The main advantages of LWRs are that they are relatively simple and compact. They have been developed over a period of more than 50 years and so the experience has led to the development of robust and efficient plants. They do, however, require the use of enriched fuel and the steam conditions are such that their thermal efficiencies are not as high as some gas-cooled plants. However, they are capable of achieving relatively high fuel burnup so their efficiency in terms of fuel utilisation is good.

LWRs have also proved themselves to be flexible plants, and can contribute to load following and frequency control. PWRs using grey rods and BWRs can load follow over their complete cycle. PWRs which use black rods can also load follow on a daily basis for most of their cycle, but it can be more difficult towards the end of a cycle, not because of any limitations on the reactors, but because of the need to process the large quantities of water involved in making changes to low boron concentrations. The limitation is then associated with the capacity of the boron recycle system or waste water processing system.

Given the time that they have been in operation we now also have experience with the decommissioning of the plants and because of their compact design they have proved to be relatively straightforward to decommission.

The design relies on a high integrity pressure vessel, which potentially limits the size of plant that can be constructed. In addition at present there are relatively few suppliers of large high integrity vessels and their components. The integrity requirements mean that the vessels are factory constructed and have to be transported as a unit. However, the size limitation is probably not an issue since vessels accommodating cores capable of supporting electrical outputs of at least 1800 MW(e) are currently envisaged.

Almost all Generation IV systems use a closed fuel cycle design or, in the case of the VHTR and MSR, offer the possibility of a closed cycle through the use of

thorium-based fuel. Fuel recycling technology is therefore an essential element of Generation IV systems. However, current technology, which involves the separation of pure plutonium oxide, does not meet the GIF requirement for proliferation resistance (Abram and Ion, 2008). Several alternative technologies are under development, which could provide significant improvements in both the safety and economics of closed cycle plants. These are discussed below.

The modelling approaches employed in the five types of computer code identified in Section 14.3.1 are described in this section. The emphasis is on fuel performance codes, since the thermo-mechanical behaviour of the fuel pins is the key aspect of fuel behaviour under irradiation.

Neutronics codes

Given the core geometry and materials, the aim of a neutronics calculation is — put simply — to determine the evolution with time of the spatial and energy distribution of the neutron flux in the core. Use of nuclear cross-section data then allows the corresponding nuclear reaction rates, and hence the core power and temperature distributions, to be evaluated. (The reality is more complicated, since the neutron flux distribution, nuclear reaction rates and core power and temperature distributions are interdependent.) Since the composition of the fuel changes with burnup, which in turn affects the neutron flux distribution, the neutronics calculation must also compute the evolution of the fuel composition with burnup. Of particular importance is the depletion of burnable absorber materials, since these strongly influence the neutron flux.

Due to computational limitations, the neutronics calculation is generally divided into two stages, the first performed by a so-called lattice code and the second by a whole core neutronics code (subsidiary codes may also be employed to perform, for example, core temperature distribution calculations). The lattice code calculates the change in composition of the fuel as a function of burnup for each assembly type. Neutron transport theory with a large number of neutron energy classes, or groups (i. e. multi-group theory), is utilised in two dimensions. An accurate nuclear data library (e. g. JEFF or ENDF/B) is also employed. The neutron transport equations (Glasstone and Sesonske, 1980) are solved via convoluted methods (e. g. the method of characteristics (Knott and Edenius, 1993)), with various approximations to take account of the heterogeneity of the core, the resonances in the nuclear reactions, neutron leakage, etc. The results are used to construct a lookup table of nuclear cross sections and reaction rates as a function of burnup. This lookup table is then an input to the whole core neutronics code, which performs neutron transport calculations in three dimensions for the entire core. To make the problem tractable, the diffusion theory approximation to neutron transport theory (Glasstone and Sesonske, 1980) is generally implemented with only a small number of neutron energy groups (i. e. the calculation is a few-group calculation). The resulting neutron transport equations are solved (typically by the nodal method (Smith and Rempe, 1988)) and the spatial and energy distributions of the neutron flux, and hence the core power distribution, are determined.

Whole core neutronics codes can be of either the steady-state or kinetics (transient) varieties. The former assumes quasi-steady-state conditions throughout irradiation, i. e. that the neutron density distribution is always in equilibrium. This simplifies the equations that need to be solved, thereby allowing a more accurate solution method with high core discretisation. The latter calculates the full time dependence of the neutron density distribution. Since the equations that need to be solved are necessarily more complex, the methods employed for their solution are generally less accurate than for steady-state codes, with more approximations. The core discretisation may also need to be coarser than with steady-state codes to allow simulation over acceptable timescales.

