click to enlarge4.1 Mining and milling of uranium ore
Uranium minerals are always associated with other elements such as radium and radon in radioactive decay series (see Appendix 2). Therefore, although uranium itself is barely radioactive, the ore which is mined must be regarded as potentially hazardous, especially if it is high-grade ore. The radiation hazards involved however are virtually all due to the associated elements and are similar to those in many mineral sands operations.
Australian uranium mines have mostly been open cut and therefore naturally well ventilated. Ore grades at Ranger, as well as at the proposed Jabiluka and Kintyre mines, are less than 0.5% U3O8. The Olympic Dam underground mine, with ore grade less than 0.1% U3O8. Any underground uranium mine is ventilated with powerful fans.
Canada's older mines at Cluff Lake, Key Lake and Rabbit Lake, as well as McClean Lake which started in 1999, are all open cut mines and well ventilated. The three newer mines are underground operations. Two of them, McArthur River and Cigar Lake, are in very high-grade ore and require special remote-control techniques for mining. There is some underground mining at Cluff Lake, and there will also be some at McClean Lake later.
The mined ore (i.e. rock containing economically recoverable concentrations of uranium) is crushed and ground. The resulting slurry is then leached, usually with sulfuric acid, to dissolve the uranium (together with some other metals). The solids remaining after the uranium is extracted are pumped as a slurry to the tailings dam, which is engineered to retain them securely. Tailings contain most of the radioactive material in the ore, such as radium.
Two new Australian mines are in situ leaching (ISL) operations, with recovery of the uranium from the sandy ore taking place underground. A slightly acidic and heavily oxygenated solution is circulated through bores and the uranium is extracted in plant at the surface, with the liquor being recirculated.
In each case the leach liquor goes through a solvent extraction or ion exchange process followed by precipitation to remove the uranium from solution as a bright yellow precipitate ("yellowcake"). After high-temperature drying, the uranium oxide (U3O8), now khaki in colour, is packed into 200-litre drums for shipment. The radiation dose rate one metre from such a drum of freshly processed U3O8 is about half that (from cosmic rays) received by a person on a commercial jet flight.
In Australia all these operations are undertaken under the Australian Code of Practice on Radiation Protection in the Mining and Milling of Radioactive Ores, administered by state governments. In Canada, the Canadian Nuclear safety Commission regulations apply. In both countries they set strict health standards for gamma radiation and radon gas exposure* as well as for ingestion and inhalation of radioactive materials. Standards apply to both workers and members of the public.
* 20 mSv/yr averaged over 5 years is the maximum allowable radiation dose rate for workers, including radon (and radon daughters) dose. This is in addition to natural background and excludes medical exposure. See also Appendix 1 and glossary for definitions.
The gamma radiation comes principally from isotopes of bismuth and lead. The radon gas emanates from the rock (or tailings) as radium decays.* It then decays itself to (solid) radon daughters, which are energetic alpha-emitters. Radon occurs in most rocks and traces of it are in the air we all breathe. However, at high concentrations it is a health hazard since its short half-life means that disintegrations giving off alpha particles are occurring relatively frequently. Alpha particles discharged in the lung can eventually give rise to lung cancer.
* "Radon" here normally refers to Rn-222. Another isotope, Rn-220 (known as "thoron"), is given off by thorium, which is a constituent of many Australian mineral sands. At Port Pirie in S.A. a rare earths treatment plant operating between 1968 and 1972 produced tailings containing thorium. These gave rise to minor Rn-220 emissions in addition to Rn-222 emanating from earlier Radium Hill uranium tailings at the same site. See also Appendix 2.
A number of precautions are taken at a uranium mine to protect the health of workers:
*Both lead and uranium are toxic and affect the kidney. The body progressively eliminates most Pb or U, via the urine.
Since the fifteenth century many miners who had worked underground in the mountains near the present border between East Germany and the Czech Republic contracted a mysterious illness, and many died prematurely. In the late 1800s the illness was diagnosed as lung cancer, but it was not until 1921 that radon gas was suggested as the possible cause. Although this was confirmed by 1939, between 1946 and 1959 much underground uranium mining took place in the USA without the precautions which might have become established as a result of the European experience. In the early 1960s a higher than expected incidence of lung cancer began to show up among miners who smoked. The cause was then recognised as the emission of alpha particles from radon and, more importantly, its solid daughter products of radioactive decay. The miners concerned had been exposed to high levels of radon 10-15 years earlier, accumulating radiation doses well in excess of present recommended levels.
