Criticality is an uncontrolled nuclear chain reaction (divergence) within fissile materials, which are the odd-numbered isotopes of uranium and plutonium (233U, 235U, 239Pu and 241Pu). The risk of criticality is due to the presence of a significant amount of fissile material in some part of the facility and this is therefore broken down into several units so that adequate control modes may be defined. These modes can be characterized by limits given by up to three parameters while also taking into account the presence of neutron poisons, if any:

• fissile material mass

• geometrical dimension of the piece of equipment

• moderation ratio

To measure the damage resulting from a nuclear incident or accident, there is an international nuclear and radiological event scale, the INES scale (Fig. 16.24).

Major challenges for P&T are the specification, fabrication and performance validation of transmutation fuels and targets. In addition to MA content and composition impact on P&T, it is necessary to have a comprehensive understanding of the fuel performance and behavior under a wide range of irradiation conditions and burn-ups. This understanding should encompass such effects as helium release, high-temperature gradients and cladding properties. Some of the current fuel type options for the different strategies discussed in this section are given in Table 17.6 (extracted from Ref. 58).

Notes:

*SFR: sodium-cooled fast reactor; LFR: liquid metal-cooled fast reactor; GFR: gas-cooled fast reactor; ADS: accelerator driven (sub-critical) system tASS: austenitic stainless steel; FMS: ferritic martensitic steel; ODS: oxide dispersion steel.

Issues related to different types of fuels for homogeneous recycling (oxide, metal, nitride, carbide) are being actively investigated and international comparisons are underway.

As for IMF, the main advantage of their utilization is the potential to destroy Pu and MAs more effectively than with the fuels containing U because this avoids further Pu production during irradiation. These fuels have been envisaged as ADS fuels and for the targets of the heterogeneous recycling mode (see Section 17.3).

In general, the criticality, radiation protection and cooling requirements, due to the presence of large amounts of isotopes with high decay heat and neutron production by spontaneous fission (both even-mass Pu isotopes and MAs) in the transmutation fuel, will determine the production capacity of a fabrication facility. In this respect the neutron production issue discussed in Sections 17.2 and 17.3 is a crucial one. In this respect, if one excludes the case of very high build-up of Cf-252 (see Section 17.2), curium is the most problematic element for fabrication plants. It is a powerful source of spontaneous neutrons and has a decay heat of 2.8 W/g from its main isotope, Cm-244, so that its handling is problematic. In terms of impact curium is followed by Pu-238, whose accumulation during recycling also raises similar issues, although to a lesser extent than curium. The next, in terms of difficulties, is Am-241. Appropriate (thick) neutron and у ray protection are needed, implying remote handling. The transport of fuel assemblies and in-reactor handling operations are also affected by these thermal and radiological constraints which, of course, would significantly raise costs.

The Transport Regulations are supported by guidance documents that include:

• TS-G-1.1, Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (IAEA, 2008a)

• TS-G-1.2, Planning and Preparing for Emergency Response to Transport Accidents Involving Radioactive Material (IAEA, 2002)

• TS-G-1.3, Radiation Protection Programmes for the Transport of Radioactive Material (IAEA, 2007b)

• TS-G-1.4, The Management System for the Safe Transport of Radioactive Material (IAEA, 2008b)

• TS-G-1.5, Compliance Assurance for the Safe Transport of Radioactive Material (IAEA, 2009b)

In addition to the documents already mentioned, two others are key to understanding the basis of, and application of, the requirements and guidance provided by the IAEA documents already mentioned. One is known as the fundamental safety principles (IAEA, 2006), and the other is the basic safety standards for radiation protection upon which many of the transport requirements are based (IAEA, 1996).

PCGE uses constant fuel costs that it says ‘are comparable with the assumptions used in the World Energy Outlook (IEA, 2009)’. Reference to the World Energy Outlook (WEO) 2009 10 indicates that the PCGE gas and coal prices broadly correspond to the fossil-fuel price assumptions in the reference scenario at around 2015. Predicting long-term fuel prices is, clearly, very difficult (one reason, perhaps, why WEO 2009 runs to almost 700 pages). Nevertheless, as Eq. 5.3 makes clear, when comparing technologies for future electricity generation, it is not today’s price of fuel but the long-term average that is relevant. Consequently, future movements in fuel prices are crucial, especially those of fossil fuels, which are likely to remain the key determinants of electricity prices for the foreseeable future. Sensitivity of LCOE to fuel price variability is discussed in Section 5.2.6.

