Enrichment activities are very carefully monitored by the international community to ensure that enriched uranium that has been declared to be for civil use is not diverted for military use, or that facilities are not secretly producing uranium at the enrichment levels required for military use. This is referred to as the nuclear safeguards programme. Commercial facilities implement rigorous accounting procedures to show that there are no discrepancies between the quantity of uranium that enters the facility and the quantity of enriched and depleted material that leaves. Operations are subject to regular monitoring by International Atomic Energy Agency (IAEA) inspectors, or organisations such as EURATOM that act on their behalf and teams of safeguards inspectors are a common sight at commercial enrichment facilities.

The international community is also well aware of the different technologies that could be used to generate HEU for military use. Each technology requires key knowledge, components and technology so that alerts are raised should any organisation or country seek to acquire such things without good reason.

In order to avoid fuel failure, mechanical stresses must be limited. Such stresses are generated by the expanding pellets pressing against the cladding radially and pulling it axially, also called pellet-clad mechanical interaction (PCMI). The total strain induced in the cladding tube is usually limited to 1% elastic + plastic hoop strain or 2.5% equivalent plastic strain.

Fuel pellet dimensional changes have three main components: thermal expansion as an instantaneous response to temperature increase, fuel densification (in-core shrinkage) during the first MWd/kg of burn-up, and fuel swelling due to fission products. The development of swelling and densification is illustrated with Halden Reactor in-core measurements in Fig. 9.11.

Fuel swelling is a slow process where the volume change is typically in the order of 0.5-0.8% AV/V per 10 MWd/kgU burn-up, and the fuel and cladding accommodate to each other during steady-state operation. Failure because of mechanical overstraining is then mostly related to rapid power changes and the resulting thermal expansion of the fuel pellets. In some cases with high power and temperature, expansion due to swelling caused by gaseous fission products will also play a role. Fuel vendors prescribe the rate of permissible power change for their products and subject them to ‘ramp testing’, i. e., a rapid power increase, in

test reactors. An example of such ramp testing is shown in Fig. 9.12. The cladding failed in this case as indicated (in the figure) by the decrease in clad strain and the increase in detectable gamma activity. The failure limit is a function of many parameters: the burn-up of the fuel, the conditioning level before the start of the power increase, the power increment, the ramp rate and the final power level all play a role. Vendors therefore qualify their respective products with extensive ramping campaigns.

The release of fission products from the pellets to the rod’s free volume has two performance limiting consequences. One is the effect of the fission product iodine, which plays a role in so-called stress corrosion cracking (SCC) and is available in sufficient quantity in a fuel rod for the SCC mechanism to work. Liner or barrier fuel, as described in Section 9.4.2, is the remedy adopted for mitigating the problem in BWRs and CANDU, and power change restrictions are applied in PWRs and BWRs.

Another limitation stems from the release of krypton and xenon. These noble gases are generated with about 28 cm3 (stp) per MWd from fission of U-235 in a thermal neutron spectrum. If the fuel temperature is high enough, they diffuse out of the fuel matrix to the rod’s free volume and cause a pressure increase. In Pu-MOX fuel, the generation and release of helium contributes in addition. When the rod pressure exceeds the outside system pressure, the cladding will creep outwards, and a gap may open between the cladding and the fuel pellets if creep — out is faster than fuel swelling. As a consequence, the fuel temperature will increase, which in turn will lead to more fission gas release, even higher pressure and more creep-out. Consequently, safety regulations set a limit on rod pressure

(NEA, 2003). Fission gas release and rod pressure build-up is a life-limiting factor, but it was found that considerable overpressure can be sustained before the fuel temperature responds with an increase, see Fig. 9.13 (Wiesenack et al., 2006). A remedy is to use large grain fuel (50-100 pm), which has a longer average diffusion path to the grain boundaries and thus retains the fission gases better than standard grain fuel (8-10 pm).

Another possible novel used-fuel recycling process for LWR fuel would remove uranium by making use of the volatility of uranium hexafluoride. Some fission products would also be removed leaving plutonium, minor actinides and some residual fission products. This so-called ‘plutonium ash’ could be used as CANDU fuel (Dyck et al., 2005).

