Nuclear Fuel Fabrication
(Updated October 2013)
- Fuel fabrication is the final stage in nuclear fuel preparation prior to use in a reactor.
- Nuclear fuel assemblies are specifically designed for particular types of reactors and are made to exacting standards.
- Utilities and fabricators have collaborated to greatly improve fuel assembly performance.
Nuclear reactors are powered by fuel containing fissile material. The fission process releases large amounts of useful energy and for this reason the fissioning components – U-235 and/or Pu- 239 – must be held in a robust physical form capable of enduring high operating temperatures and an intense radiation environment. Fuel structures need to maintain their shape and integrity over a period of several years within the reactor core, thereby preventing the leakage of fission products into the reactor coolant.
The standard fuel form comprises a column of ceramic pellets of uranium oxide, clad and sealed into zirconium alloy tubes. For light water reactor (LWR) fuel, the uranium is enriched to various levels up to about 4.8% U-235. Pressurised heavy water reactor (PHWR) fuel is usually unenriched natural uranium (0.7% U-235), although slightly-enriched uranium is also used.
The fabrication of fuel structures – called assemblies or bundles – is the last stage of the front end of the nuclear cycle shown in Figure 1. The process for uranium-plutonium mixed oxide (MOX) fuel fabrication is essentially the same – notwithstanding some specific features associated with handling the plutonium component.
Figure 1: The closed nuclear fuel cycle, showing primary and recycled materials flow
Nuclear fuel fabrication – process overview
There are three main stages in the fabrication of the nuclear fuel structures used in LWRs and PHWRs:
- Producing pure uranium dioxide (UO2) from incoming UF6 or UO3.
- Producing high-density, accurately shaped ceramic UO2 pellets.
- Producing the rigid metal framework for the fuel assembly – mainly from zirconium alloy; and loading the fuel pellets into the fuel rods, sealing them and assembling the rods into the final fuel assembly structure.
These steps are illustrated in Figure 2.
Figure 2: The fuel fabrication process
UO2 powder production
Uranium arrives at a fuel manufacturing plant in one of two forms, uranium hexafluoride (UF6) or uranium trioxide (UO3), depending on whether it has been enriched or not. It needs to be converted to uranium dioxide (UO2) prior to pellet fabrication. Most fabrication plants have their own facilities for effecting this chemical conversion (some do not, and acquire UO2 from plants with excess conversion capacity). Chemical conversion to and from UF6 are distinct processes, but both involve the handling of aggressive fluorine compounds and plants may be set up to do both.
Conversion to UO2 can be done using ‘dry’ or ‘wet’ processes. In the dry method, UF6 is heated to a vapour and introduced into a two stage reaction vessel (eg, rotary kiln) where it is first mixed with steam to produce solid uranyl fluoride (UO2F2) – this powder moves through the vessel to be reacted with H2 (diluted in steam) which removes the fluoride and chemically reduces the uranium to a pure microcrystalline UO2 product.
Wet methods involve the injection of UF6 into water to form a UO2F2 particulate slurry. Either ammonia (NH3) or ammonium carbonate ((NH3)2CO3) is added to this mixture and the UO2F2 reacts to produce; ammonium diuranate (ADU, (NH3)2U2O7) in the first case, or ammonium uranyl carbonate (AUC, UO2CO3.(NH3)2CO3) in the latter case. In both cases the slurry is filtered, dried and heated in a reducing atmosphere to pure UO2. The morphology of UO2 powders deriving from the ADU and AUC routes are different, and this has a bearing on final pellet microstructure.
Wet methods are slightly more complex and give rise to more wastes, however, the greater flexibility in terms of UO2 powder properties is an advantage.
For the conversion of UO3 to UO2 , water is added to UO3 so that it forms a hydrate. This solid is fed (wet or dry) into a kiln operating with a reducing atmosphere and UO2 is produced.
Manufacture of ceramic UO2 pellets
The UO2 powder may need further processing or conditioning before it can be formed into pellets:
- Homogenization: powders may need to be blended to ensure uniformity in terms of particle size distribution and specific surface area.
- Additives: U3O8 may be added to ensure satisfactory microstructure and density for the pellets. Other fuel ingredients, such as lubricants, burnable absorbers (e.g. gadolinium) and pore-formers may also need to be added.