Steady-state codes are used for core design (loading pattern acceptability) calculations where a quasi-steady-state assumption is adequate. Kinetics codes are used for analysis of specific faults as part of reactor safety studies. For faults where the neutronics and the thermal-hydraulics are coupled (e. g. a steamline break in a PWR or power-flow oscillations in a BWR), kinetics codes require either a simplified core/primary circuit thermal-hydraulics model, or coupling to a core/system thermal-hydraulics code.

An alternative to the lattice code plus whole core neutronics code approach to neutronics calculations is use of a Monte Carlo code (Carter and Cashwell, 1975). In this case, the neutron transport equations are ignored; instead, the underlying stochastic behaviour of individual neutrons is simulated. A neutron produced by fission is assigned an initial position, energy, speed and direction by random sampling from the appropriate probability distributions. Given the core geometry, core materials and nuclear reaction cross sections, the evolution of the neutron’s position, energy, speed and direction are then evaluated, taking into account any scattering or fission initiated by the neutron, until the neutron is absorbed or has escaped from the core. If the neutron induces fission, the position, energy, speed and direction histories of the further neutrons generated are also evaluated. This process is repeated many times and the tallies of the neutron path lengths and nuclear reactions in pre determined regions, or cells, of the core, and of neutrons passing through surfaces between cells, are accumulated. These tallies can then be used to estimate the neutron flux and power distributions in the core. The uncertainties in the estimated values decrease as the number of evaluated neutron histories is increased, but at the expense of an increased computation time.

Due to the computational overheads, use of the Monte Carlo method for standard core design is not currently feasible. However, Monte Carlo codes may be used to benchmark (support the predictions of) lattice codes, or for modelling non-standard core configurations for which standard neutronics codes are not applicable or are inaccurate.

In addition to the high-level requirements described above, a wide range of other constraints must also be respected. These include measures to avoid undue radiation exposure of personnel, limits on contamination in manned areas and protection against criticality. Additionally, to avoid any misappropriation of fissile material, physical protection systems are deployed and a rigorous and permanent accounting of materials is monitored by Euratom (European Atomic Energy Community) and the International Atomic Energy Agency (IAEA).

Another consideration is that the reprocessing plant is exposed to high levels of radiation, which may lead to various forms of radiation damage. A particular issue concerns degradation of the solvent by radiolysis in which Pu-238 (half-life 88 years) has an important role. This is created from U-235 and increases in concentration in spent fuel with burn-up.

All these constraints must be taken into account in the design and operations of reprocessing plants.

The key P&T feasibility issues are, it appears, found in fuel cycle performance. Table 17.4 presents two key parameters — decay heat and neutron production after post-irradiation cooling — which largely determine the ease of the fuel fabrication step. Data are shown for some of the transmutation strategies considered previously, comparing these with standard PWR fuel.

The MA/Pu ratios in Table 17.4 correspond to different TRU management strategies. The MA/Pu ratio of -0.1 is used because it corresponds approximately to the TRU composition of PWR fuel at discharge. The ratio MA/Pu -1 corresponds to the case of a fuel with a high MA content, which could be loaded in a dedicated transmuter (FR or ADS).

In the case of homogeneous recycling in PWRs and FRs, the large difference in the neutron source at fabrication is essentially due to the impact of the 252Cf spontaneous neutron fission (-1012 n/g/s) contribution, due to the different mechanisms of its build-up in the two different types of spectrum (see Fig. 17.2). Since 252Cf is a very powerful neutron emitter (see Table 17.1), it results in an unacceptably high neutron source at fuel fabrication. This is one of the factors that suggests that grouped TRU transmutation in thermal reactors should be avoided.6

Moreover, the strong increase in the decay heat and neutron source values observed for a) all low CR critical FRs with high MA/Pu loading ratios, b) TRU target heterogeneous recycling with more than -10% MAs in the target and c) U-free loaded ADSs, can also result in very significant penalties on the associated specific fuel cycle (spent fuel handling, fuel fabrication, etc.). However, it should be noticed that the specific fuel cycle associated with those systems will be only a limited part of the overall fuel cycle, as will be discussed later (Sections 17.4.1 and 17.4.2).