The small unventilated uranium "gouging" operations in the USA which led to the greatest health risk are a thing of the past. In the last 40 years, individual mining operations have been larger, and efficient ventilation and other precautions now protect underground miners from these hazards. Open cut mining of uranium virtually eliminates the danger. There has been no known case of illness caused by radiation among uranium miners in Australia or Canada. While this may be partly due to the lack of detailed information on occupational health from operations in the 1950s, it is clear that no major occupational health effects have been experienced in either country.
After mining is complete most of the orebody, with virtually all of the radioactive radium, thorium and actinium materials, will end up in the tailings dam.* Hence radiation levels and radon emissions from tailings will probably be significant. In the unlikely event of someone setting up camp on top of the material, they could eventually receive a radiation dose exceeding international standards, just as they could from outcropping orebodies. Therefore, the tailings need to be covered over with enough rock, clay and soil to reduce both gamma radiation levels and radon emanation rates to levels near those naturally occurring in the region. A vegetation cover can then be established.
* About 95% of the radioactivity in the ore is from the U-238 decay series (see Appendix 2), totalling about 450 kBq/kg in ore with 0.3% U3O8 (eg from Ranger). The U-238 series has 14 radioactive isotopes in secular equilibrium, thus each represents about 32 kBq/kg (irrespective of the mass proportion). When the ore is processed, the U-238 and the very much smaller masses of U-234 (and U-235) are removed. The balance becomes tailings, and at this point has about 85% of its original intrinsic radioactivity. However, with the removal of most U-238, the following two short-lived decay products (Th-234 & Pa-234) soon disappear, leaving the tailings with a little over 70% of the radioactivity of the original ore after several months. The controlling long lived isotope then becomes Th-230 which decays with a half life of 77,000 years to radium-226 followed by radon-222. (Alex Zapantis, Supervising Scientist Group, Australia).
Radon emanation from tailings during mining and before they are covered is sometimes seen as a general environmental hazard. However, traces of radon are emitted by minerals present in most rock and soil. Thus, apart from the local hazards mentioned above, the regional increase in radon release due to mining operations is very small (see also notes on radiation in 6.3).
Process water, which arises from settled tailings, contains radium and other metals that would be undesirable in the outside environment. This water is retained and evaporated so that the contained metals are retained in safe storage, as in an orebody. In fact process water is never released to natural waterways, but is stored in the tailings dam and (at Ranger) the original mine pit, and evaporated from there.
At Ranger, rainfall run-off is segregated in accordance with water quality, and high quality water from relatively undisturbed catchments is released during flood times*. Poorer quality water is retained on site and treated. Originally this more contaminated water was to be released to the nearby creek during periods of high flow, but in practice none has ever been discharged.
* Radionuclide levels are not to exceed drinking water standards.
The former Australian uranium mine at Rum Jungle is best known to some people as a source of water pollution. Here the uranium ore was associated with a lot of sulphide mineralisation. In accordance with the low standards of the 1950s, few precautions were taken to prevent river pollution from the site either at the time or following mining**. Large heaps of both waste rock and low-grade ore in the monsoonal climate caused a large amount of acidic run-off, known as acid mine drainage.
**Metal sulphides in contact with water and air in a warm climate tend to react readily, particularly in the presence of certain bacteria. Sulphuric acid is generated and toxic heavy metals such as copper then go into solution and may be carried downstream.
Fuel cycles describe the way in which fuel gets to nuclear reactors and what happens to it when it comes out.
All aspects of obtaining and preparing the fuel, using it, and the management of spent fuel together make up what is known as the fuel cycle. As the term suggests it has been the intention with nuclear power to recycle the unused part of the spent fuel so that it is incorporated into the fresh reactor fuel elements.
Figure 10. The open fuel cycle
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Unlike coal, uranium ore cannot be fed directly into a power station; it has to be purified, concentrated (usually) and made up into special fuel rods. Figure 10 shows the so-called "open fuel cycle" for nuclear power, which is the system as it stands today in most countries using the most common kinds of reactors.
