The Nuclear Fuel Cycle
(Updated May 2015)
- The nuclear fuel cycle is the series of industrial processes which involve the production of electricity from uranium in nuclear power reactors.
- Uranium is a relatively common element that is found throughout the world. It is mined in a number of countries and must be processed before it can be used as fuel for a nuclear reactor.
- Fuel removed from a reactor, after it has reached the end of its useful life, can be reprocessed to produce new fuel.
The various activities associated with the production of electricity from nuclear reactions are referred to collectively as the nuclear fuel cycle. The nuclear fuel cycle starts with the mining of uranium and ends with the disposal of nuclear waste. With the reprocessing of used fuel as an option for nuclear energy, the stages form a true cycle.
To prepare uranium for use in a nuclear reactor, it undergoes the steps of mining and milling, conversion, enrichment and fuel fabrication. These steps make up the 'front end' of the nuclear fuel cycle.
After uranium has spent about three years in a reactor to produce electricity, the used fuel may undergo a further series of steps including temporary storage, reprocessing, and recycling before wastes are disposed. Collectively these steps are known as the 'back end' of the fuel cycle.
Uranium is a slightly radioactive metal that occurs throughout the Earth's crust (see page on Uranium and Depleted Uranium). It is about 500 times more abundant than gold and about as common as tin. It is present in most rocks and soils as well as in many rivers and in sea water. It is, for example, found in concentrations of about four parts per million (ppm) in granite, which makes up 60% of the Earth's crust. In fertilisers, uranium concentration can be as high as 400 ppm (0.04%), and some coal deposits contain uranium at concentrations greater than 100 ppm (0.01%). Most of the radioactivity associated with uranium in nature is in fact due to other minerals derived from it by radioactive decay processes, and which are left behind in mining and milling.
There are a number of areas around the world where the concentration of uranium in the ground is sufficiently high that extraction of it for use as nuclear fuel is economically feasible. Such concentrations are called ore.
Both excavation and in situ techniques are used to recover uranium ore. Excavation may be underground and open pit mining.In general, open pit mining is used where deposits are close to the surface and underground mining is used for deep deposits, typically greater than 120 m deep.
Open pit mines require large holes on the surface, larger than the size of the ore deposit, since the walls of the pit must be sloped to prevent collapse. As a result, the quantity of material that must be removed in order to access the ore may be large. Underground mines have relatively small surface disturbance and the quantity of material that must be removed to access the ore is considerably less than in the case of an open pit mine. Special precautions, consisting primarily of increased ventilation, are required in underground mines to protect against airborne radiation exposure.
An increasing proportion of the world's uranium now comes from in situ leach (ISL) mining, where oxygenated groundwater is circulated through a very porous orebody to dissolve the uranium oxide and bring it to the surface. ISL may be with slightly acid or with alkaline solutions to keep the uranium in solution. The uranium oxide is then recovered from the solution as in a conventional mill.
The decision as to which mining method to use for a particular deposit is governed by the nature of the orebody, safety and economic considerations.
For more detailed information see the information pages on:
Milling, which is generally carried out close to a uranium mine, extracts the uranium from the ore (or ISL leachate). Most mining facilities include a mill, although where mines are close together, one mill may process the ore from several mines. Milling produces a uranium oxide concentrate which is shipped from the mill. It is sometimes referred to as 'yellowcake' and generally contains more than 80% uranium. The original ore may contain as little as 0.1% uranium, or even less.
In a mill, the ore is crushed and ground to a fine slurry which is leached in sulfuric acid (or sometimes a strong alkaline solution) to allow the separation of uranium from the waste rock. It is then recovered from solution and precipitated as uranium oxide (U3O8) concentrate. After drying and usually heating it is packed in 200-litre drums as a concentrate, sometimes referred to as 'yellowcake' (though it is usually khaki).
U3O8 is the uranium product which is sold. About 200 tonnes is required to keep a large (1000 MWe) nuclear power reactor generating electricity for one year.