Conditioned UO2 powder is fed into dies and pressed biaxially into cylindrical pellet form using a load of several hundred MPa – this is done in pressing machines operating at high speed. These ‘green’ pellets are then sintered by heating in a furnace at about 1750°C under a precisely controlled reducing atmosphere (usually argon-hydrogen) in order to consolidate them. This also has the effect of decreasing their volume. The pellets are then machined to exact dimensions – the scrap from which being fed back into an earlier stage of the process. Rigorous quality control is applied to ensure pellet integrity and precise dimensions.
For most reactors pellets are just under one centimetre in diameter and a little more than one centimetre long. A single pellet in a typical reactor yields about the same amount of energy as one tonne of steaming coal.
Burnable absorbers (or burnable 'poisons') such as gadolinium may be incorporated (as oxide) into the fuel pellets of some rods to limit reactivity early in the life of the fuel. Burnable absorbers have a very high neutron absorption cross-section and compete strongly for neutrons, after which they progressively ‘burn-out’ and convert into nuclides with low neutron absorption leaving fissile (U-235) to react with neutrons. Burnable absorbers enable longer fuel life by allowing higher fissile enrichment in fresh fuel, without excessive initial reactivity and heat being generated in the assembly.
MOX pellets – see later section.
Manufacture and loading of the fuel assembly framework
Nuclear fuel designs dictate that the pellet-filled rods have a precise physical arrangement in terms of their lattice pitch (spacing), and their relation to other features such as water (moderator) channels and control-rod channels. The physical structures for holding the fuel rods are therefore engineered with extremely tight tolerances. They must be resistant to chemical corrosion, high temperatures, large static loads, constant vibration, fluid and mechanical impacts. Yet they must also be as neutron-transparent as possible.
Assembly structures comprise a strong framework made from steel and zirconium upon which are fixed numerous grid support pieces that firmly hold rods in their precise lattice positions. These are made from zirconium alloy and must permit (and even enhance) the flow of coolant water around the fuel rod. The grid structures grip the fuel rod and so are carefully designed to minimise the risk of vibration-induced abrasion on the cladding tube – called ‘fretting’ wear.
All fuel fabricators have highly sophisticated engineering processes and quality control for the timely manufacture of their assembly structures.
Pellets meeting QA specifications are loaded into tubes made from an appropriate zirconium alloy, referred to as the ‘cladding’. The filled tube is flushed with helium and pressurized with tens of atmospheres (several GPa) of this gas before the ends are sealed at each end by precision welding. A free space is left between the top of the pellet stack and the welded end-plugs – this is called the ‘plenum’ space and it accommodates thermal expansion of the pellets and some fission product gases. A spring is usually put into the plenum to apply a compressive force on the pellet stack and prevent its movement.
The completed fuel rods are then fixed into the prefabricated framework structures that hold the rods in a precisely defined grid arrangement.
In order to maximize the efficiency of the fission reaction the cladding and indeed all other structural parts of the assembly must be as transparent as possible to neutrons. Different forms of zirconium alloy, or zircaloy, are therefore the main materials used for cladding. This zircaloy includes small amounts of tin, niobium, iron, chromium and nickel to provide necessary strength and corrosion resistance. Hafnium, which typically occurs naturally with zirconium deposits, needs to be removed because of its high neutron absorption cross-section. The exact composition of the alloy used depends on the manufacturer and is an important determiner in the quality of the fuel assembly. Zircaloy oxidizes in air and water, and therefore it has an oxidized layer which does not impair function.
Rigorous quality control measures are employed at all process points in order to ensure traceability of all components in case of failures.
The major process safety concerns at nuclear fuel fabrication facilities are those of fluoride handling and the risk of a criticality event if insufficient care is taken with the arrangement of fissile materials. Both risks are managed through the rigorous control of materials, indeed, fuel fabrication facilities operate with a strict limitation on the enrichment level of uranium that is handled in the plant – this cannot be higher than 5% U-235, essentially eliminating the possibility of inadvertent criticality.
Types of nuclear fuel assemblies for different reactors
There is considerable variation among fuel assemblies designed for the different types of reactor. This means that utilities have limited choice in suppliers of fabricated fuel assemblies, especially for PWRs.
Pressurised water reactors (PWRs) are the most common type of nuclear reactor accounting for two-thirds of current installed nuclear generating capacity worldwide. A PWR core uses normal water as both moderator and primary coolant – this is kept under considerable pressure (about 10 MPa) to prevent it from boiling, and its temperature rises to about 330°C after its upward passage past the fuel. It then goes through massive pipes to a steam generator.