Starting in uranium mines such as Olympic Dam or Ranger in Australia or the northern Saskatchewan mines of Canada, ore is mined and milled to produce uranium in the form of uranium oxide concentrate. It is a mixture of two oxides, commonly known as U3O8. This material, a khaki-coloured powder, is shipped to customers. It has the same isotopic ratio as the ore, where uranium-235 (U-235) is present to the extent of about 0.7 percent. The rest is a heavier isotope of uranium - U-238 (with traces of U-234). Most reactors, including the common light water type (LWR), cannot run on natural uranium, so the proportion of U-235 must be increased to about 3.5 percent. This is called enrichment. Canadian reactors use unenriched uranium - see below.
Enrichment is a fairly high-technology physical process which requires the uranium to be in the form of a gas. The simplest way to achieve this is to convert the uranium oxide to uranium hexafluoride, which is a gas at little more than room temperature. This form of uranium is commonly referred to as UF6 or "hex". Hence the first destination of uranium oxide concentrate from a mine is a conversion plant where it is purified and converted to uranium hexafluoride.
The UF6 is then fed to an enrichment plant* which increases the proportion of the fissile U-235 isotope. In the process about 85% of the natural uranium feed is rejected as "depleted uranium" or "tails" (mainly U-238) which is stockpiled **. Thus, after enrichment about 15% of the original quantity is available as enriched uranium containing about 3.5 percent U-235.
*Most enrichment has so far been undertaken using the expensive and energy intensive gaseous diffusion process. Newer plants are mostly based on very much more efficient gas centrifuge technology. The next generation of enrichment plants may use advanced laser technology.
**This material cannot be used in current types of reactors, its only significant use is as a feed for fast breeder reactors, or to dilute ex-military uranium, see sections 4.4 & 3.5. It is stored as UF6 in steel cylinders. Usually less than 0.3% U-235 remains in it.
The enrichment methods now in use are based on the slight difference in atomic mass of U-235 and U-238. Much of the installed capacity relies on the gaseous diffusion process, where the UF6 gas is passed through a long series of membrane barriers which allow the lighter molecules with U-235 through faster than the U-238 ones. More modern plants use high-speed centrifuges to separate the molecules of the two isotopes.
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The large Tricastin enrichment plant in France (beyond cooling towers)
The four nuclear reactors in the foreground provide over 3000 MWe power for it.
Enriched uranium then goes on to a fuel fabrication plant where the reactor fuel elements are made. The UF6 is converted to uranium dioxide, a ceramic material, and formed into small cylindrical pellets about 2 cm long and 1.5cm in diameter. The pellets are loaded into zirconium alloy or stainless steel tubes about 4 metres long to form fuel rods. These are assembled into bundles about 30 cm square to form reactor fuel assemblies. Fuel assemblies of this type are used to power the US-developed light water power reactor, currently the most popular design (see Table 5). A 1000 MWe reactor has about 75 tonnes of fuel in it.
click to enlarge
A PWR fuel assembly
Canadian CANDU (CANadian Deuterium Uranium) reactors have a different design, and run on natural (ie unenriched) uranium. Instead of a single large pressure vessel containing the core, they have multiple (eg 300-600) horizontal pressure tubes, each containing fuel and heavy water coolant. The pressure tubes extend through the reactor vessel, or calandria, which contains the heavy water moderator*. CANDU fuel bundles are only 10 cm in diameter and 50 cm long.
* Heavy water, or deuterium oxide, contains deuterium, which is an isotope of hydrogen having a neutron in the nucleus.
Inside all kinds of operating reactors a fission chain reaction occurs in the fuel rods, as described in 3.1. Fast neutrons are slowed by the water, heavy water or graphite moderator so that they can cause fission. Neutron-absorbing control rods are inserted or withdrawn to regulate the speed of the reaction. Heat from the fission reaction is conveyed from the reactor core and is used to make steam, which in turn is used to generate electricity.
In a light water reactor the fuel stays in the reactor for about three years, generating heat from fission of both the U-235 and also the fissile plutonium (eg Pu-239) which is formed there. After three years or so, the level of fission products and other neutron-absorbers has built up so that the reaction is slowing down, and the spent fuel assemblies are therefore removed. About one third of the fuel is changed each year. In a CANDU type, fuel stays in the reactor only 18 months or so.
When removed, spent fuel is hot and radioactive. It is therefore stored under water to remove the heat and to provide shielding from radiation, pending the next step. This may be reprocessing in the case of countries such as UK, France and Japan, which have chosen to close the fuel cycle, or it may be final disposal in the case of countries such as USA, Canada and Sweden, which have chosen the "open fuel cycle". Storage is initially at the reactor site. It may then be transferred elsewhere, or to an engineered dry storage facility.
