Spent fuel reprocessing is most usually performed using the well proven hydrometallurgical PUREX (plutonium uranium refining by extraction) process.

Several organic compounds were tested and eventually tri-butyl-phosphate (TBP), diluted in an alkane solvent, was selected. TBP is an organo-phosphorus compound with the formula (CH3CH2CH2CH2O)3PO.

CH3-CH2-CH2-CH2-0

CH3-CH2-CH2-CH2-0-P = о

/

CH3-CH2-CH2-CH2-0

16.3.1 Description of the main steps of the PUREX process

The main steps (Fig. 16.2) include:

• Head operations leading to the dissolution of the fuel. These comprise reception at the reprocessing plant, storage, shearing, dissolution, clarification (Section 16.3.2).

• Operations for separating uranium and plutonium from fission products followed by uranium/plutonium partition (Section 16.3.3).

• Purification of plutonium and uranium (in order to comply with the required purity for end products) and, possibly, conversion into oxides.

• Ancillary operations for conditioning solid wastes and effluent treatments.

With respect to the transmutation of FPs, much less work has been performed in the last decades. In part this is because, compared to the transuranics, their radiotoxicity is lower and shorter lived so that, after about 250 years, most of them have decayed. Nevertheless, some FPs are very long lived and may be mobile in the environment so that, often, they are major contributors to the very long-term radiological impact of deep geological disposal. The long-lived FPs (LLFPs) that deserve most attention in this respect are 2 29I, 135Cs, 79Se, 99Tc and 126Sn (see Table 17.5).

Reactor type PWR FR ADS ‘ч Fuel type Parameter MOX (Pu only, reference) Homog TRU recycle Pu only Homog. TRU recycle, CR = 1 and MA/Pu ~0.1 Homog. TRU recycle, CR = 0.5 and MA/Pu ~0.1 Homog. TRU recycle, CR = 0.5 and МА/ Pu ~1 MA targets (Heterog. recycle) U-free fuel CR = 0 and МА/ Pu ~1 Decay heat 1 x3 x0.5 x2.5 x12 x38 x40 x100 Neutron source 1 x8000 ~1 x150 x1000 x4000 x5000 x20000

512 Nuclear fuel cycle science and engineering Table 17.5 Properties of the main long-lived FPs (LLFPs)

Unlike transuranics, FP transmutation does not produce supplementary neutrons as fission products are neutron consumers. LLFP transmutation therefore requires a large neutron surplus to be available. If this is expressed in terms of the fraction of a reactor fleet that would be needed to perform the LLFP transmutation, the value comes out at between 8 and 15%,23 i. e., a very large (and, very probably, unrealistic) figure.

Finally, it should be emphasized that the short-term heat production is essentially related to 90Sr and 1 37Cs, which are not candidates for transmutation because their relatively short half-life (~30 years) is such that no transmutation process can provide a comparable ‘transmutation half-life’, (evaluated as the product of the microscopic capture cross section of the isotope considered times the neutron flux). One way to handle these two isotopes could be to recover them with an appropriate chemical process, then store them and let them decay.

In summary, the transmutation of the so-called LLFPs is no longer envisioned by any major international program.

Many near-surface disposal facilities for radioactive waste exist throughout the world[31] — some having operated for decades whilst a few are now in post-closure institutional control. Designs vary depending on the type of waste to be accepted and the environmental conditions at the disposal site, especially the potential for water to come into contact with the waste. In broad terms we expect the number of engineered barriers to increase with the hazard presented by the waste and, similarly, as we move from dry to temperate and tropical climates. In all cases the overall aim is to contain and isolate the waste so that radionuclides do not enter the human environment.

With respect to containment, a crucial element is the avoidance of water contact with the waste. In a naturally dry environment there may be no need for any more barriers than those offered by the waste and its package but in wetter climates a repository cap can provide the necessary protection against infiltration of rainwater. Where the groundwater table is close to the surface, facilities are normally created above grade. Otherwise, below grade facilities have a number of advantages including less need for lateral support for the walls, simpler operation and a reduced susceptibility to erosion because of the lower profile (Fig. 18.3). For VLLW in a humid environment, the provision of a cap may be enough to supply the needed level of containment. For LILW, however, it is usual to install additional engineered barriers in the form of concrete, which is used to immobilise the waste within the package, to fill the gaps between waste packages in the waste stack and to provide the engineered structures (walls, base and cap) of the facility.