The remainder of the ore, containing most of the radioactivity and nearly all the rock material, becomes tailings, which are emplaced in engineered facilities near the mine (often in a mined out pit). Tailings need to be isolated from the environment because they contain long-lived radioactive materials in low concentrations and maybe also toxic materials such as heavy metals. However, the total quantity of radioactive elements is less than in the original ore, and their collective radioactivity will be much shorter-lived.
Conversion and enrichment
The uranium oxide product of a uranium mill is not directly usable as a fuel for a nuclear reactor and additional processing is required. Only 0.7% of natural uranium is 'fissile', or capable of undergoing fission, the process by which energy is produced in a nuclear reactor. The form, or isotope, of uranium which is fissile is the uranium-235 (U-235) isotope. The remainder is uranium-238 (U-238). For most kinds of reactor, the concentration of the fissile uranium-235 isotope needs to be increased – typically to between 3.5% and 5% U-235. Isotope separation is a physical process to concentrate (‘enrich’) one isotope relative to others. The enrichment process requires the uranium to be in a gaseous form. The uranium oxide concentrate is therefore first converted to uranium hexafluoride, which is a gas at relatively low temperatures.
At a conversion facility, the uranium oxide is first refined to uranium dioxide, which can be used as the fuel for those types of reactors that do not require enriched uranium. Most is then converted into uranium hexafluoride, ready for the enrichment plant. The main hazard of this stage of the fuel cycle is the use of hydrogen fluoride. The uranium hexafluoride is then drained into 14-tonne cylinders where it solidifies. These strong metal containers are shipped to the enrichment plant.
The enrichment process separates gaseous uranium hexafluoride into two streams, one being enriched to the required level and known as low-enriched uranium; the other stream is progressively depleted in U-235 and is called 'tails', or simply depleted uranium.
The main enrichment process in commercial plants uses centrifuges, with thousands of rapidly-spinning vertical tubes. As they spin, the physical properties of molecules, specifically the 1% mass difference between the two uranium isotopes, separates them. A laser enrichment process is in the final stage of development.
The product of this stage of the nuclear fuel cycle is enriched uranium hexafluoride, which is reconverted to produce enriched uranium oxide. Up to this point the fuel material can be considered fungible (though enrichment levels vary), but fuel fabrication involves very specific design.
Enrichment is covered in detail in the page on Uranium Enrichment.
A small number of reactors, notably the Canadian and Indian heavy water type, do not require uranium to be enriched.
Reactor fuel is generally in the form of ceramic pellets. These are formed from pressed uranium oxide (UO2) which is sintered (baked) at a high temperature (over 1400°C) a . The pellets are then encased in metal tubes to form fuel rods, which are arranged into a fuel assembly ready for introduction into a reactor. The dimensions of the fuel pellets and other components of the fuel assembly are precisely controlled to ensure consistency in the characteristics of the fuel.
In a fuel fabrication plant great care is taken with the size and shape of processing vessels to avoid criticality (a limited chain reaction releasing radiation). With low-enriched fuel criticality is most unlikely, but in plants handling special fuels for research reactors this is a vital consideration.
Some 27 tonnes of fresh enriched fuel is required each year by a 1000 MWe reactor.
Power generation and burn-up
Several hundred fuel assemblies make up the core of a reactor. For a reactor with an output of 1000 megawatts (MWe), the core would contain about 75 tonnes of low-enriched uranium. In the reactor core the U-235 isotope fissions or splits, producing a lot of heat in a continuous process called a chain reaction. The process depends on the presence of a moderator such as water or graphite, and is fully controlled.
Some of the U-238 in the reactor core is turned into plutonium and about half of this is also fissioned, providing about one third of the reactor's energy output.
As in fossil-fuel burning electricity generating plants, the heat is used to produce steam to drive a turbine and an electric generator, in this case providing over 7 billion kilowatt hours of electricity in one year.
