Radiotoxicity has been used as a measure of the potential benefit of TRU transmutation. If 99.5-99.9% recovery of TRUs is assumed, it can be shown that, compared to a closed or open cycle PWR, use of P&T causes the radiotoxicity of buried highly active materials (essentially the losses at reprocessing) to be reduced by about two orders of magnitude after several hundred years. By comparing the calculated radiotoxicity with the ‘reference level’ it can be seen that, with P&T, the buried material is less radiotoxic than the initial uranium ore after approximately three hundred years. In comparison, with a once-through (direct disposal) fuel cycle it takes more than 1-200000 years to achieve this. A typical example is given in Fig. 17.7.

The upper curve in Fig. 17.7 represents the natural reduction in radiotoxicity with time for the direct disposal of spent LWR-UOX fuel, at a burn-up of 51 GWd/ MTIHM (once-through cycle). The middle curve corresponds to the multi­recycling of Pu in a fast reactor so that FPs plus MAs plus Pu losses at reprocessing are sent to a geological disposal. The lowest curve corresponds to the multi­recycling of both Pu and MAs in a fast reactor so that only FPs plus Pu+MA losses at reprocessing are sent for geological disposal. In all reprocessing cases, the recovery factor is assumed to be 99.9%. This factor is already reached at industrial scale in the case of Pu (e. g. at the La Hague plant in France) and it has been reached at laboratory scale (or pre-industrial scale) in the case of MAs (e. g. at the ATALANTE facility in France43).

As indicated above, one interesting result shown in Fig. 17.7 is that, if Pu+MA are multi-recycled and only losses at reprocessing are sent to a geological disposal together with FPs, the radiotoxicity of the disposed material reaches the level of the radiotoxicity of the initial uranium ore, after only few hundred years, while in the case of direct disposal of the irradiated fuel, it takes 2-300 000 years to get the same result.

R. B. POPE, Consultant, USA

Abstract: This chapter addresses both the packaging and the transport of the key radioactive materials associated with the fuel cycle. It focuses on the safety principles associated with these transport activities, the regulations that result from the worldwide application of these safety principles, and the security requirements and guidance that are also applied to these shipments. It then summarizes transport packaging and operational experience associated with the various elements of the fuel cycle, and discusses current technologies and future trends.

Key words: packaging, package, transport operations, transport safety, transport security.

18.3 Introduction

It has been stated that the transport of radioactive material provides ‘the lifeblood of the nuclear fuel cycle’ (Malesys, 2011). The basis for this statement is that, by connecting all its various phases, transport allows the nuclear fuel cycle to function (Fig. 19.1). This figure, and this chapter, both focus on the uranium — based fuel cycle. However, it is possible to have a similar thorium-based fuel cycle, and the packaging and transport issues addressed herein would apply equally well to such a cycle.

Figure 19.1 combines two typical variations of the uranium-based fuel cycle: (a) the closed fuel cycle, where spent fuel (SF) is reprocessed providing uranium and plutonium for producing new fuel; and (b) the open fuel cycle (or the once — through fuel cycle) where SF is treated as waste and is not reprocessed. It is noted that SF is also sometimes called spent nuclear fuel (SNF) and, for the open fuel cycle, SF is sometimes called irradiated nuclear fuel (INF). The closed fuel cycle depicted and used in this chapter is based upon recycling of uranium oxide fuel, resulting in the production of mixed oxide (MOX) fuels utilizing recycled plutonium and depleted uranium as the feedstock for the MOX fuel, which would then be utilized in light-water reactors (LWRs).

Each circle in Fig. 19.1 represents an activity at a facility in the fuel cycle, while each arrow represents the transport of one or more types of radioactive material between two activities. The numbers in small boxes (e. g. 19.5.1) represent the section in this chapter where the packaging and transport associated with the carriage of that material (e. g. for 19.5.1 the material is uranium ore) is discussed.

In some cases, multiple transport operations are possible from a given facility. For example, with the power generation activity, two arrows show fresh fuel arriving (either fresh uranium fuel or fresh MOX fuel), and there are three arrows emanating from the activity representing, respectively, the transport of SF from the power reactor to (a) a reprocessing facility, (b) a SF and HLW disposal facility and (c) a SF storage facility.

