Metal casks, concrete casks and silos are variations of the container storage systems. Common to all cask and silo designs is that they: [29]

Metal casks The structural material for metal casks may be forged steel, nodular cast iron or a steel/lead sandwich structure. The casks are fitted in most cases with an internal basket or metal canister. They usually have a double lid closure system, which may be bolted or seal welded and may be monitored for leak tightness. In many countries, dual-purpose metal casks were developed and are licensed for storage and off-site transportation.6 One example of a metal cask for storage and transportation is shown in Fig. 15.10. In the USA, a triple-purpose cask was developed (TAD: transportation, aging, disposal). There are problems with the application of containers for triple purposes. It is obvious that the licensing process is more complicated for multipurpose casks and containers as in some cases separate licences must be obtained for each operation. The licensing of a cask for disposal, with uncertainties about necessary compatibilities when there is no fuel disposal facility in operation, is difficult.

Drain and vent orifice

Secondary lid

Resin

Basket for 37 _ PWR assemblies

Trunnions

Resin

Heat conductors

Rear shock absorbing cover

15.10 TNT M24 transport and storage cask (courtesy of AREVA).

Recycling in an ordinary water reactor

Mixed oxide (MOX) fuel provides about 2% of the new nuclear fuel used today and is manufactured from plutonium recovered from used reactor fuel. Its use also provides a means of burning military weapons-grade plutonium to produce electricity. Reprocessing to separate plutonium for recycling as MOX is increasingly being seen as more worthwhile from an economic point of view as uranium prices increase.

European and Japanese reactors use MOX extensively. It is generally employed as about one third of the core of reactors in these areas, but some reactors will accept up to 50% MOX assemblies. The EPR or AP1000, among other advanced light-water reactors, will be able to accept complete fuel loadings of MOX if required. If up to 50% of MOX is used, the operating characteristics of a reactor are not changed. A plant must be designed from the outset to take this fuel though, or adapted as necessary (for example more control rods are needed). If more than 50% MOX is used, plants either need to be designed from the outset to accommodate this, or changed significantly.

One advantage of MOX is that the addition of small extra amounts of plutonium easily raises, the fissile concentration of the fuel, whereas enriching uranium with higher levels of U-235 is relatively expensive. MOX use is becoming more attractive since reactor operators are looking to burn fuel harder and longer, increasing burn-up from around 30 000 MW days per tonne, which was typical earlier in the century, to over 50 000 MWd/t nowadays. MOX use is also attracting more interest since reducing the volume of spent fuel is increasingly desirable (World Nuclear Association, MOX, Mixed Oxide Fuel, 2009).

I. CROSSLAND, CrosslandConsultingLtd, UK

Abstract: This chapter provides a description of key issues in disposal of solid radioactive waste namely the nature of the waste itself, the measures needed to convert it to a form that is suitable for disposal, the institutional arrangements that allow disposal to take place and, finally, the disposal methods currently in use or proposed to be used. The chapter concludes with a discussion of possible future developments in this field.

Key words: radioactive waste disposal, high level waste (HLW), spent nuclear fuel (SNF), borehole disposal, vitrification.

18.1 Introduction

Industrial waste is an inevitable consequence of human exploitation of the environment. Today, the old tin and copper mines of Cornwall are tourist attractions but, when applying to have a group of mines registered as a World Heritage Site, the area was said to include some of the most polluted land in the UK1 and a warning was given that remediation could disturb areas associated with historic former mines. It seems that, given enough time, even industrial pollution can come to be seen as a form of heritage.

What our successors will think of radioactive waste — a by-product of twentieth and twenty-first century nuclear technology — is impossible to say but one of the key principles of its modern-day management2 states that its production should not place an undue burden, whether technical or financial, on future generations. This immediately rules out any form of storage as a potential solution, remembering that, according to the IAEA,3 all storage is temporary because it implies retrieval. What is needed is something permanent. Tasked with finding a solution for the long-term management of the UK’s radioactive wastes, the Committee on Radioactive Waste Management (CoRWM) examined every possible option, including firing the waste into space, sea dumping and deposition into subduction zones. After two years’ deliberation, the committee recommended4 geological disposal as ‘the best available approach’.