For nuclear fuel, costs are relatively constant across the world and, unless plant-specific data are available, PCGE uses a standard figure of $9.33 per MWh(e). This is made up of two elements: front — and back-end costs. The former include uranium mining, conversion, enrichment and fuel fabrication and amount to $7 per MWh(e). Back-end costs include spent fuel storage, reprocessing and disposal and are estimated at $2.33 per MWh(e). This figure is taken by PCGE to apply whether the fuel cycle is open or closed and, presumably, is based on an assumption that the additional cost involved in fabricating fuel elements from mixed (i. e. uranium-plutonium) oxide fuel is offset by the savings on mining, conversion, enrichment and disposal.

When allowance is made for thermal efficiency (assumed to be 33% net for nuclear), fuel price is a little above $3 per MWh(th). The cost of uranium (U3O8) is a relatively small part of this but it is the component that, with an expansion of nuclear power, is most likely to increase. PCGE and WNA point to a long-term price of around $23 per pound of U3O8 (and rising), which broadly corresponds to a natural uranium price of around $0.5 per MWh(th).[9] This increase could be offset to some extent by the economies of scale for conversion, enrichment and fabrication. Assuming that uranium prices will double, a long-term nuclear fuel price of $3.5 per MWh(th) is assumed here (Table 5.2). In coming to this figure, we assume that mixed oxide (MOX) fuel will not serve as a significant brake on uranium prices due to the high cost of reprocessing. In passing we note that this assumption is not entirely compatible with the view taken by PCGE that the costs of open and closed fuel cycles are comparable.

The PCGE study provides detailed data on the cost of coal of various types in a range of countries that produce a mean price of $8.2 per MWh(th). There is wide variability from one country to another with a minimum of less than $3 per MWh(th) occurring in Australia and South Africa. WEO 2009 suggests that, partly because of movements away from coal-fired electricity generation, the price of steam coal in OECD countries will increase only slowly to produce a 2020 price of around $104 per tonne which corresponds to about $13 per MWh(th); this is the value assumed here (Table 5.2) although the large regional differences need to be recognised.

For gas-fired plants, the PCGE study provides 27 fuel cost values, which, after allowing for thermal efficiency differences, produce a mean of $30.7 per MWh(th). Maximum and minimum values are about $16 and $40 per MWh(th) respectively. This spread reflects regional differences — the highest prices occur in Japan and Korea and the lowest in China and Russia. Since PCGE was published, however, gas prices have fallen significantly due to reduced demand in the economic downturn and increased supply due to new fields being opened and the development of so-called unconventional gas resources. A 2011 edition of the IEA World Energy Outlook11 — subtitled ‘Are we entering a golden age for gas?’ — points out that conventional gas reserves are sufficient to last 150 years at the current rate of use and that unconventional reserves (e. g. shale gas) are at least equal to that. Furthermore, gas reserves have a wide geographical distribution. Consequently, IEA sees gas consumption peaking in 2035 at a level that is about 54% greater than today. USA prices are expected to almost double while European prices will increase by almost 50% over the 2009 average. These changes produce 2035 prices that are very similar to those used in PCGE and, for that reason, a figure of $30 per MWh(th) is adopted here (Table 5.2).

Secondary supplies today account for the equivalent of about 17 000 tU per year. This will drop sharply in 2014 when the supply of blended-down Russian high — enriched uranium to USA ceases, but in most scenarios will recover to at least 16 000 tU/yr by 2020.

The most obvious secondary source of uranium is civil stockpiles held by utilities and governments. The amount held here is difficult to quantify, due to commercial confidentiality. As at January 2011 some 120 000 tU total inventory was estimated for utilities, 10 000 tU for producers and perhaps 20 000 tU for fuel cycle participants, making a total of 150 000 tU (WNA Market Report 2011). These reserves are expected not to diminish, but to be replaced as they are drawn down and perhaps show a steady net increase to provide energy security for utilities and governments, especially in China.

Recycled uranium and plutonium another source, and currently save are 1500-2000 tU per year of primary supply, depending on whether just the plutonium or also the uranium is considered. In fact, plutonium is quickly recycled as MOX fuel, whereas the reprocessed uranium (RepU) is mostly stockpiled. Some 100 000 tonnes of used fuel has been reprocessed so far in the civil sector.