For SF being transported from the power generation facility (nuclear power plant), one arrow depicts shipment directly from the power plant to SF disposal. In some cases (e. g. for many power plants in the US), spent fuel is stored on site (either in reactor pools or dry storage facilities) for considerable time, and ultimately — when a SF disposal facility becomes available — will be transported directly from the power plant to the disposal facility. Similarly, for HLW generated at the reprocessing facility, one arrow depicts shipment directly from the reprocessing facility to the HLW disposal facility. This is possible in the event that the HLW is stored on site at the processing facility until an HLW disposal facility becomes available, at which time the HLW will be transported directly from the processing facility to the disposal facility. Of course, in both cases, transport of the SF and HLW from the power generation and reprocessing facilities to an interim storage facility is also possible.

In addition to the transport arrows shown in Fig. 19.1, each of the activities represented by the circles may generate low-level radioactive waste (LLW) and/ or intermediate-level radioactive waste (ILW), which will need to be packaged for transport and then shipped to appropriate waste disposal facilities. Also, at the end of life of any of the facilities, their decontamination and decommissioning will lead to additional transport of radioactive wastes (some of which may be very large components) for disposal. These are not depicted in Fig. 19.1 but they are discussed in sections 19.5.9 and 19.5.10, respectively.

The following addresses both the safety and security aspects of the packaging and transport of the key radioactive materials associated with the fuel cycle as just discussed. It also summarizes transport packaging experience, operations and current technologies, and discusses future trends — all associated with the fuel cycle.

Canada

In Canada, uranium ores first came to public attention in the early 1930s when the Eldorado Gold Mining Company began operations at Port Radium, Northwest Territories, to recover radium. Radium is one of the decay products of uranium and is therefore found in all uranium ores.

Exploration for uranium began in earnest in 1942, in response to a demand for defence purposes. By 1956 thousands of radioactive occurrences had been discovered and three years later 23 uranium mines with 19 treatment plants were in operation. The main production centre was around Elliot Lake in Ontario, but northern Saskatchewan hosted some plants. This first phase of Canadian uranium production peaked in 1959 when more than 12 000 tonnes of uranium was produced. This uranium yielded more in export revenue than any other mineral export from Canada that year.

In response to the development of civil nuclear power, uranium exploration revived during the 1970s, with the focus on northern Saskatchewan’s Athabasca Basin. The Rabbit Lake, Cluff Lake and Key Lake mines started up 1975 to 1983. Exploration expenditure in the region peaked at this time, resulting in the discoveries of Midwest, McClean Lake and Cigar Lake. Then in 1988 the newly formed Cameco Corporation discovered the massive McArthur River deposit.

Canada’s share of known world uranium resources is currently about 8%, but it produces almost one fifth of the mined uranium supply (almost 9145 tU in 2011) making it the second largest producer in the world behind Kazakhstan. Most uranium is exported, but about one fifth is used domestically.

Canada has made a transition from second-generation uranium mines (started 1975-1983) to new high-grade ones, all in northern Saskatchewan, making its uranium mining operations among the most advanced in the world.

Cameco operates the McArthur River mine, which started production at the end of 1999. Its ore is milled at Key Lake, which once contributed 15% of world uranium production but is now mined out. Its other former mainstay is Rabbit Lake.

McArthur River has enormous high-grade reserves of over 20% uranium ore at a depth of about 600 metres. It opened at the end of 1999 and is now the largest uranium mine in the world by a wide margin. Remote-control raise-boring methods are used for mining and the ore is trucked 80 km south to the modified Key Lake mill, where it is blended with ‘special waste rock’ to produce 7200 tU/ yr. Tailings are deposited in a mined-out pit. Cameco is the operator and majority owner, with Areva (30.2%) as partner.

Areva Resources operates the McClean Lake mine, which commenced operation in mid 1999. It has new plant and other infrastructure and uses the first mined-out pit for tailings disposal (the ore having been stockpiled). McClean Lake has four open pits and later will become an underground mine. Annual production depends critically on ore grade being treated, though the mine has recently been relicensed to produce as much as 3100 tU/yr.