Fuel for western PWRs is built with a square lattice arrangement and assemblies are characterized by the number of rods they contain, typically, 17×17 in current designs. A PWR fuel assembly stands between four and five metres high, is about 20 cm across and weighs about half a tonne. The assembly has vacant rod positions – space left for the vertical insertion of a control rod. Not every assembly position requires fuel or a control rod, and a space may be designated as a "guide thimble" into which a neutron source rod, specific instrumentation, or a test fuel segment can be placed.
A PWR fuel assembly comprises a bottom nozzle into which rods are fixed through the lattice and to finish the whole assembly it is crowned by a top nozzle. The bottom and top nozzles are heavily constructed as they provide much of the mechanical support for the fuel assembly structure. In the finished assembly most rod components will be fuel rods, but some will be guide thimbles, and one or more are likely to be dedicated to instrumentation. A PWR fuel assembly is shown in Figure 3. PWR fuel assemblies are rather uniform compared with BWR ones, and those in any particular reactor must have substantially the same design.
An 1100 MWe PWR core may contain 193 fuel assemblies composed of over 50,000 fuel rods and some 18 million fuel pellets. Once loaded, fuel stays in the core for several years depending on the design of the operating cycle. During refueling, every 12 to 18 months, some of the fuel - usually one third or one quarter of the core – is removed to storage, while the remainder is rearranged to a location in the core better suited to its remaining level of enrichment.
Russian PWR reactors are usually known by the Russian acronym VVER. These fuel assemblies are characterized by their hexagonal arrangement, but are otherwise of similar length and structure to other PWR fuel assemblies.
Figure 3: Schematic view of PWR fuel assembly (Mitsubishi Nuclear Fuel)
Figure 4: A PWR fuel assembly
Figure 5: A VVER-1000 fuel assembly
Boiling water reactors (BWRs) are the second most common nuclear reactor type accounting for almost one-quarter of installed nuclear generating capacity. In a boiling water reactor water is turned directly to steam in the reactor pressure vessel at the top of the core and this steam (at about 290°C and 7 MPa) is then used to drive a turbine.
BWRs also use fuel rods comprising zirconium-clad uranium oxide ceramic pellets. Their arrangement into assemblies is again based on a square lattice, with pin geometries ranging from 6x6 to 10x10. Fuel life and management strategy is similar to that for a PWR.
But BWR fuel is fundamentally different from PWR fuel in certain ways: (i) Four fuel assemblies and a cruciform shaped control blade form a 'fuel module', (ii) each assembly is isolated from its neighbours by a water-filled zone in which the cruciform control rod blades travel (they are inserted from the bottom of the reactor), (iii) each BWR fuel assembly is enclosed in a zircaloy sheath which directs the flow of coolant water through the assembly and during this passage it reaches boiling point, (iv) BWR assemblies contain larger diameter water channels – flexibly designed to provide appropriate neutron moderation in the assembly.
The zircaloy tubes are allowed to fill with water thus increasing the amount of moderator in the central region of the assembly. Different enrichment levels are used in the rods in varying positions – lower enrichments in the outer rods, and higher enrichments near the centre of the bundle. A BWR reactor is designed to operate with 12-15% of the water in the top part of the core as steam, and hence with less moderating effect and thus efficiency there.
For many BWR models, control of reactivity to enable load-following can be achieved by changing the rate of circulation inside the core. Jet pumps located in the annulus between the outer wall of the vessel and an inner wall called the shroud increase the flow of water up through the fuel assembly. At high flow rates steam bubbles are removed more quickly, and hence moderation and reactivity is increased. When flow rate is decreased, moderation decreases as steam bubbles are present for longer and hence reactivity drops. This allows for a variation of about 25% from the maximum rated power output, enabling load-following more readily than with a PWR.
Control rods are used when power levels are reduced below 75%, but they are not part of the fuel assembly as in a PWR. They are bottom-entry – pushed upwards so that rods intercept the lower, more reactive, zone of the fuel assemblies first.
BWR fuel fabrication takes place in much the same way as PWR fuel.
A cross sectional diagram of a BWR assembly is shown in Figure 6. BWR fuel assemblies therefore operate more as individual units, and different designs may be mixed in any core load, giving more flexibility to the utility in fuel purchases.