The aim of this chapter is simply to describe the nature of radioactive waste and the techniques developed for its safe disposal:

• Section 18.2 describes the various kinds and categories of radioactive waste and the manner of their arising.

• Section 18.3 covers pre-disposal, by which is meant processing and conditioning of radioactive waste for storage and disposal.

• Section 18.4 addresses the framework for disposal: the organisational arrangements needed to drive and control a disposal programme.

• Section 18.5 outlines modern practice in radioactive waste disposal and is divided into sub-sections on near-surface disposal, deep disposal and borehole disposal.

• Section 18.6 looks at possible future trends.

• Section 18.7 suggests some further reading.

The Transport Regulations — TS-R-1 (IAEA, 2009a) — establish a sound basis for both the design of packages used for the transport of radioactive materials and the operations associated with their use.

19.4.1 Regulatory requirements for packages

The packages used for the transport of radioactive material are structured into five categories: (1) excepted packages, (2) industrial packages, (3) Type A packages, (4) Type B packages and (5) Type C packages; where the activity and/or risk posed by the contents, and the design requirements both increase from the excepted packages to the Type B and Type C packages.

Excepted packages are used for the transport of very small quantities of radioactive material. Minimal design and operational requirements are imposed on these packages because of the very low risk posed by their contents. They are expected to be able to survive routine conditions of transport (i. e. incident free) without loss of contents.

Industrial packages, which come in three types (IP-1, IP-2 and IP-3), are used for the transport of low specific activity material and surface-contaminated objects, where the risk posed by these contents and, therefore, the design requirements again increase from IP-1 to IP-3. Over the range of contents and package requirements, these packages might be expected to retain their shielding and containment integrity under both routine and normal conditions of transport (i. e. minor mishaps), depending upon the level of activity of the contents involved.

Type A packages are used for the transport of limited quantities of radioactive material, where the content limits are established by the individual radionuclides using a methodology for assessing risks, which assumes that the packaging is capable of retaining its shielding and containment integrity under both routine and normal conditions of transport.

Type B packages are used for the transport of higher activity contents (i. e. higher than Type A quantities). Their designs undergo a rigorous review and certification by competent authorities before they can be used. They are designed to withstand, without significant loss of shielding or containment (the values of which are specified quantitatively in the Regulations), not only routine and normal conditions of transport, but also accident conditions.

Type C packages are similar to Type B packages, but must be designed to even more stringent conditions and are for use in the air transport of significant quantities of radioactive material. The accident conditions of transport that are imposed on these designs are more demanding than those imposed on Type B packages.

In addition to the above, if the contents of the package are fissile, they must be assessed under normal and accident-related conditions to demonstrate that criticality will not occur. For fissile material, the packages may be either Industrial — Fissile (Type IF), Type A Fissile (Type AF) or Type B Fissile (Type B(U)F or Type B(M)F where the ‘U’ and the ‘M’ designate unilateral approval and multilateral approval requirements as set forth in the Regulations).

Finally, if the package contains uranium hexafluoride, specific design requirements apply, and they are given the unique identifiers H(U) and H(M), unilaterally and multilaterally approved, respectively.

Although this may appear to be a complex structure for specifying regulatory packaging requirements, it is structured in a sound manner, uses a graded approach, and is based upon the experience of the development and application of the Regulations for over 50 years.

To illustrate the graded approach, Table 19.2 shows the increase in package design requirements as the contents move from the very small quantities of radioactive material allowed in excepted packages, to the very large quantities that are allowed in the Type B and Type C packages. The increase in design requirements from Type B to Type C results from the potentially more severe environments during normal transport and accidents that a Type C package might experience in air transport. The table does not show the additional requirements imposed on fissile or uranium hexafluoride packages.