Earlier generations of reactors, such as are still operating in the UK, use uranium metal fuel instead of uranium oxide, and are gas-cooled. For these reactors reprocessing operations have been going on for some time, so that the fuel elements are not held very long in cooling ponds. This, and the corresponding arrangement for light water reactors, is illustrated by the more complex diagram in Figure 11 which is known as the "closed fuel cycle" system.
Figure 11. The closed fuel cycle
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In the closed fuel cycle for light water reactors fuel is supplied in exactly the same way as before. Starting with uranium mines and mills the uranium goes through conversion, enrichment, and fuel fabrication to the reactor.
But after being removed from the reactor the fuel rods are put through a reprocessing plant where they are chopped up and dissolved in acid. Various chemical processes recover and separate the two valuable components: plutonium and unused uranium. This leaves about 3% of the fuel as high-level waste. After solidification it is reduced to a small volume of highly radioactive material suitable for permanent disposal. See also sections 5.2 - 5.3.
About 96% of the uranium which goes into the reactors emerges again in the spent fuel, albeit depleted to less than 1% U-235. As shown in Figure 14 some of the remainder is converted into heat and radioactive fission products and some into plutonium and other actinide elements. Hence reprocessing spent fuel has some economic benefits in recovering the unused uranium and the plutonium which has been generated and not burned in the reactor. It also substantially reduces the volume of material to be disposed of as high-level waste, which also has economic benefit.
Plutonium comprises about 1% of the spent fuel. It is a very good nuclear fuel which needs no enrichment process, it can be mixed with depleted uranium, made into fuel rods in a mixed oxide (MOX) fuel fabrication plant and put back into the reactor as fresh fuel (see 5.2). Alternatively it could be used to fuel future breeder reactors (see below).
The recovered uranium can go back to be enriched, and on into fresh fuel for a reactor. The closed fuel cycle is thus a more efficient system for making maximum use of the uranium dug out of the ground (by about 30%, in energy terms) and that is why the industry originally favoured this approach. However, due largely to many years of low uranium prices (since about 1980), plans for widespread reprocessing of spent reactor fuel have not eventuated. France, Germany, UK, Russia and Japan are proceeding with the closed fuel cycle for oxide fuels, and across Europe over 35 reactors are licensed to load 20-50% of their core with MOX fuel containing up to 7% reactor-grade plutonium.
Today's nuclear reactor technology is distinctly better than that represented by most of the world's operating plants, and the first advanced reactors are now in service in Japan.
Reactor suppliers in North America, Japan and Europe have nine new nuclear reactor designs either approved or at advanced stages of planning, and others at a research and development stage. These incorporate safety improvements including features which will allow operators more time to remedy safety problems and which will provide greater assurance regarding containment of radioactivity in all circumstances. New plants will also be simpler to operate, inspect, maintain and repair, thus increasing their overall reliability and economy.
The new generation reactors:
The new designs fall into two broad categories: evolutionary and developmental. The evolutionary designs are those which are basically new models of existing, proven designs. The developmental designs depart more significantly from today¼s plants and require more testing and verification before large-scale deployment.
Table 8
ADVANCED THERMAL REACTORS
| Country and developer | Reactor | Size MWe | Design Process | Main Features |
| US-Japan (GE-Hitachi-Toshiba) | ABWR | 1300 | Commercial operation in Japan since 1996-7. In US: NRC final design certification 1997, FOAKE | |
| USA
(Westinghouse), South Korea | System 80+, APR (PWR) | 1300 1400 | NRC final design certification May 1997. Further developed for new S. Korean Shin Kori 3 & 4. | |
| USA (Westinghouse) | AP-600 AP-1000 (PWR) | 600 1000 | AP-600: NRC final design certification 1999, FOAKE | |
| Japan (utilities, Westinghouse, Mitsubishi) | APWR | 1500 | Basic design in progress, twin unit planned at Tsuruga | |
| France-Germany (Framatome ANP) | EPR (PWR) | 1550-1750 | Confirmed as future French standard, design completed 1997 | |
| Germany (Framatome ANP) | SWR (BWR) | 1000 | Under development | |
| Sweden (Westinghouse) | BWR 90+ | 1500 | Under development | |
| Russia (Atomenergoproject & Gidropress) | V-407 V-392 (PWR) | 640 1000 respectively | Construction of first V-407 unit pending, V-392 units planned |
|
| Russia (AEE) | VVER-91 (PWR) | 1000 | Two being built at Tianwan in China |
|
| Canada (AECL) | CANDU-9 | 925-1300 | Licensing approval 1997 | |
| Canada (AECL) | ACR | 700 1000 | Development to 2005. | |
| South Africa (Eskom, BNFL) | PBMR | 110 (module) | prototype due to start building in 2002 | |
| USA-Russia et al (General Atomics - Minatom) | GT-MHR | 285 (module) | Under development in Russia by multinational joint venture |
In USA, the Department of Energy and the commercial nuclear industry have developed three advanced reactor types.