Cigar Lake will be a 450 m deep underground mine in poor ground conditions, using ground freezing and high-pressure water jets for excavation of ore. High — grade ore slurry from remote mining will be trucked for treatment at Areva’s expanded McClean Lake mill, 70 km north-east to produce 7000 tU/yr from about 2013. A major flood in 2006 and another in 2008 set the project back several years and pushed costs up from C$660 million to more than C$1.8 billion. The joint venture is managed by Cameco which holds 50%, while Areva holds 37%.

Areva’s large Kiggavik deposit in the Nunavut Territory has evident potential, as do several other smaller but significant deposits.

General properties of thorium-based fuels in reactors

The high fission efficiency of U-233 (the ‘eta’ value) results in a swing in the fissile content and reactivity over the in-core lifetime of thorium-based fuel. However, this effect is much smaller than in uranium fuel. Thus, thorium reactor cores are more manageable than uranium cores because, over the lifetime of the core, the variations in reactivity and power distribution (power peaking) are less. Further, at low average fluxes and low burn-up, the in situ breeding is also better than for uranium fuel. This property gives significantly greater flexibility to programmes based on thorium use from a reactor operation standpoint. In general, the conversion factor in current thermal reactors is higher for Th-U/233 cycles, compared to uranium-plutonium cycles (the increase is usually between 20 to 30%).This point will be further developed at the end of this section.

It must be mentioned also that U-233 is much more flexible in thermal reactors than plutonium, because its nuclear properties (mainly its cross sections) provide greater margins for fuel management in the core. Partly, this is because the three main isotopes of plutonium have great resonances at a very low energy, which complicate the neutronic behaviour of plutonium fuels.

Other specific features of thoriumbased fuel are as follows:

Thorium and its oxide (ThO2 ) have better behaviour under irradiation than uranium and its oxide (UO2), allowing higher burn-ups. This is a consequence of the higher melting point and superior thermal conductivity of both thorium and thorium oxide when compared to uranium and UO2.[17] Furthermore, the chemical interaction of metallic thorium with water and steam is less intense than for metallic uranium.

There is a significant weaker neutron spectrum dependence of U-233 thermal cross sections, compared to those of plutonium isotopes. This is favourable for reactor safety (temperature effects) and operation (power changes), especially when switching LWR from ‘cold’ to ‘hot’ conditions (and conversely).

The yield of fission products affecting reactor poisoning during operation (such as xenon and samarium) is significantly lower for U-233, compared to U-235 and plutonium. The average cross-section values of neutron absorption by U-233 fission products is decreased by about 25 to 30%. Hence, reactivity loss is lower and the core lifetime (i. e. burn-up) increases. This also contributes to a better global neutron economy.

In terms of reactor operation and fuel performance, therefore, fuel based on Th/U-233 has many advantages over uranium-plutonium based fuel.

Conversely, one of the main drawbacks to the use of thorium fuel in reactor cores is the production of Pa-233, a neutron absorber, in rather high concentration. This is explained by the relatively long decay period of Pa-233 (27 days half-life forming U-233), compared to its equivalent in uranium fuel, Np-239 (2.3 days forming Pu-239). It results in a ‘delayed reactivity’ increase after reactor shutdown that must be carefully accounted for.

In a reactor, the rate of loss of neutrons by Pa-233 capture is proportional to the number of neutrons, thus to the neutron flux, and to the capture cross section of Pa-233, which is high for thermal neutrons. Consequently, the concentration of Pa-233 during reactor operation is particularly penalizing for high flux thermal neutron reactors. The loss of a Pa-233 nucleus by neutron capture is equivalent to the loss of a U-233 nucleus, which would otherwise have been formed by normal radioactive decay of Pa-233. This phenomenon leads to a significant reduction of the conversion ratio. Because it is most significant at high thermal neutron fluxes, studies of thorium fuel have tended to be done on reactor cores with low thermal neutron flux and, therefore, low power density. Such cores are not so attractive economically, of course, but they are a feature of ADS, which is one reason why these reactors have been considered to be particularly suitable for thorium fuel.