All disposals, at whatever depth, are susceptible to human intrusion but in the case of deep disposal it is widely accepted that this is confined to exploratory drilling, the probability of which is reduced by avoiding sites with mineral resources. In contrast, human intrusion into a near-surface disposal can take many forms such as construction of buildings and roads, pipeline burial and archaeological excavations. A 1987 NEA study25 indicated that excavations for domestic buildings rarely go deeper than 3 m while the maximum depth for a road construction was taken as 10 m. These figures may be compared with recent proposals for intermediate depth disposal in Japan26 and France27 where the specified target depths are 50-100 m and 15-200 m respectively. In general, it may be reasonably assumed that, with a few exceptions, disposal below a few tens of metres will avoid the potential for large-scale inadvertent excavation. The exceptions are largely site specific: the possibility that deep road or railway cuttings or tunnels might be constructed will be determined by the hilliness of the terrain; in places where high-rise buildings could be placed, the foundation depth (normally a few tens of metres) will be dependent on soil strength.

It is tempting and sometimes even useful to equate low-, intermediate — and high-level wastes with near-surface, intermediate-depth and deep disposal, respectively. There is a difficulty, however, because containment and isolation do not inevitably increase monotonically with depth. It is possible, for instance, that a bedded salt formation lying between 200 and 300 m would provide superior containment to a fractured basement rock at 600 m. Similarly, a formation that does not lie close to a mineral resource will provide better isolation than one that does, even if the latter is at the greater depth. The important point is that when deciding on an appropriate facility depth, there are no fixed rules: each case must be addressed individually.

I. HORE-LACY, WorldNuclearAssociation, UK

Abstract: Uranium as an essential energy mineral is produced in similar fashion to many other metal minerals. Mining is underground or in open pits, with the ore dug out and treated in a mill, or increasingly by in situ leaching, with only the dissolved mineral removed and recovered. Some uranium is recovered as a by-product of copper or other mining. Uranium resources are abundant, especially in the light of technological means of using it much more fully. Secondary sources are significant, and much has moved from military stockpiles to power generation. The chapter surveys the main uranium mines and mining regions.

Key words: underground and open pit mining, in situ leaching, tailings management, abundant resources.

6.1 Introduction

In the last sixty years uranium has become one of the world’s most important energy minerals. It is used almost entirely for making electricity, though a small proportion is used for the important task of producing medical isotopes. Some is also used in marine propulsion, especially naval.

Physical properties of thorium as an atomic nucleus

All known isotopes of thorium are unstable but one of them, thorium-232 (Th-232) has a very long half-life of approximately 1.41 x 10 1 0 years (it is an alpha emitter). The half-lives of all other isotopes being less than 100 000 years, Th-232 is the sole thorium isotope of naturally occurring thorium,[12] which has an atomic weight of 232.038 g/mol. It undergoes natural disintegration and is eventually converted through a 10-step chain of isotopes to lead-208, a stable isotope. Alpha and beta particles are emitted during this decay. One intermediate product is the gas radon-220 also called thoron.

Th-232 and U-238 are fertile materials. Just as the absorption of a neutron by U-238 generates Pu-239, so, too, U-233 is generated from Th-232. The reactions are very similar.

232Th П > 233

In a reactor core, Th-232 absorbs a neutron to first produce Th-233, which decays very rapidly (with a radioactive decay period of 22 min) into protactinium-233 (Pa-233), which itself decays (with a radioactive period of 27 days) to produce U-233.

It is also possible, however, for Pa-233 to capture a neutron so that the formation of U-233 is, in effect, in competition with the formation of U-234, the balance depending on the average flux level:

Th-232+n ^Th-233 (22 m) ^Pa-233+ n ^Pa-234 (6.7 h) ^ U-234

For thermal neutrons, U-233 has a higher neutron yield per neutron absorbed than either uranium-235 or plutonium-239. The average number of fission neutrons produced per absorption of a thermal neutron (called the ‘eta’ factor) is typically 2.27 for U-233 in a standard PWR compared to 2.06 for U-235 and 1.84 for Pu-239. This is one of the principal advantages of the thorium cycle: the high eta value of the generated fissile isotope, U-233, makes it the best fissile isotope in the thermal range among all existing fissile isotopes. It is therefore theoretically possible to achieve breeding in today’s reactors using Th/U-233 based fuel.

To generate U-233, fissile materials — such as U-235 or Pu-239 — are required to provide the neutrons that will transform the Th-232 into U-233. After being discharged from the reactor, used fuel can be reprocessed. The fissile materials (U-233, U-235 and Pu) as well as the remaining fertile Th-232 are then retrieved to be recycled into new fuel assemblies.