Two of these fall into the category of large "evolutionary" designs which build directly on the experience of operating light water reactors in the United States, Japan and Western Europe. These reactors are in the 1300 megawatt range.
One is an advanced boiling water reactor (ABWR), two examples of which are in commercial operation in Japan, with two more under construction in Taiwan. The other type, System 80+, is an advanced pressurised water reactor (PWR), which is ready for commercialisation. Eight System 80 reactors in South Korea incorporate many design features of the System 80+, which is the basis of the Korean Next Generation Reactor program.
The US Nuclear Regulatory Commission (NRC) gave final design certification for both in May 1997, noting that they exceeded NRC "safety goals by several orders of magnitude". The ABWR has also been certified as meeting European requirements for advanced reactors.
Another, more innovative US advanced reactor is smaller - 600 MWe - and has passive safety features (its projected core damage frequency is nearly 1000 times less than today's NRC requirements). The Westinghouse AP-600 gained final design certification from the NRC in Dec 1999.
These NRC approvals are the first such generic certifications to be issued and will be valid for 15 years. As a result of an exhaustive public process, safety issues within the scope of the certified designs have been fully resolved and hence will not be open to legal challenge during licensing for particular plants. Utilities will be able to obtain a single NRC licence to both construct and operate a reactor before construction begins.
Separate from the NRC process and beyond its immediate requirements, the US nuclear industry has selected one standardised design in each category - the large ABWR and the medium-sized AP-600, for detailed first-of-a-kind engineering (FOAKE) work. The US$ 200 million program, was half funded by DOE. It means that prospective buyers now have firm information on construction costs and schedules.
The Westinghouse AP-1000, scaled-up from the AP-600, has now been submitted to the NRC for full design certification. It is under active consideration for building in UK and USA and is capable of running on a full MOX core if required. In Japan, the first two ABWRs have been operating since 1996 and are expected to have a 60 year life. Several more are under construction in Japan and also Taiwan.
A large (1500 MWe) advanced PWR is being developed by four utilities together with Westinghouse and Mitsubishi. It will be the basis for the next generation of Japanese PWRs. In addition, Mitsubishi is participating in development of Westinghouse's AP-1000 reactor.
In South Korea, the APR-1400 Advanced PWR design has evolved from the US System 80+ and has been known as the Korean Next-Generation Reactor.
In Europe, three designs are being developed to meet the European Utility Requirements (EUR) of French and German utilities, which have stringent safety criteria. The first two are ready for commercial deployment.
Framatome ANP is developing a large (1550 and up to 1750 MWe) European pressurised water reactor (EPR), which was confirmed in mid 1995 as the new standard design for France. It is derived from the French N4 and German Konvoi types.
Together with German utilities and safety authorities, Framatome ANP is also developing another evolutionary design, the SWR 1000, a 1000-1290 MWe BWR. The design was completed in 1999 and development continues, with US design certification being sought.
In Sweden, Westinghouse is developing its evolutionary BWR 90+ (1500 MWe) design, working with Scandinavian utilities to meet EUR requirements. In Russia, two advanced reactor designs have been developed. Both are advanced PWR with passive safety features.
Construction of the first 640 MWe V-407 (VVER-640) unit is ready to begin near St Petersburg. Four are planned for there and Kola. One or two 1000 MWe V-392 (advanced VVER-1000) units are planned for Novovoronezh. In addition, the VVER-91 (1000 MWe) has been developed with western control systems and two are being built in China at Jiangsu Tianwan.
Small floating nuclear power plants are also being developed.