Package design / test requirement*

General requirements for all packagest

Increased capability to withstand specific temperature and pressure environments expected during transport by air

Limit smallest package dimension to 10cm

Withstand drop onto an unyielding target from a height of 0.3 to 1.2 m depending upon the mass of the package; and compressive load equal to greater of 5 times the package mass or 13 kPa

Incorporate a seal providing evidence the package has remained unopened, and provide a securely closed containment system

Incorporate tiedown attachments so that package integrity will be retained during normal and accident conditions of transport Design to package temperatures ranging from -40°C to +70°C; and retain contents to reduced pressure of 60kPa Design to national and international standards

Withstand a water spray test for one hour and a penetration test by a 6kg bar dropped 1 mf

Withstand a drop from 9m onto an unyielding target so as to suffer maximum damage

Withstand a drop from 1 m onto a rigidly mounted bar so as to suffer maximum damage

Industrial Type A Type В Type C

packages package package package

Package design / test requirement Excepted Industrial Type A Type В Type C package packages package package package Withstand impact of 500kg plate from a height of 9m (some Type В and all Type C packages) # X Withstand a thermal test at 800°C for 30min so as to suffer maximum damage X Withstand 8 hour immersion in water at 15m depth in attitude for maximum damage X X Withstand 1 hour immersion in water at 200 m depth (some Type B, all Type C packages) # X Withstand burial at steady state in a thermal insulating environment X Withstand drop of 250 kg puncture probe from height of 3m forsmall packages, or drop of package onto probe for large packages X Withstand a thermal test at 800°C for 60 min so as to suffer maximum damage X Withstand impact onto unyielding target at 90 m/s velocity so as to suffer maximum damage X

Notes:

* ‘#’ applies to some of the packages within that type, ‘X’ applies to all of the packages within that type.

t General requirements are covered in paragraphs 606-616 of TS-R-1; addressing such items as properly securing the package, ensuring lifting attachments cannot fail, having easily decontaminated external surfaces and preventing collection of water, capable of withstanding routine transport accelerations and vibrations, having physically and chemically compatible materials, protecting any valves, and taking into account routine transport ambient temperatures and pressures.

t A drop height of the 6kg bar of 1.7 m is required for Type A packages containing liquids or gases.

Packaging and transport of radioactive material 575

Main results

Table 5.3 presents the application of Eq. 5.3 when used with the data just described and a discount rate of 7.5%. Decommissioning, fuel, carbon and O&M costs feed directly into the LCOE value while the capital cost is discounted. When the cost of carbon is zero, gas — and coal-fired generation are cheapest. When, however, a carbon cost is included at ($50 per tonne of CO2) nuclear is the cheapest followed by gas. Interestingly, even with CO2 at $50 per tonne, the addition of carbon capture technology (at the coal prices used) does little to make coal a more economic option.

Three of the technologies — nuclear, coal with carbon capture and wind — share a common feature in that all are capital intensive. In all three cases capital costs make up more than about 50% of the LCOE. As expected, the LCOE for gas generation is dominated by fuel and carbon costs, which comprise around 80% of the total.

It is noteworthy that onshore wind struggles to be competitive with the other technologies even with CO2 priced at $50 per tonne. This is a consequence of the high capital cost which, in turn, results from the low availability of this technology. This clearly demonstrates that siting — finding reliably windy places — is an important determinant of the cost of the electricity produced by this technology

and explains why, despite significantly higher overnight costs, offshore wind might be considered viable.

Because thermal efficiency affects fuel and carbon costs, it is clear that improvements in this parameter will mostly affect technologies where fuel and carbon make a large contribution to the LCOE, i. e. coal and gas. Thermal efficiency improvements in nuclear plant, while obviously welcome, will not have a large impact on overall costs.