Figure 6: Schematic view of BWR fuel assembly (Nucleartourist and GE)
PHWR (CANDU) fuel
Pressurised Heavy Water Reactors (PHWRs) are originally a Canadian design (also called “CANDU”) accounting for ~6% of world installed nuclear generating capacity. PHWRs use pressure tubes in which heavy water moderates and cools the fuel. They run on natural (unenriched) or slightly-enriched uranium oxide fuel in ceramic pellet form, clad with zirconium alloy.
PHWR fuel rods are about 50 cm long and are assembled into ‘bundles’ approximately 10 cm in diameter. A fuel bundle comprises 28, 37 or 43 fuel elements arranged in several rings around a central axis (see Figure). Their short length means that they do not require the support structures characteristic of other reactor fuel types. PHWR fuel does not attain high burn-up, nor does it reside in the reactor core for very long and so the fuel pellets swell very little during their life. This means that PHWR fuel rods do not need to maintain a pellet-cladding gap, nor be highly pressurized with a filling gas (as for LWR fuel), indeed, the metal cladding is allowed to collapse onto the fuel pellet thereby assuring good thermal contact.
The fuel bundles are loaded into horizontal channels or pressure tubes which penetrate the length of the reactor vessel (known as the calandria), and this can be done while the reactor is operating at full power. About twelve bundles are loaded into each fuel channel depending on the model – a 790 MWe CANDU reactor contains 480 fuel channels composed of 5,760 fuel bundles containing over 5 million fuel pellets.
The on-load refueling is a fully-automated process: new fuel is inserted into a channel at one end and used fuel is collected at the other. This feature means that the PHWR is inherently flexible with its fuel requirements, and can run on different fuels requiring different residence times, eg natural uranium, slightly enriched uranium, plutonium-bearing fuels and thorium-based fuels.
Figure 7: Indian PHWR fuel bundles
The Advanced Gas-Cooled Reactor (AGR) is a second-generation UK-designed nuclear reactor only used in UK. AGRs account for about 2.7% of total global nuclear generating capacity. They employ a vertical fuel channel design, and use CO2 gas – a very weak moderator – as the primary coolant.
AGR fuel assemblies consist of a circular array of 36 stainless steel clad fuel pins each containing 20 enriched UO2 fuel pellets, and the assembly weighs about 43 kilograms. Enrichment levels vary up to about 3.5%. Stainless steel allows for higher operating temperatures but sacrifices some neutron economy. The assembly is covered with a graphite sheath which serves as a moderator. Eight assemblies are stacked end on end in a fuel channel, inserted down through the top of the reactor. During refueling this whole stack is replaced. Fuel life is about five years, and refueling can be carried out on-load through a refueling machine.
Figure 8: cutaway of an AGR fuel assembly
The RBMK reactor is an early Soviet design, developed from plutonium production reactors. Eleven units are in operation (3% of world total), with control systems and fuel greatly modified since 1990. It employs vertical pressure tubes (just under1700 of these, each about 7 metres long) running through a large graphite moderator. The fuel is cooled by light water water, which is allowed to boil in the primary circuit, much as in a BWR.
RBMK fuel rods are about 3.65 metres long, and a set of 18 forms a fuel bundle about 8 cm diameter. Two bundles are joined together and capped at either end by a top and bottom nozzle, to form a fuel assembly with an overall length of about 10 metres, weighing 185 kilograms. Since 1990 RBMK fuel has had a higher enrichment level, increasing from about 2% to average 2.8% (varying along the fuel element from 2.5% to 3.2%) and it now includes about 0.6% erbium (a burnable absorber). This has the effect of improving overall safety and increasing fuel burn-up. This new fuel can stay in the reactor for periods of up to six years before needing to be removed. All RBMK reactors now use recycled uranium from VVER reactors.
As with other pressure tube designs such as the PHWR, the RBMK reactor is capable of on-load refueling.
Fast neutron reactor fuel
There is only one commercially operating fast reactor (FNRs) in service today – the BN-600 at Beloyarsk in Russia. There are two FNRs under construction – a 800 MWe unit in Russia and a 500 MWe unit in India (which expects to build five more). Two BN-800 units are planned in China.
