains in a facility designated for safeguards and listed under the bilateral agreement at all times.

All imported nuclear power plants – from France, Canada and Russia – are under IAEA safeguards, as is the Russian Hanzhun centrifuge enrichment plant in Shaanxi.

A significant number of military production reactors and other plants, with the related Chinese Academy of Engineering Physics, are in Sichuan province.

Notes & references

Notes

a. The Fuzhou mine in the southeastern Jiangxi province is in a volcanic deposit, as is Quinglong.
Xinjiang's Yili basin in the far west of China, in which the Yining (or Kujiltai) ISL mine sits, is contiguous with the Ili uranium province in Kazakhstan, though the geology is apparently different.
The other mines are in granitic deposits.
Source: Uranium 2009: Resources, Production and Demand, OECD Nuclear Energy Agency and International Atomic Energy Agency (2010). [Back]

b. MYRRHA (Multipurpose Hybrid Research Reactor for High-tech Applications) will be a sub-critical assembly relying on accelerated neutrons to achieve periods of criticality in a low-enriched uranium core. As well as being able to produce radiosiotopes and doped silicon, Myrrha's research functions would be particularly well suited to investigating transmutation. Earl in 2010, the Belgian government approved its share of funding of the facility at SCK-CEN's Mol site in northern Belgium. Belgium is to contribute 40% towards the €960 million ($1.3 billion) investment the project will require, but SCK-CEN is looking to set up an international consortium to ensure additional financing. Myrrha itself is scheduled for operation in 2023, but a reduced power model, Guinevere, became operational at Mol in March 2010. [Back]

c. This October 2009 agreement followed a call 12 months earlier by the Russian-Chinese Nuclear Cooperation Commission for construction of an 800 MWe demonstration fast reactor similar to the OKBM Afrikantov design being built at Beloyarsk 4 and (then) due to start up in 2012. In June 2009 Rosatom and CNNC had signed an agreement for construction of two BN-800 demonstration reactors in China, and St Petersburg Atomenergoproekt (now Atomproekt) said it was starting design work on a BN-800 reactor for China. [Back]

References

1. Criticality for fast reactor, World Nuclear News (22 July 2010); Chinese fast reactor nears commissioning, World Nuclear News (7 April 2009) [Back]

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China Guangdong Nuclear Power Group website

China National Nuclear Corporation website

Dynabond Powertech

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Yang Hongyi, CIAE & CNNC, Fast Reactor Progress and Cooperation with French (November 2015)

Hongjie Xu, SINAP, Status and Perspective of TMSR in China, presented at the Generation IV International Forum (GIF) Molten Salt Reactor Workshop at the the Paul Scherrer Institute on 24 January 2017

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Senin, 04 Maret 2019

Uranium Enrichment

http://www.world-nuclear.org/information-library/nuclear-fuel-cycle/conversion-enrichment-and-fabrication/uranium-enrichment.aspx

International Enrichment Centres, Multilateral approaches

Enrichment Processes

Centrifuge process

Laser processes

Gaseous diffusion process

Electromagnetic process

Aerodynamic processes

Enrichment of reprocessed uranium

Enrichment of depleted uranium tails

After enrichment

Environmental Issues

Notes & References

General sources

Nuclear Fuel Cycle / Conversion Enrichment and Fabrication / Conversion and Deconversion

Conversion and Deconversion

Conversion process

Secondary sources of conversion supply

Depleted uranium and deconversion

Notes & references

General references

Enriched uranium

Contents

Grades

Reprocessed uranium (RepU)

Low enriched uranium (LEU)

Highly enriched uranium (HEU)

Enrichment methods

Diffusion techniques

Gaseous diffusion

Thermal diffusion

Centrifuge techniques

Gas centrifuge

Zippe centrifuge

Laser techniques

Atomic vapor laser isotope separation (AVLIS)

Molecular laser isotope separation (MLIS)

Separation of Isotopes by Laser Excitation (SILEX)

Other techniques

Aerodynamic processes

Electromagnetic isotope separation

Chemical methods

Plasma separation

Separative work unit

Cost issues

Downblending

Global enrichment facilities

See also

References

External links

Navigation menu

China's Nuclear Fuel Cycle

Domestic uranium resources and mining

Mineral exploration

International uranium sources

CNNC initiatives abroad

CGN initiatives abroad

Alternative sources of uranium

Nuclear fuel industrial parks

Conversion

Enrichment and enriched uranium imports

Enriched uranium

Fuel fabrication

Fuel cycles

Reprocessing, recycling

Mixed oxide fuel

Waste management

Storage of used fuel and disposal of HLW

Low- and intermediate-level waste

Decommissioning

Heavy manufacturing industrial parks

Research & development

Research Reactors

Reactor and fuel cycle development

Thorium molten salt reactor programme

Recycled uranium in PHWRs; thorium in PHWRs

HTR demonstration: HTR-10

Commercial HTRs: Shidaowan HTR-PM, Ruijin or Wan'an