So far as isolation is concerned, the extensive use of concrete will help to reduce erosion and make the structure more resistant to inadvertent human intrusion. But in near-surface disposal, by far the most important element in this respect is institutional control, which should be considered as an engineered barrier in its own right. While there may be an intention to maintain everlasting institutional control over a facility, this is too bold a claim to be used in a safety case. Based on the longevity of existing human institutions and allowing some margin for conservatism, a commonly claimed institutional control period is

300 years. After that period, the safety case must assume that control will be lost and humans may reoccupy the site. To obtain approval for a proposed site, the safety case will need to demonstrate that this can be done safely. A radionuclide that is often prevalent in LILW is caesium-137. This has a half-life of about 30 years, which means that a 300-year control period will produce a decrease in radioactivity of about 1000 times (210). Let us suppose that humans could safely live on a site where specific activities were at exemption levels. For caesium-137, this is 10 Bq/g.14 Let us further allow that mixing of the waste with uncontaminated material (from the cap for instance) reduces the specific activity of the resulting soil by a factor of ten. In that case, the maximum permissible specific activity of Cs-137 in a waste for disposal would be about 100 000 Bq/g (100 kBq/g). This is of the same order of magnitude as the value allowed for near-surface disposal of caesium-137 in France, which is 330 kBq/g.28

Examples of the silo design (Fig. 18.3) may be found in Sweden and Finland, where they are constructed at a depth of about 100 m. A more recent development is one proposed to be built at Vrbina in Slovenia. This is about 12 m below the surface with a maximum depth of more than 50 m.29 The greater depth of disposal that is available with the silo design provides more protection from human intrusion and may thus enable the disposal of LILW.

Uranium is a naturally occurring element with an average concentration of 2.8 parts per million in the Earth’s crust. Traces of it occur almost everywhere. It is more abundant than gold, silver or mercury, about the same as tin, and slightly less abundant than cobalt, lead or molybdenum. Vast amounts of uranium also occur in the world’s oceans, but in very low concentrations.

Uranium mines operate in some twenty countries, though 55% of world production comes from just ten mines in six countries, these six providing 85% of the world’s mined uranium (Fig. 6.1). Most of the uranium ore deposits at present supporting these mines have average grades in excess of 0.10% of uranium — that is, greater than 1000 parts per million. In the first phase of uranium mining to the 1960s, this would have been seen as a respectable grade, but today some Canadian mines have huge amounts of ore up to 20% U average grade. Other mines, however, can operate successfully with very low grade ores, down to about 0.02% U.

Some uranium is also recovered as a by-product with copper, as at Olympic Dam mine in Australia, or as a by-product from the treatment of other ores, such as the gold-bearing ores of South Africa, or from phosphate deposits such as in

Morocco and Florida. In these cases the concentration of uranium may be as low as a tenth of that in orebodies mined primarily for their uranium content. An orebody is defined as a mineral deposit from which the mineral may be recovered at a cost that is economically viable given the current market conditions. Where a deposit holds a significant concentration of two or more valuable minerals then the cost of recovering each individual mineral is reduced as certain mining and treatment requirements can be shared. In this case, lower concentrations of uranium than usual can be recovered at a competitive cost.

Generally speaking, uranium mining is no different from other kinds of mining unless the ore is very high grade. In this case special mining techniques such as dust suppression, and in extreme cases remote handling techniques, are employed to limit worker radiation exposure and to ensure the safety of the environment and general public.

Searching for uranium is in some ways easier than for other mineral resources because the radiation signature of uranium’s decay products allows deposits to be identified and mapped from the air.

Thorium is a possible alternative source of nuclear fuel, but the technology for using this is not established. Thorium requires conversion to a fissile isotope of uranium actually in a nuclear reactor (see Chapter 9) . However, supplies of thorium are abundant, and the element currently has no commercial value. Accordingly, the amount of resource is estimated rather than directly measured as with uranium. Thorium is reported to be about three times as abundant in the Earth’s crust as uranium. The 2009 IAEA-NEA ‘Red Book’ lists 3.6 million tonnes of known and estimated resources as reported, but points out that this excludes data from much of the world, and estimates about 6 million tonnes overall. Making normal assumptions regarding how it might be used, this represents a far greater energy source than the same amount of uranium used in today’s reactors, but about the same if fast neutron reactors are envisaged.

Thorium has been considered as a complement to uranium-based fuel since the earliest days of the nuclear industry, initially based on considerations of resource utilization, and more recently also as a result of concerns about proliferation and waste management. These last two points will be further developed later in this chapter as well as the primary incentive for considering the thorium cycle in the past: potential uranium savings.