To maintain efficient reactor performance, about one-third of the spent fuel is removed every year or 18 months, to be replaced with fresh fuel.
Typically, some 44 million kilowatt-hours of electricity are produced from one tonne of natural uranium. The production of this amount of electrical power from fossil fuels would require the burning of over 20,000 tonnes of black coal or 8.5 million cubic metres of gas.
An issue in operating reactors and hence specifying the fuel for them is fuel burn-up. This is measured in gigawatt-days per tonne and its potential is proportional to the level of enrichment. Hitherto a limiting factor has been the physical robustness of fuel assemblies, and hence burn-up levels of about 40 GWd/t have required only around 4% enrichment. But with better equipment and fuel assemblies, 55 GWd/t is possible (with 5% enrichment), and 70 GWd/t is in sight, though this would require 6% enrichment. The benefit of this is that operation cycles can be longer – around 24 months – and the number of fuel assemblies discharged as used fuel can be reduced by one third. Associated fuel cycle cost is expected to be reduced by about 20%.
In CANDU reactors using natural uranium, burn-up is much less, about 7.5 GWd/t, but in terms of efficiency this is equivalent to almost 50 GWd/t for enriched fuel.
Burn-up in GWd/t is the conventional measure for oxide fuels, and 60 GWd/t U is equivalent to about 6.5% atomic percent burn-up, i.e. about 6.5% of the original uranium atoms are burned directly, or indirectly via transformation to fissile plutonium. (With metal fuels, the atomic percent metric is used, and a new light water reactor metal fuel is targeting 21 atomic percent burn-up when it is deployed in 2020s.)
As with as a coal-fired power station about two thirds of the heat is dumped, either to a large volume of water (from the sea or large river, heating it a few degrees) or to a relatively smaller volume of water in cooling towers, using evaporative cooling (latent heat of vapourisation).
With time, the concentration of fission fragments and heavy elements formed in the same way as plutonium in the fuel will increase to the point where it is no longer practical to continue to use the fuel. So after 18-36 months the used fuel is removed from the reactor. The amount of energy that is produced from a fuel assembly varies with the type of reactor and the policy of the reactor operator. Used fuel will typically have about 0.9% U-235 and 0.6% fissile plutonium (almost 1% Pu total).
When removed from a reactor, the fuel will be emitting both radiation, principally from the fission fragments, and heat. It is unloaded into a storage pond immediately adjacent to the reactor to allow the radiation levels to decrease. In the ponds the water shields the radiation and absorbs the heat, which is removed by circulating the water to external heat exchangers. Used fuel is held in such pools for several months and sometimes many years. It may be transferred to naturally-ventilated dry storage on site after about five years.
Depending on policies in particular countries, some used fuel may be transferred to central storage facilities. Ultimately, used fuel must either be reprocessed or prepared for permanent disposal. The longer it is stored, the easier it is to handle, due to decay of radioactivity.
There are two alternatives for used fuel:
- reprocessing to recover and recycle the usable portion of it
- long-term storage and final disposal without reprocessing.
Used fuel still contains about 96% of its original uranium, of which the fissionable U-235 content has been reduced to less than 1%. About 3% of the used fuel comprises waste products and the remaining 1% is plutonium (Pu) produced while the fuel was in the reactor and not 'burned' then.
Reprocessing separates uranium and plutonium from waste products (and from the fuel assembly cladding) by chopping up the fuel rods and dissolving them in acid to separate the various materials. It enables recycling of the uranium and plutonium into fresh fuel, and produces a significantly reduced amount of waste (compared with treating all used fuel as waste). See page on Processing of Used Nuclear Fuel. The remaining 3% of high-level radioactive wastes (some 750 kg per year from a 1000 MWe reactor) can be stored in liquid form and subsequently solidified.
Uranium and plutonium recycling
The uranium recovered from reprocessing, which typically contains a slightly higher concentration of U-235 than occurs in nature, can be reused as fuel after conversion and enrichment.