As with all reactors the fission process is controlled by balancing the reactivity in the core. The core itself contains structural materials, which absorb neutrons, and the fission process is influenced by the energy of the neutrons and hence the effectiveness of the moderation. Reactivity is controlled by means of absorbers in the form of control rod clusters and by the addition of dissolved absorbers to the water. The control rods contain either a silver/indium/cadmium alloy or in some cases hafnium. Boric acid is used as the dissolved absorber to provide bulk reactivity control.

Since PWRs are periodically shutdown for refuelling the reactivity held in the core at the beginning of the cycle (BOC) must be high enough to last for the complete cycle. This is achieved by enriching the fuel to increase the proportion of 235U, which is about 0.7% in natural uranium, to between 3 and 5%, depending on the cycle length. The fission processes will consume the 2 35U and so reduce reactivity. In addition some of the fission products, referred to as ‘neutron poisons’, act as absorbers, further reducing reactivity. However, neutron capture by 238U will lead to the formation of 2 39Pu, which is fissile and so adds reactivity. A proportion of the plutonium will be consumed later in the cycle.

The core reactivity therefore falls during the cycle and is compensated for by reducing the amount of dissolved boron in the water. This is achieved by the chemical and volume control system (CVCS) (Fig. 10.82 . A small amount of water passes through the CVCS system at all times to clean up the coolant, including the removal of radioactive material. It can also be used to either add borated water or demineralised water to change boron concentration. The advantage of using dissolved boron is that the absorption is uniform and so the neutron flux profile across the core remains undisturbed.

The absorber rods provide rapid control; they are divided into different groups with some designated for control and the others held out of the core for shutdown. The boron concentration is maintained so the plant normally operates with the control rods just inserted into the top of the core. In this position the control is sensitive and rapid shutdown can be achieved by dropping the control and shutdown rods under gravity. Fast power changes will be carried out using the rods but slower variations including compensation for burnup will be achieved by changing the boron concentration.

Changes in boron concentration involve adding more concentrated boric acid or adding water. In both cases fluid must be drained from the circuit and treated by the waste water plant. To avoid the need to treat large quantities of waste water some plants make use of so called ‘grey rods’. When normal (‘black’) rods are inserted into the core this changes the flux shape (see Pouret et al, 2009) because they essentially absorb all the neutrons that impinge on them. This increases peaking factors as well as leading to non-uniform xenon transients. Grey rods are

Containment sump

10.8 Chemical and volume control system (Source: USNRC).

Rod Cluster Control Assemblies (RCCAs), which contain lower densities of absorber material and so do not absorb all the incident neutrons. They can therefore be more deeply inserted into the core without unduly disturbing the flux shape. They are used to change load without the need to change the concentration of the dissolved boron. They do, however, still have local effects on the flux and therefore do affect fuel utilisation.

As was noted above, PWRs are under-moderated, which means that increases in moderator density will increase reactivity. This tends to lead to a negative moderator temperature coefficient (MTC), which is good for control. However, because the moderator generally contains dissolved boron, the extent of the effect will depend on the boron concentration since changes in the moderator density will also change the absorber content. At the end of cycle (EOC) when the boron concentration will be very low the MTC will be strongly negative. However, high boron concentrations at BOC could give a positive MTC. To avoid this, burnable poisons are introduced. Burnable poisons were initially used in the form of discrete rods, which were inserted into fuel assemblies that did not contain RCCAs. These provide negative reactivity early in the cycle but are burnt out during the cycle. It is now more common to use integral burnable poisons in which the burnable poison is incorporated into the fuel rods either by mixing (e. g. gadolinia) with the fuel or as a surface coating (e. g. zirconium diboride) on the fuel pellets.

In the UK, the preferred method for handling discharged fuel has been to store it underwater in fuel cooling ponds. The reactivity of the cladding requires an elevated pH to prevent corrosion, and the usual method is to dose with sodium hydroxide to a pH of around 11. Reduction of corrosion and removal of decay heat requires cooling plant, with typical operating temperatures of around 13 °C.

Many early Magnox reactor sites had open-air cooling ponds. This led to problems of control of pond conditions, and admission of debris in the form of atmospheric dust and bird guano. High winds at coastal sites led to foaming of the water surface and airborne activity being released from the pond to the surrounding land. As a consequence, the ponds were eventually all roofed.