Used fuel is about 96% uranium (burned down to around 1% U-235), about 1% plutonium (with about one third non-fissile isotopes) and 3% fission products and actinides, which are wastes. Reprocessing the fuel separates these so that the plutonium is promptly recycled into mixed oxide (MOX) fuel and the uranium is either recycled via a conversion plant or stockpiled as strategic reserve. It is more expensive to convert and enrich than fresh uranium from the mine, due to two new uranium isotopes.

Recycling the plutonium is simply continuing and expediting something which occurs already: some of the uranium-238 comprising most of the fuel is progressively turned into plutonium, and the fissile part of this burns along with the uranium-235. Consequently about one third of the energy from any power reactor comes from burning plutonium. If MOX fuel is used, of course much more does.

Re-enrichment of depleted uranium (DU, enrichment tails) is another secondary source. There is about 1.5 million tonnes of depleted uranium available, from both military and civil enrichment activity since the 1940s, most at tails assay of 0.25­0.35% U-235. Non-nuclear uses of DU are very minor relative to annual arisings of over 35 000 tU per year. This leaves most DU available for mixing with recycled plutonium on MOX fuel or as a future fuel resource for fast neutron reactors. However, some DU that has relatively high assay can be fed through under­utilised enrichment plants to produce a natural uranium equivalent, or even enriched uranium ready for fuel fabrication. Russian enrichment plants have treated 10 000-15 000 tonnes per year of DU assaying over 0.3% U-235, stripping it down to 0.1% and producing a few thousand tonnes per year of natural uranium equivalent. This Russian program treating Western tails has now finished, but a new US one is expected to start when surplus capacity is available, treating about 140 000 tonnes of old DU assaying 0.4% U-235.

The world’s nuclear weapons stockpiles provide an important source of nuclear fuel. Since 1987 the United States and countries of the former USSR have signed a series of disarmament treaties to reduce the nuclear arsenals of the signatory countries by approximately 80%.

The weapons contained a great deal of uranium enriched to over 90% U-235 (i. e. up to 25 times the proportion in reactor fuel). Some weapons have plutonium-239, which can be used in mixed-oxide (MOX) fuel for civil reactors. From 2000 the dilution of 30 tonnes of military high-enriched uranium has been displacing about 10 600 tonnes of uranium oxide per year from mines, which represents about 15% of the world’s reactor requirements. Over 2000 to 2013 this Russian ex-weapons material is supplying almost half US power generation requirements, hence accounting for about 10% of US electricity, but that supply terminates in 2013. Under this program, by August 2011, 425 tonnes of high-enriched uranium (HEU) had been downblended into some 12 250 tonnes of low-enriched uranium (LEU) for reactor fuel, representing 75 million SWU of enrichment and about 17 000 warheads, at a cost of $7.2 billion (paid by electricity consumers).

On the US side, 174 tonnes of military high-enriched uranium has been declared to be surplus and available for civil power generation. A start has been made on downblending this and the first fuel fabricated from it has been shipped to Tennessee Valley Authority (TVA) power plants. The US Department of Energy’s (DOE’s) National Nuclear Security Administration (NNSA) has awarded contracts to downblend 17.4 tonnes of HEU from dismantled warheads to yield about 290 tonnes of LEU, 230 tonnes of which will be part of a new American Assured Fuel Supply (AFS) program. In June 2009, the NNSA awarded a further contract for downblending 12.1 tonnes of HEU, which will yield some 220 tonnes of LEU by 2012.

There is insufficient up-to-date information and experience available to develop a meaningful cost projection for thorium-fuelled reactors at the present time. Nevertheless, it should be possible to make a few comparisons with the equivalent uranium-fuelled systems. The main cost item for an LWR system comes from the up-front expense of preparing the site and building the reactor and its associated cooling circuits. In general, the capital cost and operation and maintenance cost of reactors will hardly be affected by the type of fuel being used.

Fuel constitutes a small but still significant component of the overall cost of nuclear power — the most recent OECD-NEA study 1 3 indicates that the fuel cycle represents around 20% of the cost of a typical nuclear generation operation. This is not insignificant, of course, and, because of past efforts to minimize capital and operational costs, it is of increasing importance the remaining cost component that is influenced by essentially external factors such as uranium price,
front-end and back-end fuel cycle service costs, etc. The uncertainties associated with these external factors (e. g. higher and more volatile uranium prices) make it necessary to consider possible alternatives, of which the thorium cycle is one. Within this 20%, the costs can be further broken down:

Uranium

Uranium conversion Enrichment Fuel fabrication Back-end activities Total

where the ‘back-end activities’ include interim storage, reprocessing and waste disposal.