Detailed reactor studies and MOX fuel fabrication and irradiation tests have confirmed the feasibility of MOX fuel use, including ex-weapons plutonium, in the CANDU reactor (Chan et al, 1997; Dimayuga, 2003; Dimayuga et al., 2005) as well as in the ACR-1000 (Ovanes et al, 2009).

Some countries are looking at alleviating public concerns about the back-end of the fuel cycle by partitioning and transmuting the used fuel. In this concept, some of the long-lived actinides from used LWR fuel would be fissioned in fast reactors designed for that purpose. By utilizing the actinides in the used LWR fuel first as fuel in a CANDU reactor, the total number of fast reactors needed for ultimate destruction of the LWR actinide waste can be significantly reduced. In this role, the CANDU reactor would be an effective intermediate burner between LWRs and fast reactors, by reducing both the decay heat and the radiotoxicity of the used LWR fuel (Hyland and Dyck, 2007; Hyland et al. , 2009a; Hyland and Gihm, 2010).

Thorium fuel

A large variety of thorium fuel cycles could be utilized in the CANDU reactor, the ACR-1000 or the PT-SCWR. The fissile component could be provided by plutonium from reprocessed LWR fuel, LEU or from recycled U-233. The use of thorium in the CANDU reactor can be tailored to meet national considerations, such as availability of fissile and fertile material, the availability of recycling and fuel fabrication facilities and strategic objectives. The use of thorium would provide protection against the increasing cost of uranium as resources dwindle, and would help assure long-term resource supply and diversity.

The CANDU reactor and its evolutionary variants offer the potential of a staged approach to thorium fuel cycles. The simplest, near-term fuel cycle option is the ‘once-through’ thorium fuel cycle, in which economic and resource benefit is derived from the use of thorium without the need for recycling (although the fissile U-233 created is available for future recycling when and if needed). In the near term, the fissile material required to initiate the thorium cycle could be provided by LEU fuel. Alternatively, if plutonium were available from conventional reprocessing or advanced recycling technologies, it could provide the external source of neutrons needed to initiate the cycle. In this case, plutonium would be consumed and fissile U-233 would be created.

Regardless of the source of fissile material, the CANDU reactor provides several options for configuring the arrangement of fissile driver material and fertile Th-232 (Boczar et al, 2002a, 2002b):

• a homogeneous fuel in which the fissile and fertile components are co-mixed in the same fuel pellet and bundle (Hyland et al, 2009b)

• heterogeneous ‘mixed bundle’ designs in which the fissile and fertile components are in separate elements in the same bundle

• heterogeneous ‘mixed channel’ designs, in which the bundles containing thorium and the bundles containing the driver fuel are irradiated in separate channels. This option allows a different irradiation time for the thorium and driver fuel bundles/channels

The full benefit of the thorium fuel cycle would require recycling of the U-233 produced during irradiation, which could be done in existing CANDU reactor designs. Looking to future developments, the Self-Sufficient Equilibrium Thorium (SSET) cycle offers the potential of a near-breeder thermal reactor, which is self­sustaining in fissile material (recycled U-233). To achieve the SSET cycle would require even greater improvements in neutron economy. This could be achieved by further optimizing the lattice design, removing the adjuster rods, increasing heavy-water purity, reducing the flux level (which would reduce parasitic absorption in Pa-233), removing the isotope Zr-91 from the zirconium structural materials in the fuel and core, which is a stronger absorber of neutrons, and ultimately by using a gas coolant (which would add around 15 mk reactivity and eliminate the positive coolant void reactivity). To illustrate the potential of this reactor and fuel cycle, in Canada, if all the plutonium from used natural uranium fuel from the existing CANDU reactors operating over their lifetime were to be used to initiate the SSET cycle, two to three times the existing nuclear capacity in Canada (currently about 15 GWe) could be sustained indefinitely (Boczar et al, 2010).

The CANDU reactor offers a phased approach to the use of thorium, starting with the simple technology existing today and progressing through stages to a closed thorium cycle. For example, China is a country having both LWR and CANDU reactors and has indicated interest in pursuing the thorium fuel cycle in their CANDU reactors (WNN, 2009). A phased approach has been proposed towards achievement of energy sustainability through the use of the thorium fuel cycle in CANDU reactors in China (Boczar et al, 2010). The main elements of that strategy are more widely applicable:

• In the short term, obtain benefit from and experience in the use of thorium in existing CANDU 6 reactors using a mixed-bundle approach (a CANFLEX bundle containing thorium in the larger central eight elements with enriched uranium in the outer 35 elements). Pure thorium bundles could also be irradiated for long periods of time in the peripheral channels.