The ponds were equipped with clean-up plant to remove particulate and dissolved activity. Typical plant comprised sand pressure filters for removal of solids (magnesium and aluminium hydroxides and the, fission product strontium) and ion exchange beds for removal of (principally) caesium isotopes.

Chapelcross Power Station did not chemically dose its ponds, but relied on corrosion of the fuel cladding to raise the pH. It was unique in that each of its two ponds was emptied for cleaning on a two-yearly cycle, and radio-caesium levels were controlled by pumping water through in-pond enclosed skips filled with zeolite.

In later years, several fuel storage ponds were equipped with in-pond IONSIV units. These comprised a pumped ion exchange resin cartridge with an extremely high degree of affinity for caesium, equipped with a pre-filter to remove particulate and a post-filter to prevent highly active resin fines being returned to the pond. These have been very successful in controlling water activity levels.

Fuel storage was in open-topped boxes (skips), designed to fit into transport flasks to enable the fuel to be removed to the Sellafield reprocessing plant. (Hinkley Point A had a unique design of storage skip, necessitating transfer to a transport skip). Other than at Hunterston A, which used aluminium skips, all skips were made of painted mild steel.

UK Magnox ponds were equipped with machinery for removing splitters or lugs from the fuel, enabling a higher density of packing in the transport skips. A variety of designs of desplittering or delugging machine were used, some more successfully than others. Desplittering machines fell into two categories: ram-and-die, in which the fuel element was forced through a die thereby removing splitters or lugs, and designs in which jaws closed around the fuel element, severing the bands holding the splitter cage assemblies, which then sprang off. The latter designs were by-and-large good at avoiding fuel damage; however, the ram-and-die machines frequently damaged the end fittings of fuel elements, leading to exposure of fuel or breaking of the fuel element. Later designs of ram — and-die machine used a split die, which opened as the end fitting passed, thereby reducing the potential for damage.

Fuel element debris (FED), comprising splitter cages or lugs, frequently with end fittings containing Nimonic springs and occasionally fuel fragments, was removed for storage in FED vaults. The Nimonic springs in particular were highly active, each containing Co-60 at GBq levels, resulting in handling challenges for the stored FED.

Difficulties in desplittering, and transport issues, led to large backlogs of fuel in some Magnox ponds. As a consequence, extensive fuel corrosion was an issue at many sites, with the ponds becoming highly contaminated, and access to the pond environment requiring respiratory protection.

Wylfa Power Station was unique in the UK in having dry fuel storage facilities. Discharged fuel was initially stored in large carousels cooled under a CO2 atmosphere, before being moved to a longer term air-cooled store in skips similar to those used in ponds. Problems of fuel corrosion under water were thereby avoided, along with the much reduced need for water treatment plant and reduced arisings of mobile wastes (sludges and resins).

In the case of ceramic fuel pellets, gas atoms and vacancies in the as-manufactured pores can diffuse into the surrounding fuel matrix or be ejected into the matrix by the disruptive action of fission fragments. Once in the matrix, the vacancies diffuse to grain boundaries where they are absorbed. The result is a densification (i. e. a reduction in volume) of the fuel pellets over the initial few months of irradiation. The increase in the pellet-cladding gap size tends to increase fuel temperatures, while the increase in fuel thermal conductivity due to the reduction in the porosity volume fraction tends to decrease fuel temperatures. Since the

former is a stronger effect, the net outcome is an increase in maximum fuel temperature at any given power. The extent of densification is dependent upon the fuel type, the initial porosity volume fraction, the porosity size distribution and the evolution of the fuel temperature and fission density distributions with time.

Concrete casks are movable structures with a single storage cavity. They are used for storage and sometimes for transport of spent fuel. Structural strength and radiological shielding are provided by reinforced regular or high-density concrete.