Considering the items in this list and evaluating the possible impact of a change from uranium to thorium, we can see that, at the present time, as a by-product of rare-earth production and, with no great demand, thorium is essentially ‘free’. This would change, of course, if the thorium cycle were to be widely adopted. Nevertheless, in comparison to uranium, we would expect its greater availability and easier mining conditions to allow it to be produced at no greater cost.

Moving down the list, it is clear that, once a closed thorium cycle is established in which U-233 was the fissile material, conversion and enrichment would not be needed at all and this may represent a significant cost saving. Before that situation could be reached, however, a supply of fissile material — U-235 and/or Pu — would be needed to supply the neutrons needed to transmute thorium to U-233. In the case of U-235, conversion and enrichment will, of course, be needed. Indeed, if medium enriched uranium (MEU) is used, five times more SWU (Separative Work Unit) would be needed than for conventional uranium fuel (enrichment 4 to 5%) although the smaller amount of MEU in U/Th-fuels will compensate for this. If Pu is used as seed material, then conversion and enrichment would not be needed, but Pu-based fuel fabrication is approximately three to five times more expensive than for UOX-based fuel. One may conclude from this that, while a closed thorium cycle will not incur conversion and enrichment costs, the establishment of the cycle will entail costs for these processes (or their equivalent for Pu seed fuel) that are likely to be greater than for a conventional once-through cycle.

As explained, because of the need for remote handling, fuel fabrication with U-233 will be more expensive than U-235. The refabrication cost will therefore be higher than MOX fuel fabrication where glove boxes are sufficient. Similarly, the greater technical difficulties of the THOREX, as compared with the PUREX, process will also increase costs. It is, however, difficult to assess the extra-cost since no industrial feedback is available regarding back-end operations. Regarding waste management, the THOREX process might generate 50-70% more vitrified waste than PUREX (cf. Section 8.3.3), so that the interim storage and long-term repository costs might also engender extra costs.

Given the wide variety of thorium-fuel options being investigated today, each of these variants has its own specific economic potentials and limitations and a general statement on the economic performance of thorium-fuel options in LWRs is difficult because the specifics of the individual options would need to be considered, which is beyond the scope of this chapter. Nevertheless, today’s renewed interest in thorium-fuel options is specifically directed at the formulation of a well-founded technical-economic assessment of its industrial viability. This work relies on updated knowledge and experience resulting from R&D undertaken worldwide during the last decades.

8.2 Conclusion

Thorium is not a direct competitor to uranium since thorium does not contain fissile isotopes. It thus must be used in combination with fissile isotopes from another source (enriched uranium, plutonium or U-233). Nevertheless, thorium has always been considered as an attractive fuel cycle option for future development of nuclear energy for the following main reasons, which have been discussed and assessed in this chapter:

• the enhancement of fuel resources by producing a new fissile isotope, U-233, which is moreover the best fissile isotope for thermal neutrons

• the existence in some countries of domestic thorium and, conversely, shortages of natural uranium, combined with the knowledge that thorium natural resources in the world are probably greater than those of natural uranium

• the good in-core neutronic and physical behaviour of thorium fuel, allowing it to reach high burn-ups, high conversion factors compared to U-233 and even breeding (i. e. a conversion factor superior to 1) in thermal reactors

Today, these benefits are more relevant than ever in the context of the nuclear renaissance, and possible uranium scarcity in the decades to come. In addition, new priorities have also stimulated renewed interest in thorium-based fuels. Among them, two main reasons are to be cited: (a) the fact that the thorium cycle strongly reduces the global inventory of long-lived minor actinides (and thus the long-term radiotoxic inventory of the finally disposed waste), (b) the fact that the use of thorium allows very efficient plutonium burning.

Another argument, which is sometimes quoted in favour of thorium, is its ability to be more proliferation resistant. This argument is not very compelling because certain physical properties of U-233 make it attractive to potential weapon use. Nevertheless, the discussion presented here shows that several routes do exist to impede such utilization and that, overall, thorium fuel may be no less proliferation-resistant than uranium fuel.