In summary, reasons for considering the introduction of a thorium-based fuel cycle include:

• increasing fissile resources by breeding U-233 from thorium

• improving fissile fuel utilization in thermal reactors

• significantly reducing U-235 enrichment requirements

• decreasing production of plutonium and other transuranic (TRU) elements compared to the uranium fuel cycle

• advantageous neutronic and physical properties of thorium-based fuel (e. g., higher thermal conductivity, higher melting point, better behaviour under irradiation, higher burn-up achievable)

Possible disadvantages are associated with the handling of U-233 fuel and the reprocessing of thorium-based fuel/blankets next to the need to deploy a parallel nuclear fuel cycle to the U/Pu fuel cycle when deploying a full Th/U-233 fuel cycle. These topics will be further developed in this chapter.

Fuel assembly designs and their improvements as well as fuel behaviour are comprehensively reported in international conference series and their proceedings. The yearly recurring TopFuel/Water Reactor Fuel Performance meetings are organised by national and international nuclear societies and provide overviews of current issues and developments as well as detailed research papers. Some proceedings can be downloaded from the internet, e. g., the 2006 issue from www. euronuclear. org.

The conference series ‘Zirconium in the Nuclear Industry’ is organised approximately every other year and covers all aspects of zirconium alloy applications and their behaviour in the reactor core. The proceedings are available through the ASTM bookstore www. astm. org/Standard/books_journals.

Reactor physics affecting fuel assembly design, associated tools and methods are covered by the biennial conferences on ‘Physics of Reactors’ (Physor). The proceedings of some recent meetings are available through the ANS bookstore (www. new. ans. org/store).

Finally, a book that is both informative and entertaining is Canada Enters the Nuclear Age. It contains many details of CANDU assembly and fuel designs for which there was no room in this chapter (Atomic Energy of Canada Limited, 1997; ISBN 0-7735-1601-8).

The Magnox reactor type was the first gas-cooled reactor to be produced in any quantity. However, almost all designs used on individual sites were unique. This was in part no doubt due to the rapidly developing technology of nuclear power and also the fact that there were a number of different consortia involved in the construction. This approach resulted in a lack of economy of scale such as that seen in the building of series of PWR and BWR plants in France and the USA.

11.3.1 Main plant features

The majority of Magnox reactors constructed were of the steel pressure vessel (SPV) type. The exceptions are Oldbury and Wylfa (UK) and Chinon A3, St Laurent des Eaux and Bugey 1 (France), which had concrete pressure vessels.

Steel pressure vessel reactors

The general layout of most SPV reactors was of a spherical vessel of diameter typically ~20 m containing the reactor core, with typically six hot and six cold ducts leading to external vertical heat exchangers. However, the early reactors (Chapelcross, Calder Hall and Berkeley) had cylindrical pressure vessels. In those early designs each of the heat exchangers was in a separate building, fed with external, exposed ductwork. Later designs saw the heat exchangers housed in the same building as the reactor and bioshield, giving notionally better protection from external impact events.

The material used was essentially a mild steel, of thickness in the range 50-100 mm. The construction was by welding plates in situ, with the reactor pressure vessel (RPV) being supported either on a cylindrical skirt or on a system of rollers to allow for expansion movement.

The cylindrical graphite core, typically of around 15 m diameter, was mounted on support plates, in turn sitting on the diagrid. The latter was a massive steel support structure, the name apocryphally coming from an abbreviated note indicating ‘diameter of grid’ written on an engineering drawing. The diagrid and support structure had many penetrations to allow coolant flow to the fuel channel in the graphite stack.

The graphite core could be as massive as 4000 Te and the fuel weight totalled between 113 Te and 350 Te (uranium only). Earlier designs of reactor used a ball­bearing interface between the graphite and support structure to accommodate differential thermal expansion, whereas later designs used spigots to locate the graphite. The latter was enabled by development of a core restraint very similar to that used in the later AGR design. Both designs used quite complex linkages to tie the graphite to a steel structure, which expanded at different rates. The core restraint of the earlier reactors used a temperature compensated steel circumferential

hoop, designed to expand at the same rate as the graphite (using similar techniques to those deployed centuries earlier in pendulum clocks). Later designs used a steel restraint structure in which the graphite expands radially at the same rate as the steel. This was accomplished by rigid tie rods set between the core restraint and the outer bricks, with differential vertical expansion being accommodated by pivoting ‘Warwick’ links.