Canada has had two designs under development which are based on its reliable CANDU-6 reactors. The first of two improved CANDU-6 units has started up in China. The CANDU-9 (925-1300 MWe) has been developed from an existing design but is larger and has flexible fuel requirements ranging from natural uranium through slightly-enriched uranium, recovered uranium from reprocessing spent PWR fuel, mixed oxide (U & Pu) fuel, direct use of spent PWR fuel, to thorium.
The Advanced Candu Reactor (ACR) is a more innovative concept, also derived from the CANDU-6. While retaining the low-pressure heavy water moderator, it incorporates some features of the pressurised water reactor, including low-enriched fuel. Adopting light water cooling and a more compact core reduces capital cost, and because the reactor is run at higher temperature and coolant pressure, it has higher thermal efficiency. It is moving towards design certification in Canada, USA and UK.
Building on the experience of several innovative reactors built in the 1960s and 1970s, new high-temperature gas-cooled reactors (HTRs) are being developed which will be capable of delivering high-temperature (up to 950&deeg;C) helium either for industrial application or directly to drive gas turbines for electricity (the Brayton cycle) with almost 48% thermal efficiency possible. Technology developed in the last decade makes HTRs more practical than in the past, though the direct cycle means that there must be high integrity of fuel and reactor components.
Fuel for these reactors is in the form of particles less than a millimetre in diameter. Each has a kernel of uranium oxycarbide, with the uranium enriched up to 9% U-235. This is surrounded by layers of carbon and silicon carbide, giving a containment for fission products which is stable to 2000°C.
There are two ways in which these particles are arranged: in blocks - hexagonal 'prisms' of graphite, or in billiard ball-sized pebbles of graphite encased in silicon carbide, each with about 15,000 fuel particles and 9g uranium. Both have a high level of inherent safety, including strong negative temperature coefficient whereby fission slows as temperature rises.
South Africa's Pebble Bed Modular Reactor (PBMR) is being developed by a consortium led by the utility Eskom, and drawing on German expertise. It aims for a step change in safety and economics. Modules with a direct-cycle gas turbine generator will be of 110 MWe and thermal efficiency about 45%. Up to 450,000 fuel pebbles recycle through the graphite-lined reactor continuously (about ten times each) until they are expended, giving an average enrichment in the fuel load of 5-6%. Each unit will finally discharge about 19 tonnes/yr of spent pebbles to ventilated on-site storage bins.
A larger US design, the Gas Turbine - Modular Helium Reactor (GT-MHR), will be built as modules of 285 MWe each directly driving a gas turbine at 48% thermal efficiency. It is being developed by General Atomics in partnership with Russia's Minatom, supported by Framatome ANP and Fuji (Japan). Initially it will be used to burn pure ex-weapons plutonium at Seversk in Russia. The development timeline is for a prototype to be constructed in Russia 2006-09 following regulatory review there now under way. Concurrently, the Russian design is being converted to US standards for a U-burning type and this could be available for construction from 2007, following NRC licensing.
HTRs can potentially use thorium-based fuels, such as HEU with Th, U-233 with Th, and Pu with Th. Most of the experience with thorium fuels has been in HTRs.
Fast neutron reactors are a different technology from those considered so far. They generate power from plutonium by much more fully utilising the uranium-238 in the reactor fuel assembly, instead of needing just the fissile U-235 isotope used in most reactors. If they are designed to produce more plutonium than they consume, they are called Fast Breeder Reactors (FBR). If they are net consumers of plutonium they are sometimes called „burners¾. For many years the focus has been on the potential of this kind of reactor to produce more fuel than they consume, but today, with low uranium prices and the need to dispose of plutonium from military weapons stockpiles, the main interest is in their role as incinerators.
Several countries have research and development programs for Fast Breeder Reactors (FBR), which are, generically, Fast Neutron Reactors. Over 290 reactor-years of operating experience has been gained on this type of plant. See Table 9.
In the closed fuel cycle it can be seen that conventional reactors produce two "surplus" materials; plutonium (from neutron capture, separated in reprocessing) and depleted uranium (from enrichment). The fast neutron reactor uses plutonium as its basic fuel while at the same time converting depleted (or natural) uranium, basically U-238, comprising a "fertile blanket" around the core, into fissile plutonium. In other words it "burns" and can "breed" plutonium*, as shown in Figure 13. Depending on the design, it is possible to recover from reprocessing the spent fuel enough fissile plutonium for its own needs, with some left over for future breeder reactors or for use in conventional reactors (see Figure 12).