Fast neutron reactors (FNR) are unmoderated and use fast neutrons to cause fission. Hence they mostly use plutonium as their basic fuel, or sometimes high-enriched uranium to start them off (they need about 20-30% fissile nuclei in the fuel). The plutonium is bred from U-238 during operation. If the FNR is configured to have a conversion ratio above 1 (ie more fissile nuclei are created than fissioned) as originally designed, it is called a Fast Breeder Reactor (FBR). FNRs use liquid metal coolants such as sodium and operate at higher temperatures. (See also Fast Neutron Reactor information paper)
Apart from the main FNR fuel, there are numerous heavy nuclides - notably U-238, but also Am, Np and Cm that are fissionable in the fast neutron spectrum – compared with the small number of fissile nuclides in a slow (thermal) neutron field (just U-235, Pu-239 and U-233). A FNR fuel can therefore include a mixture of transuranic elements. Also it can be in one of several chemical forms, including; standard oxide ceramic, mixed oxide ceramic (MOX), single or mixed nitride ceramics and metallic fuels. Further, FNR fuel can be fabricated in pellet form or using the ‘vibro-pack’ method in which graded powders are loaded and compressed directly into the cladding tube.
The core of an FNR is much smaller that a conventional reactor, and cores tend to be designed with distinct ‘seed’ and ‘blanket’ regions according to whether the reactor is to be operated as a ‘burner’ or a ‘breeder’. In each case the fuel composition for the seed and blanket regions are different – the central seed region uses fuel with a high fissile content (and thus high power and neutron emission level), and the blanket region has a low fissile content but a high level of neutron absorbing material which can be fertile (for a breeding design, eg U-238) or a waste absorber to be transmuted.
BN-600 fuel assemblies are 3.5 m long, 96 mm wide, weigh 103 kg and comprise top and bottom nozzles (to guide coolant flow) and a central fuel bundle. The central bundle is a hexagonal tube and for seed fuel houses 127 rods, each 2.4 m long and 7 mm diameter with ceramic pellets in three uranium enrichment levels; 17%, 21% and 26%. Blanket fuel bundles have 37 rods containing depleted uranium. BN-600 fuel rods use low-swelling stainless steel cladding.
FNRs use liquid metal coolants such as sodium or a lead-bismuth eutectic mixture and these allow for higher operating temperatures – about 550°C, and thus have higher energy conversion efficiency. They are capable of high fuel burn-up.
Nuclear fuel performance in reactors
Nuclear fuel operates in a harsh environment in which high temperature, chemical corrosion, radiation damage and physical stresses may attack the integrity of a fuel assembly. The life of a fuel assembly in the reactor core is therefore regulated to a burn-up level at which the risk of its failure is still low. Fuel ‘failure’ refers to a situation when the cladding has been breached and radioactive material leaks from the fuel ceramic (pellet) into the reactor coolant water. The radioactive materials with most tendency to leak through a cladding breach into the reactor coolant are fission-product gases and volatile elements, notably; krypton, xenon, iodine and cesium.
Fuel leaks do not present a significant risk to plant safety, though they have a big impact on reactor operations and (potentially) on plant economics. For this reason, primary coolant water is monitored continuously for these species so that any leak is quickly detected. The permissible level of released radioactivity is strictly regulated against specifications which take into account the continuing safe operation of the fuel. Depending on its severity a leak will require different levels of operator intervention:
- Very minor leak: no change to operations – the faulty fuel assembly with leaking rod(s) is removed at next refueling, inspected, and possibly re-loaded.
- Small leak: allowable thermal transients for the reactor are restricted. This might prevent the reactors from being able to operate in a “load-follow” mode and require careful monitoring of reactor physics. The faulty fuel assembly with leaking rod is generally removed and evaluated at the next scheduled refueling.
- Significant leak: the reactor is shut down and the faulty assembly located and removed.
A leaking fuel rod can sometimes be repaired but it is more usual that a replacement assembly is needed (this having a matching level of remaining enrichment). Replacement fuel is one cost component associated with failed fuel. There is also the cost penalty and/or replacement power from having to operate at reduced power or having an unscheduled shutdown. There may also be higher operation and maintenance costs associated with mitigating increased radiation levels in the plant.
Fuel management is a balance between the economic imperative to burn fuel for longer and the need to keep well within failure-risk limits. Improving fuel reliability extends these limits, and therefore is a critical factor in providing margin to improve fuel burn-up.