Despite the benefits of thorium, its use presents technical challenges that were described in this chapter. To support thorium industrial implementation at a large scale, infrastructures need to be developed, (i. e. mining, milling, fuel fabrication, transport and reprocessing of thorium-based fuel). Reprocessing will be required if it is intended to recover and reuse the U-233 that is generated from the fertile thorium. Fuel fabrication using the recovered U-233 with its inseparable sister isotope U-232, and the build-up of U-232’s gamma-emitting daughters, will probably require a shielded facility. And the fabricated fuel will also need to be shielded from that point on.

Beyond these considerations, this review has shown that significant experience has been gained on thorium-based fuel in both test reactors and power reactors, but not on an industrial scale. The feasibility of the front-end fuel cycle technologies (mining, fuel fabrication) has been successfully demonstrated with generally rather old technologies. For the back-end of the cycle (reprocessing and recycling), however, experience is practically non-existent. Therefore, the use of thorium on an industrial scale would still entail quite significant R&D efforts and costs, to master and optimize all the steps of the fuel cycle (including a better knowledge of thorium resources and dedicated extraction processes). Nonetheless, modern technological breakthroughs such as remote fuel fabrication techniques already applied to MOX fuels, should lower the perceived technological hurdles of the past to allow the complete implementation of the thorium cycle, including U-233 recycling, which is required if this cycle is to be used to best advantage.

To sum up, it is clear that thorium-based fuel shows useful characteristics but they do not appear sufficient to justify an industrial development in the short­term, all the more so as these potential advantages are offset by some real drawbacks. On the other hand, in the term of a few tens of years, thorium could help to lower the radiotoxicity of radioactive waste to be disposed and, if U-233 is recycled, could reduce demand for uranium. In this latter respect, the possibility of achieving near breeding or even breeding conditions in thermal reactors represents a very attractive feature of the thorium cycle.

Finally, the future is unwritten. The appearance of new, currently unanticipated, constraints will, no doubt, modify the current context and lead to unexpected developments. This indicates the need to maintain and develop the thorium cycle as a credible option.

During normal operation the control systems maintain a balance between power generation (and hence reactivity control) and heat removal (leading to electricity generation). Key parameters required for control and protection are monitored by the control and protection systems. In general the functions of control and protection are designed to be independent and are often delivered by separate systems. The protection system is designed to monitor and detect deviations from normal operation, which are beyond the capability of the control system to correct. Small deviations can be corrected by the control system but larger ones will lead to reactor trip (or in US terminology, reactor scram) and the rapid insertion of control rods as well as the initiation of engineered safety systems to ensure core cooling and the confinement of radioactivity.

In PWRs the control and shutdown rods are released and fall under gravity. In BWRs the control rods are rapidly inserted hydraulically. Although these systems are very reliable, diverse means of shutting down the reactor are usually provided based on the injection of dissolved absorbers. In PWRs this is generally achieved by injection into the circulation loops while in BWRs the standby liquid control system injects into the inlet plenum of the reactor vessel.

When a large plant is tripped the loss of generation will disturb the grid and, although the grid connections are designed to cope with this, there is the possibility that local protection will trip the line resulting in a loss of offsite power to the plant. All nuclear power plants are therefore designed so their safety systems can deal with faults on the assumption that offsite power is lost simultaneously. To enable this emergency generators are provided, which will start automatically and power electrically driven safety systems. To provide redundancy it is usual to have between two and four divisions of emergency electrical power, each with their own diesel generator.

Fundamentally, safe operation of any nuclear plant requires that the design has:

• trip systems, which respond to conditions outside the safe operating envelope and which require shut down of the reactor

• shut-down systems

• cooling to remove residual heat from the reactors after shutdown

The AGR safety assessment considers a whole range of faults on the plant. Included in these are:

• loss of reactor coolant through a breach in the pressure circuit

• water ingress as a consequence of a boiler tube leak

• faults involving damage to the fuel stringer

• internal hazards such as fire

• external hazards such as seismic events

Each reactor fault is taken to its end point, which is the termination of the fault and to which a radiological consequence, which is the radiation exposure received by the most exposed member of the public, is allocated. This consequence is determined generally by calculation based on an assessment of:

• the damage to the fuel. If the fuel pin clad is breached in any way, there will be a release of radioactivity into the coolant

• the transient undergone by the fuel. Increased fuel temperatures will lead to additional releases from the fuel. Fuel oxidation, due to exposure to reactor coolant for example, will enhance releases from the fuel and could produce fine U3O8 particulate material

• retention within any containment. This is a significant mechanism for reducing releases to the environment

• releases to the environment in any leakage or other loss of coolant

• potential radiation doses to the public based on external exposure from the radioactive plume and any radioactivity deposited on the ground, inhalation of radioactivity and ingestion of contaminated foods

The radiological consequences of fuel route faults are addressed in the same way but the event trees are not as complicated. Waste treatment plant faults are generally of lower significance than reactor and fuel route faults because of the smaller source terms.