The plutonium can be directly made into mixed oxide (MOX) fuel, in which uranium and plutonium oxides are combined. In reactors that use MOX fuel, plutonium substitutes for the U-235 in normal uranium oxide fuel (see page on Mixed Oxide (MOX) Fuel).
According to Areva, about eight fuel assemblies reprocessed can yield one MOX fuel assembly, two-thirds of an enriched uranium fuel assembly, and about three tonnes of depleted uranium (enrichment tails) plus about 150 kg of wastes. It avoids the need to purchase about 12 tonnes of natural uranium from a mine.
Wastes from the nuclear fuel cycle are categorised as high-, medium- or low-level wastes by the amount of radiation that they emit. These wastes come from a number of sources and include:
- low-level waste produced at all stages of the fuel cycle;
- intermediate-level waste produced during reactor operation and by reprocessing;
- high-level waste, which is waste containing the highly-radioactive fission products separated in reprocessing, and in many countries, the used fuel itself. Separated high-level wastes also contain long-lived transuranic elements.
After reprocessing, the liquid high-level waste can be calcined (heated strongly) to produce a dry powder which is incorporated into borosilicate (Pyrex) glass to immobilise it. The glass is then poured into stainless steel canisters, each holding 400 kg of glass. A year's waste from a 1000 MWe reactor is contained in five tonnes of such glass, or about 12 canisters 1.3 metres high and 0.4 metres in diameter. These can readily be transported and stored, with appropriate shielding.
The uranium enrichment process leads to the production of much 'depleted' uranium, in which the concentration of U-235 is significantly less than the 0.7% found in nature. Small quantities of this material, which is primarily U-238, are used in applications where high density material is required, including radiation shielding and some is used in the production of MOX fuel. While U-238 is not fissile it is a low specific activity radioactive material and some precautions must, therefore, be taken in its storage or disposal.
See page on Waste Management in the Nuclear Fuel Cycle.
Used fuel and separated wastes: final disposal
At the present time, there are no disposal facilities (as opposed to storage facilities) in operation in which used fuel, not destined for reprocessing, and the waste from reprocessing, can be placed. In either case the material is in a solid, stable wasteform.
Although technical issues related to disposal are straightforward, there is currently no pressing technical need to establish such facilities, as the total volume of such wastes is relatively small. Further, the longer it is stored the easier it is to handle, due to the progressive decrease of radioactivity.
There is also a reluctance to dispose of used fuel because it represents a significant energy resource which could be reprocessed at a later date to allow recycling of the uranium and plutonium.
A number of countries are carrying out studies to determine the optimum approach to the disposal of used fuel and wastes from reprocessing. The general consensus favours its placement into deep geological repositories, about 500 metres down, initially recoverable before being permanently sealed.
Other Sources of Nuclear Fuel
In the 1990s uranium mines gained a competitor, in many ways very welcome, as military uranium came on to the civil market under a US-Russian agreement. Since then half of the uranium used for electricity in the USA has come from Russian military stockpiles, and worldwide about one sixth of the market has been supplied thus.
Weapons-grade uranium in stockpiles built up during 1950s and 1960s has been enriched to more than 90% U-235 and must be diluted about 1:25 or 1:30 with depleted uranium (about 0.3% U-235). This means that progressively, Russian and other stockpiles of weapons material are used to produce electricity.
Weapons-grade plutonium may also be used to make mixed oxide (MOX) fuel for use in ordinary reactors or in special reactors designed to 'burn' it for electricity. Another US-Russian agreement covers disposition of military plutonium from both countries into MOX fuel.