However, one of the principal drawbacks of the thorium cycle is U-232 production through various nuclear reactions on Th-232 and U-233.[13] U-232 is an alpha emitter with a 72 year half-life and is always associated with U-233 at concentrations ranging from tens to hundreds of parts per million. The U-232 decay chain is as follows:

U-232 (a, 72 yrs) ^ Th-228 (a, 1.9 yrs) ^ Ra-224 (a, 3.6 d / /0.24 MeV) ^ Rn-224 (a, 55 s / /, 0.54MeV) ^ Po-216 (a, 15 s) ^ Pb-212 в-,10.6 h/ /, 0.3 MeV) ^ Bi-212 (a, 60 m / /, 0.78 MeV) ^ Tl-208 (в-, 3 m / /, 2.6 MeV) ^ Pb-208 (stable)

It can be seen that this chain includes hard gamma emitters such as thallium-208 (up to 2.6 MeV). Therefore, the presence of U-232 requires that fabrication of U-233 based fuels be performed remotely in a gamma-shielded environment and this may entail significant additional cost.1

If uranium is chemically purified so that its decay products are removed, freshly separated U-233 (with significant concentrations of U-232) can be processed and converted into desired forms in lightly shielded enclosures without significant radiation exposure to workers. Depending on the U-232 concentration, it will take days or weeks for U-232 decay products that emit gamma rays to build up sufficiently to require heavy shielding to protect the workers.

The nuclear characteristics of U-233 are significantly different from those of weapons grade plutonium (WgPu) or highly enriched uranium (HEU). The minimum critical mass of U-233, in a uniform fluoride aqueous solution, is 0.54 kg (American National Standards Institute [ANSI] 1983). This is somewhat less than that of WgPu or HEU. Thus, facilities designed for WgPu or HEU might not be suitable for storage or processing of U-233 unless more restrictive criticality precautions are instituted.2 It is likely that fabrication of reprocessed U-233/ thorium based fuel would be performed in a dedicated facility whose criticality safety will be designed considering U-233.

Chemical characteristics of uranium-233

Uranium-233 is chemically identical to natural, depleted and enriched uranium. Consequently, the same chemical processes used for natural, depleted and enriched uranium are applicable to U-233. As a consequence of its shorter half-life, however, the U-233 isotope has a higher specific radioactivity than the naturally occurring isotopes of uranium (i. e., U-234, U-235 and U-238). Thus, certain radiation-induced chemical reactions are faster in uranium containing significant quantities of U-233. This is of some importance in situations such as long-term storage. The higher radiation levels of U-233 require that storage containers and U-233 storage forms should not contain either organics (plastics, etc.) or water that, through radiolysis, could degrade to form potentially explosive concentrations (unless they can somehow be vented) of hydrogen gas.

Today, due to product improvements, pellet-clad interactions and corrosion effects are not the dominating failure causes. The main failure mechanism in PWRs and BWRs is now related to debris fretting and grid-to-rod fretting, and the industry has responded by developing improved debris filters and spacer grids (see Section 9.2). However, new failure mechanisms may emerge with increasing burn-up of the fuel. One is outside-in crack growth assisted by hydride formation with incipient cracks, which propagate to the inner surface on a power increase. This mechanism, which is associated with so-called ‘delayed hydride cracking’, has been seen in ramping of high burn-up BWR fuel and was verified experimentally by Sakamoto et al. (2010).

According to an IAEA review of fuel failures in water-cooled reactors (IAEA, 2010) covering the period 1987 to 2006, a continuous decrease of the fuel rod failure rate in PWRs was achieved. The failure rate for VVERs stayed relatively constant in that period, while BWRs and CANDUs mainly improved at the beginning of the 1990s; see Fig. 9.14.

J. W. DAWSON, Consultant based in the UK, and M. PHILLIPS,

Nympsfield Nuclear Ltd, UK

Abstract: The key design features of commercial gas-cooled reactors are presented. The majority of gas-cooled reactors were built in the UK and in France, with the former country progressing their development into the 1980s and beyond whilst France concentrated on water-cooled reactors. The progression from uranium metal fuelled reactors (known as ‘Magnox’ in the UK) to Advanced Gas-Cooled Reactors is discussed, with further discussion of high-temperature gas-cooled reactors and proposed Generation IV plant. The topics discussed include: design of pressure vessels; design of fuel; refuelling; moderator and coolant chemistry; fuel storage and handling; and waste and decommissioning.