* Both U-238 and Pu-240 are "fertile" (materials), i.e. by capturing a neutron they become (directly or indirectly) fissile Pu-239 and Pu-241 respectively.
Fast neutron reactors have a high thermal efficiency due to their high-temperature operation. Cooling is by liquid sodium. Although in many ways this is difficult to handle chemically, in some respects it is more benign overall than very high pressure water. Experiments on a 19 year old UK breeder reactor before it was decommissioned in 1977 showed that the liquid sodium cooling system made it less sensitive to coolant failures than the more conventional very high pressure water and steam systems in light water reactors. More recent operating experience with large French and UK prototypes has confirmed this.
Figure 12. The fast breeder fuel cycle
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The fast breeder reactor has the potential for utilising virtually all of the uranium produced from mining operations. As described in 3.2, overall about 60 times more energy can be extracted from the original uranium by the fast breeder cycle than can be produced by the current light water reactors operating in "open cycle". This extremely high energy efficiency makes the breeder an attractive energy conversion system. However, high capital costs and an abundance of low cost uranium means that they are unlikely to be competitive for several decades, probably not much before 2050.
For this reason design work on the 1450 MWe European FBR was phased out in 1994, although research at the 1250 MWe French Superphenix FBR took place 1995-98. Research continues on the Indian FBRs, to pave the way to greater use of thorium as a fuel, and a new 500 MWe FBR is under construction at Kalpakkam. Japan's Monju prototype commercial FBR was connected to the grid in August 1995 (but was then shut down due to a major sodium leak).
The Russian BN-600 fast breeder reactor has been supplying electricity to the grid since 1980 and has the best operating and production record of all Russia's nuclear power units. The BN-350 FBR operated in Kazakhstan for about 25 years and about half of its output was used for water desalination. Russia plans to reconfigure the BN-600 reactor to burn the plutonium from its military stockpiles, and construction has started on the first BN-800.
About 20 FBRs have already been operating, some since the 1950s.
Table 9
Fast Breeder Reactors
| reactor | MW (electrical) gross | MW (thermal) | Full operation | |
|---|---|---|---|---|
| USA | EBR 1 | 0.2 | 1951-63 | |
| EBR 2 | 20 | 1963-94 | ||
| Fermi 1 | 66 | 1963-72 | ||
| SEFOR | 20 | 1969-72 | ||
| Fast Flux Test Facility | 400 | 1980-93 | ||
| UK | Dounreay DFR | 15 | 1959-77 | |
| Dounreay PFR | 270 | 1974-94 | ||
| France | Rapsodie | 40 | 1966-82 | |
| Phenix * | 250 | 1973- | ||
| Superphenix 1 | 1240 | 1985-98 | ||
| Germany | KNK 2 | 21 | 1977-91 | |
| India | FBTR | 40 | 1985- | |
| Japan | Joyo | 100 | 1978- | |
| Monju | 246 | 1994-96 | ||
| Kazakhstan | BN 350* | 135 | 1972-99 | |
| Russia | BR 5 | 5 | 1959-71 | |
| BR 10 | 10 | 1971- | ||
| BOR 60 | 12 | 1969- | ||
| BN 600* | 600 | 1980- |
Figure 13. Fission in conventional and fast neutron reactors
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Contrast between conventional ("thermal") reactor and fast neutron reactor showing how typically more neutrons are produced in the fast reactor (17 instead of 15 from 6 fissions), thus enabling the system to breed more fissile material than is consumed if desired. In this example 4 neutrons are available for breeding Pu-239 in the conventional reactor, but 7 are available in the fast reactor. The exact numbers involved will depend on design and operation.
Near-breeder or thorium cycle reactors are similar to fast breeders in that a fertile material, naturally-occurring thorium-232, will absorb slow neutrons to become (indirectly) fissile uranium-233. This will produce a chain reaction yielding heat while surplus neutrons convert more thorium to U-233. The technology is considered by some to be attractive because plutonium production is avoided, fairly abundant thorium is used as a fuel, and the efficiency of fuel use approaches that of the fast breeder reactor. However, the amount of fissile uranium produced is not quite enough to sustain the reaction, hence the term "near-breeder" is generally used. Though a focus of interest for 30 years, only in India is any commercial outcome in sight.
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