A concrete cask/container system may use sealed metal canisters or metal liners inside the storage cask to contain radioactivity. Some canisters can be removed from the concrete cask and are licensed for transportation in an off-site transportation package. Dissipation of spent fuel decay heat relies more on conductive heat transfer to the surface of the container and external natural convection by air. Concrete casks use single or double lid closure systems. Lids are welded closed and tested for leak tightness. In most cases the canisters or liner are backfilled with an inert gas to prevent corrosion and to improve heat transfer within the canister. Nitrogen or helium gases are most often used. An example of a concrete container is shown in Fig. 15.11. It is a dry storage container (DSC) for

Seal/structural weld

Vent port

Reinforced

high-density

concrete

Steel inner liner

Steel outer hner

Drain port

15.11 Dry storage container (DSC) for CANDU-type fuel (courtesy of Ontario Power Generation).

HWPR CANDU fuel where the fuel bundles are placed horizontally in the spent fuel rack. In the design and licensing of dry storage containers, the key parameters are the maximum burnup of the nuclear fuel (GWd/tHM) and the maximum heat load to the storage unit (kW).

Silos

Silo systems are monolithic or modular concrete-reinforced structures. The concrete provides shielding while containment is provided by either an integral inner metal (liner) vessel, which can be sealed after fuel loading, or by a separate metal canister. In silos, spent fuel may be stored in vertical or horizontal orientation. Fuel loading always takes place at the storage site. The NUHOMS storage system is an example of a horizontal concrete silo system. Fuel is loaded vertically into metal canisters, which are stored in a horizontal orientation inside concrete storage modules as shown in Fig. 15.12.

I rradiated MOX fuel contains more plutonium and minor actinides than the standard fuel. However, it has been demonstrated that the PUREX process works with irradiated MOX fuel without major difficulties.

In 1991-2, the CEA (French Atomic Energy Commission) reprocessed, in a pilot plant at Marcoule, about 2.1 tonnes of MOX fuel from the German Grafenrheinfeld nuclear power plant. And in 1992, 4.7 tonnes of MOX fuel from the German Obrigheim-Neckar-Unterweser nuclear power plant were reprocessed at La Hague.

Existing MOX contains 5.3% of plutonium. With irradiation, the concentration of even-numbered isotopes in the fuel increases — mainly plutonium-242. Since these isotopes are non-fissile in a pressurized-water reactor (PWR), their neutronics hinder the chain reaction. There are two possible solutions: an increase in either the plutonium content (taking it beyond 5.3%) or the uranium-235 content. The first solution would limit the number of possible recycles because of the build-up of Pu-240. The second is considered to be more promising. On this basis, several recycles of MOX fuel may be possible. Considering the relatively long time for the radioactive and thermal decay of irradiated MOX, a single cycle is likely to take about 12 years and two or three cycles will take 24 or 36 years. Provided that life extensions are permitted, these could be accommodated by existing reprocessing plants. In coming to such decisions, important points to be considered are the impact of the irradiated MOX composition on waste, effluent and releases and the change in composition of the MOX that might be introduced along with the recycles.

18.1.1 Origins of radioactive waste

While one might expect that ‘radioactive waste’ should be easily defined, the IAEA3 offers three separate definitions. The first is adequate here:

For legal and regulatory purposes, waste that contains, or is contaminated with, radionuclides at concentrations or activities greater than clearance levels as established by the regulatory body.

This definition recognises that, because all matter is radioactive to some degree, it is necessary to define a level above which it becomes ‘officially radioactive’ and subject to regulation. Radioactive waste exists in many forms and is present in virtually every country of the world. In describing its origins, it is convenient to place countries into one of three groups:

1 those with no nuclear power plants (NPPs)

2 those with NPPs

3 those with NPPs and other fuel cycle facilities

Wastes in the absence of nuclear power

In countries with no nuclear industry, radioactive waste will mostly consist of disused sealed sources. Typically, the number of sealed sources in store in a medium-sized developing country will be a few thousands. If a country has oil, gas or mineral deposits then it is likely that some wastes will be in the form of naturally occurring radioactive material (NORM). Often this is produced as so-called ‘pipe scale’: radium-rich mineral deposits that form inside pipework used to transport process water. The level of activity is generally quite low — perhaps up to a few thousand Bq/g — and it can be removed by water jetting and removed as sludge. This allows the pipes to be reused or released as scrap. In some cases it may be possible to pump the sludge underground: back into the formations from whence it came. More usually, the material is dried, drummed and put in storage awaiting a permanent solution. Finally, countries in this first group may also possess one or more low-power research reactors. Without exception, these are donated by some other country, often with an agreement that the spent fuel will be repatriated when the reactor is decommissioned.