Access for fuelling and servicing of control rods for example, is via a series of standpipes at the top or bottom of the vessel. Magnox reactors have a large number of fuel channels, varying in UK designs from 1700 for early reactors to 6150 at Wylfa. In order to reduce the number of penetrations, each standpipe served a number of fuel channels (varying from 16 to as many as 60). Most reactors were fuelled from above the reactor, but Hunterston A was unique in the UK in being fuelled from beneath; the reactor pressure vessel and bioshield were suspended above a large refuelling hall.

The integrity of the pressure vessel was a key part of the reactor safety case, resulting in complex rules being introduced to keep the vessel operating in a region where brittle fracture was not possible — i. e. in a region where, in extreme conditions, leak before break would occur. The pressure vessel temperature and pressure operating envelopes were revised over the lifetimes of the stations as understanding of radiation embrittlement developed. The pressure vessels are exposed to neutron irradiation, causing displacement of atoms in the steel grains and building up defects. Above a certain temperature the defects are annealed out of the steel and the fabric is maintained in a ductile region. Extensive programmes of monitoring and research were undertaken to understand the phenomena.

Corrosion of reactor steels was also an issue, with early limits being placed on some reactors due to failure to control the steel grade of bolts and washers adequately. Fatigue failure was also an issue, but limited in likelihood by keeping the number of major changes in temperature to a minimum (i. e. keeping the number of start-up/shut-down cycles to a minimum, and not load-following to any extent).

Expansive forces on the ductwork leading to the boilers were either taken up by designing the possibility of movement into the structure, using a system of ductwork hangers, or by use of bellows joints with rigidly fixed ductwork.

The biological shield (bioshield) comprised a separate, thick concrete cylinder surrounding the SPV. In order to control the temperature of the concrete, a shield cooling air flow was maintained by passing air between the SPV and the bioshield. This air was exposed to neutron irradiation, resulting in discharges from the cooling air stack of short-lived N-16 (7 s half-life), O-19 (26 s half-life) and Ar-41 (1.8 h half-life).

Concrete pressure vessel reactors

In the UK, only four Magnox reactors were constructed using pre-stressed concrete pressure vessels (PCPVs): two each at Oldbury and Wylfa (Fig. 12.1).

©

CAD REF. WYA/A2Z/031

72.7 A diverse range of fuel elements was produced for the Magnox reactors in the UK.

These represented the latest designs of the Magnox series, and had many features in common with the slightly later AGR reactors. In France, Chinon A3, St Laurent des Eaux and Bugey 1 were also of a concrete pressure vessel design.

The advantages of the PCPV over SPVs were many, including:

• lower cost

• greater strength (so higher coolant pressures were possible)

• combining the bioshield and pressure vessel into one structure

• avoiding working at the limits of weld technology

• flexibility of shape

• enabling much larger volumes to be incorporated, so the boilers could be built into the pressure vessel avoiding the need for external ductwork

Pre-stressed concrete pressure vessels (PCPVs) are constructed of concrete, many metres thick (5 m being a typical dimension). Each vessel contains a large number (thousands) of steel tendons in a helical formation from top to bottom threaded through mild steel tubes, which are embedded in the concrete. Each tendon consists of a number of strands of wire and each strand is of seven-core construction. Stressing galleries above and below the top and base caps of the vessel permit tensioning of the steel cables used to maintain the concrete in compression.

The construction comprises a cylindrical steel liner with lugs to locate it once the concrete is poured, with cooling pipework and penetrations (in particular for water and steam, gas circulator penetrations and refuelling standpipes) built into it.

The liner is internally insulated with foil mesh to keep hot gas away from the surface, and cooling water flowing through pipes in the concrete keep the temperature to typically ~40 °C. As repair of these pipes is to all intents impossible once the concrete has been poured, a degree of redundancy is incorporated to enable plugging of defective tubes, and strict chemical control is applied to prevent corrosion. Radiolytically generated oxygen is inhibited or removed by application of an overpressure of hydrogen or continuous degassing of pressure vessel cooling water.

As mentioned above, the boilers of Magnox reactors with a PCPV are contained within the PCPV cavity. The reactor core of these reactors was surrounded by a boiler shield wall, keeping neutron activation of the boilers (and the pressure vessel) to a minimum and permitting manned access to the boilers for inspection and repair. The refuelling arrangements and fuel channels were generally similar to the older SPV stations (bearing in mind that no two were identical).