JEFF WILKS, URENCO UK Limited, UK

Abstract: Most nuclear power stations use fuel that has concentrations of the uranium-235 isotope that are higher than found in natural uranium. The process of increasing the concentration of uranium-235 above natural is called enrichment and this chapter describes the main enrichment technologies that are in use or have been investigated. The most important enrichment technologies utilise uranium hexafluoride as a feed material and so this chapter also summarises its properties, describes processes used to manufacture it and considers a number of important issues associated with its use.

Key words: uranium hexafluoride, uranium conversion, uranium enrichment.

7.1 Introduction

The precursors to the development of nuclear power were nuclear programmes carried out for military purposes. The history of these military programmes has been documented on many occasions and is not the subject of this chapter. What is of relevance is that early atomic weapons were based on uranium and in particular, the potential for uranium to sustain a nuclear chain reaction releasing huge amounts of energy. When released in large amounts over a very short time period then this energy release forms the basis of a nuclear weapon. When the energy release is controlled, collected and used to generate electricity then it forms the basis of nuclear power generation.

Natural uranium contains three isotopes. Uranium-238 (238U) is the bulk isotope comprising over 99% of the total with uranium-235 (235U) present at around 0.71% by weight and uranium-234 (234U) at 0.0053%. Of the three isotopes, only 235U is fissile and capable of sustaining a nuclear chain reaction. The proportion of 235U in natural uranium is sufficient to sustain such a reaction under very specialised conditions, for example when using heavy water as a moderator. It is not sufficient, under any circumstances, to sustain a reaction of sufficient intensity for use in nuclear weapons. The weapons programme therefore required that technology be developed that would allow the proportion of 235U in uranium to be increased to very high levels, typically greater than 90%. The process of increasing the proportion of 235U to levels above that found naturally is known as enrichment and its application to civil nuclear power generation is the subject of this chapter.

Enriched uranium is not necessary for nuclear power generation. The first reactor to generate electricity on an industrial scale, Calder Hall in the UK, used natural uranium and successful commercial designs such as the British Magnox reactor and the Canadian CANDU reactor have done so since. The technology to enrich uranium was already established, however, and as nuclear power reactor designs were developed around the world, many of those designs chose to make use of low enriched, rather than natural uranium. Low enriched uranium allows higher power densities than can be achieved in reactors using natural uranium and generates less spent fuel, a waste requiring careful management. It also allows normal water to be used as the moderator and heat transfer medium, rather than the heavy water used in the CANDU and the graphite moderated, carbon dioxide cooled combination in Magnox reactors. Over 90% of nuclear power is now generated from low enriched uranium, typically containing 235U at levels of between 3% and 5% by weight.

The process of mining and refining uranium produces a uranium ore concentrate often referred to as yellowcake, the main component being triuranium octaoxide (U3O8). This material is impure and not suitable for enrichment and therefore needs to be converted into another chemical form. For reasons that are explained later, most enrichment technologies and specifically the two main technologies of gaseous diffusion and gas centrifuge require that the yellowcake be converted into uranium hexafluoride (UF6 , or ‘hex’) to allow enrichment to take place. This chapter describes the enrichment process and the main enrichment technologies. It also considers the properties of UF6 that have led to it being used for enrichment and the processes used to manufacture UF, from yellowcake. Finally, some relevant side issues, such as transport, sampling and analysis and tails deconversion are considered.

The design of fuel assemblies is in various ways associated with reactor physics, i. e., the science dealing with neutron chain reactions and nuclear fission. Reactor physics methods are incorporated in specialised computer codes that are able to deal with the geometry and the materials of a fuel assembly. So-called lattice codes (2D) treat a cross section of the assembly and solve the coupled space, energy, angle and time neutron transport problem. They also provide input to whole core calculation codes (3D) and fuel behaviour codes.

For the fuel, reactor physics has to address the effects of enrichment and possibly plutonium in the fuel, burnable absorbers, fuel burn-up (depletion) and the influence of control rods. For materials in an assembly, an important task is to evaluate the neutron fluence (time integrated neutron flux), which has an impact on mechanical properties and dimensional changes, for example growth and bowing in a neutron flux gradient.