Key words: Magnox-type reactors, AGR-type reactors, uranium metal fuel, uranium oxide fuel, carbon dioxide coolant, steel pressure vessel, concrete pressure vessel, high-temperature reactors, generation IV reactors.

11.3 Introduction

Gas-cooled reactors have been used since the earliest days of nuclear power — indeed, it could be argued that Fermi’s first pile in a Chicago squash court was the first gas-cooled reactor, even if the power output was only a few watts.

Gas-cooled reactors were used in the early weapons programmes. The UK Windscale Piles operated in the 1950s until the Pile 1 fire in 1957. These reactors were fuelled with natural uranium clad in aluminium, and were air cooled on a single pass arrangement. This fire was the first major nuclear incident in the western world leading to substantial releases of activity and exposure of the general population.

While much of the rest of the world was pursuing water-cooled reactors for civil power generation (to a large extent on the back of PWR-type reactors developed for underwater propulsion), the UK and France were notable for developing gas-cooled reactors. Both countries developed Magnox-type reactors, France building nine reactors and the UK eventually building 24 reactors.

The name ‘Magnox’ comes from the alloy used for the fuel cladding, in the UK predominantly magnesium with a small component of aluminium, and in France magnesium combined with low levels of zirconium. (Throughout this chapter, the term ‘Magnox reactor’ is used as a generic description of CO2-cooled, uranium metal fuelled and graphite moderated reactors.) In both cases the fuel was natural enrichment uranium metal. The designs built on the technology of early piles used for the production of plutonium, which used aluminium-clad, natural uranium metal fuel. In the UK the first eight Magnox reactors (four each at Chapelcross and Calder Hall) were used for the manufacture of weapons grade plutonium, and in the case of Chapelcross were also used for the manufacture of tritium for military purposes until shut down. They were also used for producing electricity for the civil market. Due to the increased neutron absorption of lithium capsules used in the two Chapelcross tritium-producing reactors, the fuel was low-enriched uranium. Thermodynamic efficiency was not a strong consideration.

The reactor pressure vessel of gas-cooled reactors is enormous by comparison with typical water reactor plant. This results partly from the need to accommodate the massive graphite moderator, as well as from the reduced heat removal capacity of the coolant. The latter in turn leads to the need for larger heat transfer surfaces and/or lower heat fluxes across the clad-coolant boundary.

A few other Magnox-type reactors were built: Latina in Italy, Tokai Mura in Japan and Vandellos 1 in Spain, plus several similar plants of low power constructed in North Korea. However, only in the UK were gas-cooled reactors pursued with enthusiasm. Evolving designs moved from steel pressure vessel reactors to pre­stressed concrete pressure vessels with the construction of the later plants at Oldbury and Wylfa, and the last three French Magnox class reactors of Chinon A3, St Laurent des Eaux and Bugey 1 also used pre-stressed concrete pressure vessels.

The last of the UK Magnox reactors was completed in 1971 and in the UK was superseded by the Advanced Gas-Cooled Reactor (AGR). The reasons for the development of this reactor type were part political (development of a domestic technology) and part technical, in particular the increased thermodynamic efficiency associated with the much higher coolant temperatures of the AGR (exceeding 600 °C). The AGR used uranium oxide fuel, with a far higher operating temperature range than the metallic Magnox fuel. This improved the thermodynamic efficiency to around 42%, compared to around 28% in a Magnox plant (and 32% in a modern PWR).

A total of 14 AGR reactors were built on six sites in the UK (plus a small, prototype reactor). The original design life of these reactors has by the time of writing been exceeded across the fleet by an average of five years.

Other types of gas-cooled reactor have been designed or built. High temperature gas-cooled reactors (HTRs) such as the USA’s Peach Bottom were constructed in the 1960s, and a variety of other reactors such as Germany’s AVR and the UK’s Dragon were built. Many fuel designs exist, typically using small fuel pellets encapsulated in pyrolytic carbon and silicon carbide (TRISO fuel). Proposed future generation IV (gen IV) reactors include very high temperature reactors (VHTR) similar in concept to the HTR using either pebble or prismatic fuel, as well as variations on gas-cooled fast reactor designs.

In the following, the discussion centres on Magnox (Section 12.2) and AGR plant (Section 12.3), and follows with a brief discussion of HTR and future systems (Section 12.5).

302 Nuclear fuel cycle science and engineering