• In the medium term, maximize the benefit from reprocessing used LWR fuel by using the recycled plutonium in homogeneous plutonium/thorium fuel in CANDU 6 or EC6 reactors (Mao et al. , 2009) or in ACR-1000 reactors (Ovanes et al, 2009). The RU could be used in existing CANDU 6 plants or in new EC6 reactors, either as-is or blended with DU to form NUE fuel. If actinide destruction is desired, this could also be pursued in CANDU reactors.

• In the long term, the thorium cycle would be closed by recycling the U-233 in EC6 reactors, or in CANDU reactors designed to have even higher neutron efficiency and a conversion ratio approaching unity (the SSET fuel cycle). A thorium-fuelled CANDU reactor would also be synergistic with fast reactors.

Fluoride-cooled high-temperature reactors (FHRs) combine liquid fluoride salt coolants (like MSRs), pool type cores and vessel configurations as with many SFR designs, and coated particle fuels as with HTR. Fluoride salts offer better heat transport characteristics compared to helium, enabling power densities 4 to 8 times higher and power levels over 4000 MWft . AHTRs and PB-AHTRs both use coated particle fuel embedded within a graphitic matrix cooled by liquid fluoride salt (Forsberg et al., 2008). The higher flux level in AHTR and PB-AHTR systems means more frequent reflector graphite replacement, which can be avoided by the use of inner and outer pebble blankets to reduce the radiation damage to the fixed reflector graphite (Renault et al., 2009). Two examples of FHR systems are:

1 The 1200 MWe advanced high-temperature reactor (AHTR), which employs prismatic fuel elements

2 The 410MWe pebble-bed advanced high-temperature reactor (PB-AHTR)

A commercial-scale PB-AHTR has been reported operating at ~900 MWth (Bardet et al., 2008). The characteristics of the fuel cycle are more or less the same as the VHTR, while intermediate heat transport, power conversion and balance of plant are similar to an MSR.

As LWR and PHWR (i. e. CANDU and its derivatives) reactors represent, by far, the majority in commercial use, spent fuel characteristics will be mainly discussed with respect to a UO2 matrix fuel with zircalloy cladding. Gas reactor fuels tend to be different: an AGR uses oxide fuel with stainless steel cladding and the few remaining Magnox reactors have metallic uranium in a magnesium alloy can. There is also some historic and AGR fuels with stainless steel cladding. There is very little information available about the management of fast reactor spent fuel.

In this chapter some key characteristics of spent fuel will be described. As these depend on the burn up of fuel, they will be discussed, wherever possible, in relation to burnup. During the fission process in nuclear reactors, the fuel undergoes a number of changes, such as: depletion of U-235, transformation of U-238 to Pu-239, build-up of fission products and decay products (i. e. helium gas), generation of neutron activation products, etc. The most important consequence, of course, is that the fuel becomes intensely radioactive. The main sources of this radioactivity are the fission products and the actinides, both minor (Np, Am, Cm, etc.), and major (U and Pu). It is important to remember that this changes with burnup. The contribution of the cladding to the overall radioactivity is usually very small.

The detailed composition of spent oxide fuel depends mostly on the fuel burnup at discharge (here simply referred to as ‘burnup’). For an LWR fuel assembly, this is expressed as the average power (GW) generated by the assembly, multiplied by the number of days at power (GWd) and divided by the amount of heavy metal (usually uranium) that the fuel assembly contains. While burnup varies from one reactor type to another, there has been, since the earliest days of nuclear energy, a consistent tendency for it to increase. Figure 15.3 shows how burnup increased in the period up to 2005. Such increases require higher enrichment fuel although fuel endurance, including its ability to survive unplanned events, will also impose limits.