Confinement of radioactive material involves the fuel, the circuit and the containment systems, but in this section we will focus on the containment. LWRs generally have pressure-retaining reactor containment buildings. As was noted in 10.2, the coolant in an operating LWR is always radioactive. Thus the containment is provided both to contain the activity, which may be present during normal operation, as well as any fission products released as a result of a fault.

The containment must be designed to withstand the pressure and temperature generated by the steam produced by the release of the primary coolant into the containment following a hypothetical failure of one of the main coolant pipes. In

10.16 Schematic of BWR low pressure ECCS systems (Source: USNRC).

general containments can be divided into dry containment and pressure suppression containment.

Dry containments rely on their size and the heat sink provided by the structure of the containment and its contents to limit the peak pressure following a large LOCA. Pressure suppression containments reduce the peak pressure by forcing the steam released to flow through water in suppression pools or ice baskets. This allows a lower design pressure to be used for the containment structure. The design pressure for the structure is generally set to provide a margin above the peak pressure and is evaluated using conservative design rules. They are subjected to proof pressure tests at above the design pressure. Early PWRs and BWRs used dry containments but more modern BWRs have tended to use pressure suppression containments.

N. CERULLO, University of Pisa, Italy, and G. LOMONACO,

University of Genova, Italy

Abstract: This chapter looks at Generation IV nuclear reactors, such as the very high-temperature reactor (VHTR), the supercritical water reactor (SCWR), the molten salt reactor (MSR), the sodium-cooled fast reactor (SFR), the lead-cooled fast reactor (LFR) and the gas-cooled fast reactor (GFR). Reactor designs and fuel cycles are also described.

Key words: Generation IV Initiative, very high-temperature reactor (VHTR), supercritical water reactor (SCWR), molten salt reactor (MSR), sodium-cooled fast reactor (SFR), lead-cooled fast reactor (LFR), gas-cooled fast reactor (GFR).

13.1 Introduction

From the beginning of this century, there has been increased interest in nuclear energy as the only high-capacity source of CO2-free energy available. When compared to fossil fuels, the waste and emissions generated are minimal (Generation IV International Forum, 2009). At the same time, there has been growing pressure to reduce the safety risks posed by plutonium (Pu) stockpiles and nuclear waste material. Given the Three Mile Island, Chernobyl and Fukushima accidents, there is also pressure to improve the safe operation of nuclear power plants (NPPs), although their safety has reached, at present, a relatively high level of reliability.

At present, world wide nuclear energy production is mainly from light water reactors (LWRs) that are fuelled with uranium enriched up to 5%. The discharge burn-up of a nuclear fuel element is limited by both its fissile content and its endurance, i. e. its ability to withstand exposure to a neutron fluence and high temperature. In the case of LWRs, discharge burn-up lies in a range between 30 and 60 GWd/tHM where tHM stands for tons of heavy metal, this being the initial uranium (U) plus Pu content of the fuel (Bende, 1999). This means that a LWR with an output of 1 GWe and an efficiency of 33% must burn, and subsequently discharge, about 10 to 20 tHM per full power year (FPY). An important fraction of the energy output of the fuel comes from transuranic (TRU) elements that are produced in-reactor by neutron capture.

It will be essential to use Pu as an energy source given the limited availability of other energy sources (oil, natural gas and uranium in the case of nuclear power) in the medium-long term (Cerullo et at., 2009; Bomboni et at., 2008b, 2007). However, the presence of long-lived, high-level radiotoxic elements in the waste from LWRs is becoming a more significant issue, especially from a safety point of view. The use of Pu-based fuel (e. g. MOX) in LWRs, even if it is useful as an energy source, does not allow for significant reductions of actinides because ‘new’ actinides are generated within this type of fuel. Furthermore, this choice leads, at the end of the cycle, to a growth in the quantities of minor actinides (MAs), i. e. neptunium, americium and curium, which are long-lived and very dangerous nuclides due to their radiotoxicity (Cerullo et al., 2009, 2005). Because it is not possible, in existing reactors, to entirely avoid their production, it is necessary to provide for their destruction.