Material balance in the nuclear fuel cycle
The following figures may be regarded as typical for the annual operation of a 1000 MWe nuclear power reactor such as many operating today:b
||Anything from 20,000 to 400,000 tonnes of uranium ore
||230 tonnes of uranium oxide concentrate (which contains 195 tonnes of uranium)
||288 tonnes uranium hexafluoride, UF6 (with 195 tU)
||35 tonnes enriched UF6 (containing 24 t enriched U) – balance is 'tails'
||27 tonnes UO2 (with 24 t enriched U)
||8760 million kWh (8.76 TWh) of electricity at full output, hence 22.3 tonnes of natural U per TWh
||27 tonnes containing 240 kg transuranics (mainly plutonium), 23 t uranium (0.8% U-235), 1100 kg fission products.
The following figures assume the annual operation of 1000 MWe of nuclear power reactor capacity such as in the new EPR, with 5% enriched fuel and higher (65 GWd/t) burn-up:
||Anything from 20,000 to 400,000 tonnes of uranium ore
||171 tonnes of uranium oxide concentrate (which contains 145 tonnes of uranium)
||214 tonnes uranium hexafluoride, UF6 (with 145 tU)
||23 tonnes enriched UF6 (containing 15.6 t enriched U) – balance is 'tails' (0.20%)
||17.5 tonnes UO2 (with 15.6 t enriched U)
||8760 million kWh (8.76 TWh) of electricity at full output, hence 16.5 tonnes of natural U per TWh
||17.5 tonnes containing 14.5 t uranium (0.8% U-235).
Between the above figures, Uranium 2014: Resources, Production and Demand ('Red Book'), from the OECD NEA & IAEA, said that efficiencies on power plant operation and lower enrichment tails assays meant that uranium demand per unit capacity was falling, and the report’s generic reactor fuel consumption was reduced from 175 tU per GWe per year at 0.30% tails assay (2011 report) to 163 tU per GWe per year at 0.25% tails assay. The corresponding U3O8 figures are 206 tonnes and 192 tonnes per GWe per year.
a. UO2 has a very high melting point – 2865°C (compared with uranium metal – 1132°C). [Back]
b. Figures are based on the following assumptions: enrichment to 4% U-235 with 0.25% tails assay – hence 140,000 SWU (separative work units) of enrichment needed (one SWU requires about 50kWh of electricity at a gas centrifuge enrichment plant); core load 72 tU, refuelling so that 24 tU/yr is replaced; operation – 45,000 MWday/t (45 GWd/t) burn-up, 33% thermal efficiency.
In fact, a 1000 MWe reactor cannot be expected to run at 100% load factor – 90% is more typical, so an output of around 7.75 TWh/yr is more realistic, but this simply means scaling back the inputs accordingly, e.g. to 175.5 tU.
With the higher (5%) enrichment and burn-up in the second (EPR) set of figures, enrichment input rises to 145,000 SWU. In the used fuel, transuranic and fission product numbers will be slightly lower due to high thermal efficiency.
Canadian figures for tU/GWe/yr suggest slightly lower uranium requirements and utilization for PHWRs than for light water reactors. An International Atomic Energy Agency technical report1 gives 157 tU at typical 7.5 GWd/t burn-up and 31% thermal efficiency, or 142 tU at 90% capacity factor, hence 80% of the input compared with a typical LWR above. This is 17.9 tU/TWh.
Considering just how much of the original uranium is actually used: 0.7% fissile U-235 is in natural U (Unat), on above 'typical' figures: 0.49% of Unat goes into fuel as the fissile part, 0.394% is actually fissioned, and in addition about half that much U-238 turned into Pu-239 is fissioned, giving about a 0.6% utilization of the original Unat.
With the EPR figures: 0.538% of Unat goes into fuel as the fissile part, 0.452% of that is actually fissioned, and in addition about half that much U-238 turned into Pu-239 is fissioned, giving about a 0.67% utilization of the original Unat. [Back]
1. Heavy Water Reactors: Status and Projected Development, Technical Reports Series No. 407, International Atomic Energy Agency, 2002, STI/DOC/010/407 (ISBN: 9201115024). PHWR data is taken from Chapter 6, HWR Fuel Cycles [Back]