After all end point exposures are assessed, including faults on the fuel route and the waste treatment plant, the risk from the plant is determined. The risk is determined from each end point exposure and the probability, as determined from plant reliability assessments, of that sequence. The acceptability of the plant is determined by assessment of this risk against criteria developed within the operator’s own organisation. The NII have their own assessment criteria (the Safety Assessment Principles or SAPs).

Operator risks are also determined from the radiation doses received during both normal operation and as a consequence of faults.

As irradiation proceeds the fuel swelling and clad creepdown (the latter only in LWRs and AGRs) lead to closure of the gap between fuel pellets or bars and the cladding. In AGRs, however, the gap is always small because the cladding is pressurised onto the fuel pellets during manufacture. Fuel creep due to the weight of the fuel stack can enhance gap closure in (high temperature) fast reactor oxide fuel (Bailly et al., 1999).

Once the gap is closed, a significant tensile stress is exerted on the cladding by the fuel pellets/bar, and a corresponding significant compressive stress is exerted on the fuel pellets or bar by the cladding. The stresses in the fuel and/or cladding are relieved by elasticity and creep, the latter causing permanent deformation of the fuel pellets/bar and cladding. In the case of ceramic fuel pellets, the wheatsheaf shape of the pellets means that clad stress is concentrated at the pellet ends, and the cladding therefore deforms to a greater extent over pellet-pellet interfaces. The result is ‘ridging’ or ‘bambooing’ of the cladding. If the stresses on the pellet are high, axial extrusion of fuel in the hot central region and/or hot pressing (removal of porosity under stress and temperature) are also possible. Axial extrusion in turn leads to partial, or even total, filling of any dishes in the pellet end faces. The mechanical effects of pellet and cladding contact are collectively known as ‘pellet-cladding mechanical interaction’, or PCMI. This is distinct from ‘pellet-cladding interaction’, or PCI, which also includes the chemical effects (see 14.2.12).

During steady-state conditions the stresses induced by fuel-cladding contact are small, since the stress accumulation is slow and there is ample time for stress relaxation due to fuel and/or cladding creep. The result is that there is little mechanical effect, other than a slow increase in clad circumferential strain due to the inexorable swelling of the fuel. This is not, however, true during fast transient increases in pin power, when fuel thermal expansion (including the resulting opening of radial fuel cracks in oxide fuel) and gaseous swelling can impose significant stresses on the cladding. If the stresses on the cladding are high enough, the cladding can fail by either stress-corrosion cracking (see 14.2.12), or ductility exhaustion (in the case of the latter, the high stresses are relieved by clad creep, generating large creep strains in the process).

Specifications include physical properties (homogeneity, grain size, specific surface area), plutonium content (>86%) and chemical impurities (<5000 ppm as an element) and specific у activity from fission products (<37 x 103 Bq/gPu).

Reprocessed plutonium includes isotopes from mass 238 to 242, all with long half-lives (>6 x 103 years) except for plutonium-238 (88 years) and plutonium-241 (14.4 years).

The quality of the plutonium that is available for recycling is a function of the time spent in reactor because of fission and the formation of the higher isotopes. Typically, about two thirds of the Pu-239 that is created during irradiation of U-238 in a light water reactor is lost through fission so that about 1/3 is available for recycling. High burn-up fuel produces greater concentrations of Pu-241. Whilst this may fission it can also decay to americium-241, which is neutron­absorbing and an a and у emitter. To avoid a build-up of Am-241, which can place onerous restrictions on the handling of plutonium in MOX manufacturing plants, reprocessed plutonium is usually made into MOX fuel as soon as possible. If instead, plutonium is stored for a long time, it may be necessary to remove the americium chemically to avoid such problems. To avoid transport of liquids, plutonium is delivered for storage or to the MOX plant as solid PuO2.