For PWRs and BWRs the overall average burnup is approaching 50 GWd/tU with some power plants reaching 59 GWd/tU.2 This burnup tendency also

applies to Russian design reactors VWER-440 and VWER-100, which reach 48-50 and 45-55 GWd/tU respectively. The Russian fuel for the channel type rector (RBMK) achieves burnups of 30-35 GWd/tU. Typically, higher burnup is achieved with higher initial enrichments of nuclear fuel, which is nowadays approaching 5% U-235. For heavy water reactors that use natural uranium, burnups are lower, Around 7.5 GWd/tU, with the latest tendencies to use slightly enriched uranium fuel (SEU) with consequently higher burnups reaching about 9.5 GWd/tU.

It has to be pointed out that many characteristics of spent mixed oxide fuels are in general similar to uranium oxide fuel but more pronounced. The isotopic characteristics of the MOX fuel are initially different.

Purified uranyl nitrate (UO2(NO3)2) solutions from the final purification cycle are concentrated by evaporation in order to obtain solutions of 300-400 g U/l, which can be reused or stored (Fig. 16.14).

‘ Less dense

phase outlet

Separating zone

И More dense

Vі phase inlet

Clean-in-place nozzles

Operating centrifugal contactor stage

16.13

Left: Cutaway view of an operating centrifugal extractor (Jack D. Law and Terry A. Todd, Idaho National Laboratory) Right: Schematics of a centrifugal extractor (F. Drain, R. Vinoche, J. Duhamet, WM’03 Conference, February 23-27, 2003, Tucson, AZ).

Plutonium nitrate is converted into PuO2 at the end of the purification cycle to avoid storage and transportation of solutions (Fig. 16.15).

The plutonium at valence IV is precipitated by oxalic acid into plutonium oxalate:

Pu4+ + 2 H2C2O4 + xH2O ^ Pu (C2O4)2. xH2O + 4 H+

The precipitate is washed, dried and calcined into oxide at 400 °C: Pu (C2O4)2. xH2O ^ PuO2 + xH20 + 2CO + 2CO2 The whole PUREX process is shown in Fig. 16.16.

I From Pu nitrate purification |

Decay heat from spent fuel and HLW is highest at the time the spent fuel is discharged from the reactor and decreases rapidly with storage time. As shown in Fig. 17.8, decay heat is dominated by fission products up to about 60 years after

discharge and by the actinide elements afterwards. Higher decay heat makes storage and disposal of the spent fuel and HLW more difficult, since temperature limits are imposed on these materials to ensure their integrity. In designing a repository for spent fuel, therefore, the distance between adjacent fuel packages is adjusted so that the heat loading does not produce unacceptable temperatures. Typically, these distances are of the order of 10 M. This is a key factor in deciding the volume of rock that must be excavated to create a repository and, therefore, the cost.

By removing the radionuclides that are responsible for the longer-term decay heat, P&T, perhaps combined with long-term interim storage, allows waste packages to be placed closer together so that the volume of rock needing to be excavated is reduced. Compared with direct disposal, we might expect a reduction of gallery length by a factor or 3-6 depending on the length of the (pre-disposal) cooling time.42 Additional gains can be foreseen by separating Cs and Sr and storing them for 100-300 years prior to disposal or, alternatively, through long­term interim storage of vitrified high-level waste without Cs and Sr separation. If longer-term decay heat is a significant issue, after 75-100 years, either from an engineering and operations viewpoint or from a more fundamental perspective in developing a repository concept, actinide P&T and interim storage prior to disposal can be effectively used to lower decay heat in the corresponding waste, allowing for an increased utilization of repository space. This general result of decay heat reduction appears to hold for different types of host rock42 indicating that P&T is an effective means of reducing disposal costs.

An international regulatory regime for the safe packaging and transport of all radioactive materials, including those materials associated with the nuclear fuel cycle, is well established. This regime is sound and long-standing; it focuses on safety in the transport of radioactive materials. In addition, a set of recommendations and guidance exist for security in the transport of these materials.

It is noted that there are some subtleties in the definitions of the two English terms — safety and security, These subtleties can lead to confusion, especially when dealing with languages other than English. In fact, in some languages (e. g. Spanish) safety and security are expressed by the same word. Thus, the following discussion focuses on the English language meaning of these two terms because that is the language in which requirements, recommendations and guidance for both safety and security during transport are developed at the international level by the International Atomic Energy Agency (IAEA) and other international governmental organizations including the International Civil Aviation Organization (ICAO) and the International Maritime Organization (IMO).