The nuclear industry has made significant performance improvements reducing fuel failure rates by about 60% in the 20 years to 2006 to an average of some 14 leaks per million rods loaded [IAEA 2010]. The reliability drive continues. Industry-wide programs led by EPRI and the US Institute of Nuclear Power Operations (INPO) have produced guidelines to help eliminate fuel failures (there was an ambitious goal to achieve zero fuel failures by 2010). Fuel engineering continues to improve, eg, with more sophisticated debris filters in assembly structures. Utilities themselves impose more rigorous practices to exclude foreign material entering primary coolant water. Global Nuclear Fuels (GNF) in 2013 had 2 million fuel rods in operation and claimed to have no leakers among them. (In the early 1970s hydriding and pellet-clad interaction caused a lot of leaks. The 1980s saw an order of magnitude improvement.)
At the same time there has been a gradual global trend toward higher fuel burnup*, however, there is a limit on how far fuel burnup can be stretched given the strict criticality safety limitation imposed on fuel fabrication facilities such that a maximum uranium enrichment level of 5% can be handled.
* Higher burnup does not necessarily mean better energy economics. Utilities must carefully balance the benefits of greater cycle length against higher front-end fuel costs (uranium, enrichment). Refueling outage costs may also be higher, depending on length, frequency and the core re-load fraction.
An equally important trend in nuclear fuel engineering is to be able to increase the power rating for fuels, ie, how much energy can be extracted per length of fuel rod. Currently this is limited by the material properties of the zirconium cladding.
Fabrication supply and demand
The current annual demand for LWR fuel fabrication services is expressed as a requirement for about 7000 tonnes of enriched uranium being made into assemblies, and this is expected to increase to about 9500 t by 2020. Requirements for PHWRs account for an additional 3,000 t/year and the Gas-cooled Reactor market for around 400 t/year.
Requirements for fuel fabrication services will grow roughly in line with the growth in nuclear generating capacity. However, fabrication requirements are also affected by changes in utilities’ reactor operating and fuel management strategies, which are partly driven by technical improvements in fuel fabrication itself. For example, LWR discharge burn-ups have increased steadily as improvements in fuel design have made this possible, and this has tended to reduce fabrication demand, as fuel remains in the reactor for a longer period (though there is a limit to how far burn-ups can be pushed without tackling the 5% enrichment limit in place for criticality safety margins at fuel fabrication plants). A parallel industry-wide focus on increasing fuel performance and reliability has also decreased the demand for fuel to replace defective assemblies.
Plans to build many new reactors affect the demand on fabrication capacity in two ways. The demand for reloads increases in line with the new installed reactor capacity, typically 16 to 20 tonnes per year per GW. Additionally the first cores create a temporary peak demand, since the amount required is about three to four times that of a reload batch in currently operated LWRs (and some of the enrichment is less). An average first core enrichment is about 2.8%.
Provision of fuel fabrication worldwide
Fuel fabrication services are not procured in the same way as, for example, uranium enrichment is bought. Nuclear fuel assemblies are highly engineered products, made to each customer’s individual specifications. These are determined by the physical characteristics of the reactor, by the reactor operating and fuel cycle management strategy of the utility as well as national, or even regional, licensing requirements.
Most of the main fuel fabricators are also reactor vendors (or owned by them), and they usually supply the initial cores and early reloads for reactors built to their own designs. However the market for LWR fuel has become increasingly competitive and for most fuel types there are now several competing suppliers – most notably perhaps, Russian fabricator TVEL competes to manufacture Western PWR fuel, and Western fuel fabricators can manufacture VVER fuel.
Currently, fuel fabrication capacity for all types of LWR fuel throughout the world considerably exceeds the demand. It is evident that fuel fabrication will not become a bottleneck in the foreseeable supply chain for any nuclear renaissance.
LWR fuel fabrication capacity worldwide is shown in Table 1. The back-conversion capacities are particularly unevenly distributed. For some fabricators it represents a bottleneck. Some fabricators do not have conversion facilities at all and have to buy such services in the market, while others with excess capacity are even sellers of UO2 powder.
Table 1: World LWR fuel fabrication capacity, tonnes/yr
||DAE Nuclear Fuel Complex
|Mitsubishi Nuclear Fuel
* Includes approx. 220 tHM for RBMK reactors
** Includes approx. 200 tHM for AGR reactors
Source: WNA Market Report 2013. NB the above figures are about 40% above operational capacities, which meet demand.