An alternative to the SFR is the lead-cooled fast reactor (LFR) (see Fig. 13.22). The LFR features a fast-neutron spectrum and a closed fuel cycle for efficient conversion of fertile uranium. It can also be used as a burner of MAs from spent fuel and as a burner/breeder. A summary of the main characteristics of different LFR designs is provided in Table 13.14. The plan is to develop the system towards commercial operation within the next five years.

The LFR is similar to the SFR, except that the coolant is either lead or a lead — bismuth eutectic (LBE). This improves safety given that lead is a relatively inert coolant (Cinotti et al., 2009). The lead coolant is contained inside a reactor vessel

surrounded by a guard vessel. Lead is preferred to LBE to avoid the formation of alpha-emitting 210Po (formed from 209Bi by neutron capture) and to avoid relying on Bi which is a scarcer material. Pb is a coolant with very low neutron absorption and moderation, so it is possible to maintain a fast neutron flux even with a large amount of coolant in the core. This allows an efficient utilization of excess neutrons and reduction of specific U consumption. Reactor designs can readily achieve a breeding ratio of about 1, and long core life and a high fuel burn-up can be achieved (Cinotti et al., 2006). The use of Pb means the LFR has a simpler design and thus has lower capital costs than the SFR, making it more competitive for electricity generation (Cinotti et al., 2006). Pb is also important in the design of sub-critical accelerator-driven transmutation systems (ADS), because the coolant can also serve as a spallation target, and because the nuclear cross sections of Pb allow high-energy neutrons to be utilized particularly efficiently in a process known as adiabatic resonance crossing (Abram and Ion, 2008).

There are, however, some disadvantages to using Pb. Experience using Pb coolant is limited compared to Na. There is a need for more research in such areas as system design, components and innovative fuel and fuel cycle development

(Vezzoni, 2011). Pb is more than 11 times denser than Na and thus requires significantly higher pumping power. This greater density also makes it harder to achieve a seismically safe design. The greatest challenge, however, is the corrosive and erosive nature of Pb, which requires careful oxygen control and the use of highly corrosion/erosion-resistant materials (Abram and Ion, 2008).

There are two reactor designs being developed within the GIF framework (see Table 13.15) (Cinotti et al., 2009):

1 a 600MWe design (Fig. 13.23) based on the previous European lead-cooled system (ELSY), now called ELFR (European lead fast reactor) (Mansani, 2011)

Table 13.15 Summary of the main characteristics of primary and secondary systems of different LFR designs (Colombo et al., 2010)

13.23 600 MWe ELFR (Mansani, 2011).

2 a small modular design of 20 MWe (Fig. 13.24) based on the small secure transportable autonomous reactor (SSTAR).

The 600 MWe design has a simple and compact primary circuit with removable components. The reactor has a secondary water loop with steam generators feeding a turbine. This simple design should mean lower capital costs and construction time. The compactness of the design also means a smaller reactor building. The ELSY core consists of an array of open fuel assemblies of square pitch, surrounded by reflector assemblies to reduce the risk of coolant flow

blockage while a closed hexagonal arrangement of assemblies has been proposed for ELFR (Mansani, 2011). The core is self-sufficient in Pu and can burn its own generated MAs with a content at equilibrium of about 1% heavy metal (Cinotti et al., 2009).

The 20MWe design uses natural circulation in the primary lead loop, with a secondary supercritical CO2 loop for power conversion in a direct Brayton cycle. The combination of compact size, Pb coolant, nitride fuel containing TRU elements and a fast spectrum core all promote a high conversion ratio. This improves proliferation resistance, fissile self-sufficiency, autonomous load following, simplicity and reliability of operation, transportability and a high degree of passive safety. However, the SSTAR design relies on further developments, including a high-performance code-qualified TRU nitride fuel (Cinotti et al., 2009).

The changes in the properties of fuel pellets are important since they affect the stress on the fuel rod cladding. The changes in the pellet are also more pronounced with a higher fuel burnup and hence there is increased impact on the cladding, which can cause fracturing of the cladding. Fuel pellet radioactivity provides also a source term for the evaluation of potential releases in accident analysis and public exposure. One of the parameters important for pellet change is grain size. Currently, PWR and BWR fuels taken to higher burnup are manufactured with grain sizes of 8-12 pm. During fuel irradiation there is some grain growth in the central hotter region of the fuel pellet. This grain growth depends on the operating temperature of the reactor but also on the irradiation time, which means with fuel burnup. During irradiation, fission gas is produced in the grains, which migrates to the grain boundaries where fission gas bubbles form. During prolonged irradiation of higher burnup fuel, the gas bubbles may become interlinked and may release gases and volatile fission products into the fuel cladding gap, which could assist the release of fission product gases in an accident or cladding breach. There is a tendency to increase the grain size in the course of fuel manufacturing in order to increase the diffusion length for fission gases within a grain.3 In conclusion, fission gas release into the cladding gap is accelerated with higher fuel burnup. In addition to gases, the release of fission product volatiles like iodine and caesium (Is) may also be important. The Cs release path follows the Xe path closely (besides being a direct fission product Cs-137 is also generated by the в decay of Xe-137, which has a short half-life) and it migrates from hotter grains in the pellet and may migrate to the pellet interface. At burnups above 40 GWd/tU a rim starts to form at the outer radius of fuel pellets. The pellet rim is characterized by much higher porosity with the formation of many smaller grains, which can retain noble gases.3 This rim generation related to burnup is also important for potential gas and particulate releases in a cladding breach.