The IAEA Safety Glossary (IAEA, 2007a) defines these two terms as follows:

• Safety is the ‘achievement of proper operating conditions, prevention of accidents or mitigation of accident consequences, resulting in protection of workers, the public and the environment from undue radiation hazards ’ ; whereas

• Security is the ‘prevention and detection of, and response to, theft, sabotage, unauthorized access, illegal transfer or other malicious acts involving nuclear material, other radioactive substances or their associated facilities’.

Based upon these definitions, and for the purposes of packaging and transport of radioactive material, these two terms are treated herein as follows: (a) safety relates to the protection of people and the environment during normal operations as well as in the event of accidents, whereas (b) security relates to the protection of people and the environment from malicious, intentional actions by man.

The following provides a brief summary of the international regulatory regime that addresses safety in the packaging and transport of all radioactive materials (which include materials associated with the nuclear fuel cycle), noting how regulatory requirements devolve downward from international organizations, starting with the IAEA Regulations for the Safe Transport of Radioactive Material, TS-R-1 (IAEA, 2009a), to regulatory requirements promulgated by individual regulatory bodies at the STATE (i. e. country) level.

In the days when electricity generation and supply were performed by state — owned monopolistic utilities, lack of competition ensured that electricity generators could set the price of electricity. When electricity generation costs increased, there were two possible remedies: either the electricity consumer paid more or the government provided a subsidy. Calculations of electricity costs rarely used the net present value method but, instead, relied on fuel costs and an assumption that the capital costs were evenly spread (‘amortised’) over the generating lifetime. i4 Individual power stations were ranked according to the calculated price of their electricity and thus placed in a ‘merit order’, which largely determined how frequently they were called upon to generate. A plant that was high in the merit order would operate 24 hours per day supplying the baseload, while those lower down were employed to meet the shorter-term peaks in demand and to be there as backup in case of planned and unplanned outages. The most likely causes of a change in the ranking of a power station were the introduction of newer generators and variations in the price of fuel. Nuclear stations were often at the top of the merit order because fuel costs are relatively low but, more importantly, because (as we shall see) this method of calculating costs is especially favourable to generators that, like nuclear, have high capital costs.

The application of net present value techniques to the economics of electricity generation has been used — or at least advocated — for perhaps 40 years.4 Instrumental in disseminating the approach more widely was ‘Projected Costs of Generating Electricity’ (PCGE), which first appeared in 1983 (latest edition 2010).5 It uses a spreadsheet methodology and is now published jointly by two agencies ofthe OECD (Organisation for Economic Cooperation and Development): the IEA (International Energy Agency) and the NEA (Nuclear Energy Agency). PCGE seeks to compare electricity costs across a range of generation systems and countries that submit, on a voluntary basis, data for individual existing or projected power plants. Thus the 2010 edition gathers together data for 190 power plants from 16 OECD countries, four non-OECD countries and four industry organisations. The range of power plants includes nuclear, coal, gas and various renewable sources. Here, for example, we can find wind-powered generation in Belgium directly compared with coal-fired generation in Australia. Apart from using a widely adopted methodology, PCGE is also a useful source of data that, given the need for commercial confidentiality, is often difficult to find.

In the 1950s, the USA had a great deal of uranium mining, promoted by federal subsidies. Peak production was 16 800 tU in 1980, when there were over 250 mines in operation. This number abruptly dropped to 50 in 1984, when 5700 tU was produced, and then there was steady decline to 2003, with most US uranium requirements being imported. By 2003 there were only two small operations producing a total of well under 1000 tU/yr, though more recently the sector has recovered, buoyed by higher uranium prices, so that in 2008 no fewer than 15 mines (ten underground and five ISL) operated for at least part the year and produced 1500 tU.

Cameco operates the Smith Ranch-Highland mine in Wyoming and the Crow Butte mine in Nebraska, both of them ISL operations. Uranium One operates the Christiansen Ranch ISL mine in Wyoming. Mestena Uranium’s Alta Mesa ISL plant in South Texas is also operational.

Conventional (non-ISL) uranium mining has returned to the USA after many years. One company, Denison Mines, operates mines on the Colorado Plateau and the Daneros mine in Utah. The ore is processed at its White Mesa mill in south-eastern Utah. Several other projects are under development, though some projects and mines have been put on standby pending market improvements.