The main drawbacks of LWR technology are therefore the limited exploitation of U resources coupled with the high-level long-term radiotoxicity of the final waste (it takes more than 100 000 years to balance the level of mine (LOM), i. e. to reach the radioactivity of the original ore). For LWR spent nuclear fuel (SNF) 96% is of U, Pu (see Table 13.1) and MA (see Table 13.2). The long-term radiotoxicity of the nuclear waste is essentially due to TRUs, which can produce energy by fission directly or by means of transmutation into fissile nuclides. A promising solution is to burn all the heavy metals (HMs), including MAs, as fuel for nuclear reactors. However, burning all HMs is not straightforward due to a range of open technological and neutronic issues (Bomboni et al., 2008b). Some are related to the very strong gamma and neutronic emissions of many MA nuclides as well as to the different dynamic behaviour of cores with a non-negligible MA inventory (Bomboni, 2009). Nevertheless, some reactor designs seem to be particularly

suitable for burning Pu and, to some extent, MAs (Bomboni et al., 2008a). Many of these new designs are addressed in the Generation IV Initiative.

A restructuring of the UO2 matrix takes place at the pellet periphery in high burnup LWR fuel. The transformed microstructure is characterised by small (sub-micron) grains, which are depleted of fission gas, and a high density of spherical, inter­granular fission gas bubbles. Since the restructuring occurs at the pellet periphery, the resulting microstructure is often referred to as ‘rim structure’. However, the modified microstructure is a result of the local conditions at the pellet rim — in particular the high local burnup and low fuel temperature — rather than of the radial position per se. Hence, it is better referred to as ‘high burnup structure’, or HBS, as originally used by Lassmann et al. (1995).

The appearance of the HBS is reproduced in Fig. 14.2, which shows a scanning electron microscope (SEM) image of a fractured sample of fuel irradiated to a burnup of 73 MWd/kgU (Noirot et al., 2008). The image constitutes a radial scan, with the pellet surface at the right of the image. The rightmost boxed section is enlarged in Fig. 14.3.

The mechanism for high burnup structure formation is as follows. Resonant neutron capture in 238U causes a build-up of 239Pu near to the pellet surface (239U rapidly decays to 239Pu via 239Np) (Lassmann et al., 1994), which results in a high local fission density and burnup. At the low temperatures prevailing at the pellet surface, the resulting fission damage cannot be fully annealed. Eventually, the accumulated damage results in local recrystallisation, usually starting around the margins of as-fabricated porosity. The recrystallisation causes the xenon and

krypton fission gases to be precipitated into small, isolated bubbles. A very similar microstructural transformation is seen in plutonium-rich agglomerates in LWR (U, Pu)O2 fuel, for those agglomerates located in the cooler outer regions of the pellet.

The local burnup and fuel temperature ranges over which HBS formation occurs have been investigated in the High Burnup Rim Project (HBRP) (Kinoshita

et al. , 2004). The results show that the HBS begins to form at a local burnup of ~45 MWd/kgU, and is fully formed at a local burnup of ~70 MWd/kgU, but only at fuel temperatures below ~1100 °C.

The rim porosity increases the pellet volume and therefore enhances fuel swelling. The rim pores also reduce the thermal conductivity of the rim relative to the non-restructured fuel region. There are several possibilities for enhanced fission gas release associated with formation of the HBS: (a) gas is released from the rim region during the recrystallisation process; (b) athermal release is enhanced by the small size of the restructured grains; (c) gas is released from the rim pores; (d) the reduced thermal conductivity of the rim increases fuel temperatures in the non-restructured regions of the fuel pellets, causing enhanced thermal release. Research on the behaviour of the HBS is ongoing; the currently available results suggest that at most only a small fraction of gas generated in the rim region is released to the pin free volume (Bremier et al., 2000). Thus, only enhanced fission gas release due to (d) is thought to be significant.