The main wastes arising from reprocessing are the structural elements of the fuel assemblies and raffinate. The former are compacted and concreted into steel drums; the latter is vitrified as described below.

Vitrification

Raffinate concentrates 99.5% of the activity — fission products and minor actinides — into a small volume. This material is not reusable and must therefore be transformed into a solid waste. In countries using the PUREX process at industrial scale (France, United Kingdom and Japan), vitrification is the selected method. The de-nitrated concentrates are calcined after which borosilicate glass frit is added. The radionuclides are chemically incorporated into the molten glass, which is then poured into special stainless steel containers — canisters. This immobilizes the radionuclides in a durable glass matrix that is virtually insoluble in water and relatively immune to the action of natural physico-chemical agents. The content of fission product oxides in a canister is around 12% by weight while the content of actinides oxides is around

1%. Containers are stored in cooled or ventilated shafts or, for foreign spent fuel, returned to their country of origin (Fig. 16.18).

Most processes use a large volume ceramic smelter, heated using the Joule effect by means of electrodes. At La Hague, however, vitrification is performed with a small volume, induction-heated metal smelter. Several production lines in parallel are used to match the flow rates of the plant and the method uses cells that are suitably shielded and remotely operated (Fig. 16.19).

An evolution and improvement of this process is the cold crucible technology (Fig. 16.20), which eliminates some of the limitations of the hot crucible technology:

• The behaviour of the material of existing hot crucibles limits the temperature to 1100-1150 °C and, consequently, the solubility and level of incorporation of radionuclides into the glass matrix.

• The requirements for regular maintenance of the vitrification lines.

A cold crucible enables higher temperatures to be reached in the crucible and simplifies the process by removing the calciner.

Unlike the melting pot, which heats glass by thermal conductivity from the walls of the pot to the core of the glass bath, the principle of the cold crucible is to induce electric currents within the glass to raise its temperature without directly heating the crucible.

The cold crucible operates according to the direct induction principle and combines the following:

16.20 Cold crucible melter (Source: Etienne Rousset, Patrice Brun, Emilien Sauvage, Armand Bonnetier, Steven Daix, French Atomic Energy Commission (CEA) Nuclear energy division, Confinement research and engineering department. Flux Users Conference, Barcelona, Spain — October 9-10 2008).

16.21 Prototype of cold crucible melter (numerical modelling of a cold crucible for direct induction melting of a glass (Source: Etienne Rousset, Patrice Brun, Emilien Sauvage, Armand Bonnetier, Steven Daix, French Atomic Energy Commission (CEA) Nuclear energy division, Confinement research and engineering department. Flux Users Conference, Barcelona, Spain — October 9-10 2008).

• the wall of the cold crucible is cooled by a pressurized water circulation system

• the sectorized cylindrical structure is made of stainless steel

• a protective layer of condensed glass then forms, called a ‘self-crucible’. This protects the metal crucible from the effects of high temperatures and corrosion caused by the bath of molten glass

• a high thermal power is released to the melt

• a temperature higher than 1200 °C can be reached

• thermal homogenization is performed with a stirrer

• this enables high throughput capacities

• a cold cap on the crucible helps to confine the hotter, more volatile glass

It is thus possible to have a low wall temperature while inside the crucible the temperature is very high. Insulated by a thin layer of glass, the wall is protected from the melted glass (Fig. 16.21).

The intense radioactivity of the resulting high-level waste creates heat, which obliges it to be cooled during storage for several years. At the end of this period, the waste could be sent for deep disposal although it will be necessary to distribute the canisters over an area that is large enough to allow the heat to be dissipated without producing an unacceptable increase in temperature (Fig. 16.22).