Table 2: World PHWR fuel fabrication capacity, tonnes/yr
||Cordoba & Eizeiza
||CNNC China Northern
||DAE Nuclear Fuel Complex
Source: WNA Market Report 2013, from IAEA.
The LWR fuel fabrication industry has rationalized in recent years, including:
- When Westinghouse Electric was bought by Toshiba, Kazatomprom acquired a 10% share of this,
- Global Nuclear Fuels was formed as a joint venture between General Electric, Toshiba and Hitachi. There are two ‘branches’ GNF-A (US) and GNF-J (Japan) with different ownership structures.
- Westinghouse has purchased 52% of Nuclear Fuel Industries (NFI) in Japan, with the remainder being held by Sumitomo (24%) and Furukawa (24%).
- Mitsubishi Heavy Industries and AREVA bought into Mitsubishi Nuclear Fuel and created a US fuel fabrication joint venture.
- Kazatomprom and AREVA agreed to build a 1200 t/yr fuel fabrication plant in Kazakhstan.*
Secondary supply from recycle
Currently about 100 t/yr year of reprocessed uranium (RepU) is produced at MSZ in Elektrostal, Russia for AREVA contracts. One production line in AREVA’s plant in Romans, France is licensed to fabricate 150 t of RepU into fuel per year and PWR assemblies of this type have already been delivered to French, Belgian and UK reactors, and an amount of RepU powder has been sent from Russia to Japan.
At present, nearly all commercial MOX fuel is fabricated in AREVA’s MELOX plant in Marcoule. With a capacity of 195 tonnes/yr and a good production rate this plant helps not only to save uranium and enrichment demand, but also frees-up LWR fabrication capacity in the market.
The UK’s Sellafield MOX plant had a designed capacity of 120 t/yr but was downgraded to 40 t/yr and never reached that level of reliable output before being closed down in 2011. Japan's Rokkasho-Mura MOX plant is planned to be operational by 2015, and the US MOX Plant in Savannah River is due to produce MOX fuel from weapons plutonium from 2018.
The MOX fuel market has weakened somewhat recently with cessation of its use in Belgium, Germany and Switzerland (moratorium), and the continued loading of MOX fuel in Japan is unclear in the aftermath of the Fukushima accident.
Table 3: World MOX fuel fabrication capacity, tonnes/yr
||DAE Nuclear Fuel Complex
* Operational by 2016
** Operational by 2018
Source: WNA Market Report 2013
Mixed uranium oxide + plutonium oxide (MOX) fuel has been used in about 30 light-water power reactors in Europe and about ten in Japan. It consists of depleted uranium (about 0.2% U-235), large amounts of which are left over from the enrichment of uranium, and plutonium oxide that derives from the chemical processing of used nuclear fuel (at a reprocessing plant). This plutonium is reactor-grade, comprising about one third non-fissile isotopes.
In a MOX fuel fabrication plant the two components are vigorously blended in a high-energy mill which intimately mixes them such that the powder becomes mainly a single ‘solid solution’ (U,Pu)O2. MOX fuel with about 7% of rector-grade plutonium is equivalent to a typical enriched uranium fuel. The pressing and sintering process is much the same as for UO2 fuel pellets, though some plastic shielding is needed to protect workers from spontaneous neutron emissions from the Pu-240 component.
Vibropacked MOX (VMOX) fuel is a Russian variant for MOX fuel production in which blended (U,Pu)O2 and UO2 powders are directly loaded and packed into cladding tubes where they sinter in-situ under their own operating temperature. This eliminates the need to manufacture pellets to high geometric tolerances, which involves grinding and scrap which are more complex to deal with for Pu-bearing fuels. Russian sources say vibropacked fuel is more readily recycled.
High Temperature Reactor Fuel
High Temperature Reactors operate at 750 to 950°C, and are normally helium-cooled. Fuel for these is in the form of TRISO (tristructural-isotropic) particles less than a millimetre in diameter. Each has a kernel (ca. 0.5 mm) of uranium oxycarbide (or uranium dioxide), with the uranium enriched up to 20% U-235, though normally less. This is surrounded by layers of carbon and silicon carbide, giving a containment for fission products which is stable to very high temperatures. Recent trials at two US laboratories confirmed that most fission products remain securely in TRISO particles up to about 1800°C – the performance being much better than previously known.
There are two ways in which these particles can be arranged in a HTR: 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. Either way, the moderator is graphite. HTRs can potentially use thorium-based fuels, such as high-enriched or low-enriched uranium with Th, U-233 with Th, and Pu with Th. Most of the experience with thorium fuels has been in HTRs.
The main HTR fuel fabrication plant is at Baotou in China, the Northern Branch of China Nuclear Fuel Element Co Ltd. From 2015 this will make 300,000 fuel pebbles per year for the HTR-PM under construction at Shidaowan. Previous production has been on a small scale in Germany.
Advanced Nuclear Fuel Technology Directions
Fuel development activities in the nuclear industry have largely focussed on improving the reliability of standard zirconium-clad uranium oxide fuels. Increasingly, however, R&D effort is being applied to evolutionary fuel forms that can offer significant improvements in terms of safety, waste management and operating economics.
Fuel technologies that seem particularly promising, and which could be commercially deployed in the forseeable future include:
- Ceramic or coated zirconium claddings that prevent the adverse interaction between steam and zironium at very high temperature
- High thermal conductivity oxide fuel, such as can be achieved by including additives like beryllium oxide (BeO). Higher conductivity provides higher safety margins and can allow higher operating powers.
- Thoria-based fuels, including mixed thorium-plutonium (Th-MOX) fuel which can achieve a high utilization factor for recycled plutonium.
- All-metal fuels and annular LWR fuel, allowing more cooling and therefore safe, high power densities for the fuel and improved economics. Babcock & Wilcox Nuclear Energy is working with Lightbridge to set up a pilot plant for metal fuel which is 50:50 Zr-U alloy, with uranium enriched to 20%.
- Pelletized coated-particle fuels, aimed at achieving high safety levels for the fuel that can be left in a light water reactor for very long periods, thereby achieving high burn-up of recycled plutonium and/or actinide waste components.
Work is underway on each of these new fuel technologies.
WNA 2009, The Global Nuclear Fuel Market ('market report')
Kok, Kenneth (ed), 2009, Nuclear Engineering Handbook (ch 2, 3, 4, 9), CRC Press
Nuclear Engineering International, Sept 2010, Fuel Design Data
Merrifield, J.S., 2005, Briefing of the U.S. Nuclear Regulatory Commission on Nuclear Fuel Performance, No. S-05-002 http://www.nrc.gov/reading-rm/doc-collections/commission/speeches/2005/s-05-002.html
Kushner, M. P. 1973, Nuclear Fuel Fabrication For Commercial Electric Power Generation, IEEE
Springfields Fuels – Technology and Capabilities http://www.westinghousenuclear.com/ProductLines/Nuclear_Fuel/springfields_site.shtm
Mitsubishi Nuclear Fuel Co. http://www.mnf.co.jp/pages2/pwr2.htm
Cameco Uranium Science, Nuclear Fuel Cycle http://www.cameco.com/uranium_101/uranium_science/nuclear_fuel/#seven
WNN, Joint efforts for new fuel plants, October 2010 http://www.world-nuclear-news.org/C-Joint_efforts_for_new_fuel_plants-2810107.html
WNN, Areva, Mitsubishi form fuel fabrication joint venture, February 2009 http://www.world-nuclear-news.org/newsarticle.aspx?id=24690
WNN, One-stop fuel shop coming for Asia, October 2009 http://www.world-nuclear-news.org/newsarticle.aspx?id=26256
WNN, Westinghouse buys into Japanese fuel maker, April 2009 http://www.world-nuclear-news.org/newsarticle.aspx?id=25144
WNN, Westinghouse rounds up tech, fuel and supply chain, January 2011 http://www.world-nuclear-news.org/C_Westinghouse_rounds_up_tech_fuel_and_supply_chain_1901111.html
WNN, Decision soon on new UK MOX plant, January 2011 http://www.world-nuclear-news.org/newsarticle.aspx?id=29109
Kamagin, D., 2003, Modification and Improvement of RBMK-1500 Fuel Assembly Design, Ignalina Youth Nuclear Association report
For more on fuel fabrication technical aspects, essentially welding of zircaloy: http://www.antinternational.com/fileadmin/Products_and_handbooks/IZNA/First_chapter_IZNA_7_STR_Weld.pdf