Open pit and underground mining

Mining methods have been changing. In 1990, 55% of world production came from underground mines, but by 1999 this had shrunk dramatically to 33%.

From 2000 the new Canadian mines increased it again, and with Olympic Dam it is now back to one third (Table 6.1). In situ leach (ISL or ISR) mining has been steadily increasing its share of the total, mainly due to developments in Kazakhstan.

Where orebodies lie close to the surface, they are usually accessed by open cut mining, involving a large pit and the removal of much overburden (overlying rock) as well as a lot of other waste rock. Where orebodies are deeper, underground mining is usually employed, involving construction of access shafts and tunnels but with less waste rock removed and less environmental impact. In either case, grade control is usually achieved by measuring radioactivity as a surrogate for uranium concentration. (The radiometric device detects associated radioactive minerals, which are decay products of the uranium, rather than the uranium itself.)

At Ranger in north Australia, Rossing in Namibia, and most of Canada’s Northern Saskatchewan mines through to McClean Lake, the orebodies have been accessed by open cut mining. Other mines such as Olympic Dam in Australia, McArthur River, Rabbit Lake and Cigar Lake in Northern Saskatchewan, and Akouta in Niger are underground, up to 600 metres deep. At McClean Lake and probably Ranger, mining will be completed underground.

Natural thorium, which, as noted above, has only one isotope, is a relatively abundant element with an average concentration of 7.2 ppm in the earth’s crust. This is significantly higher than uranium (2.5 to 3 ppm), reflecting the longer half-life of Th-232 (1.4 x 101 0 years) compared to 4.5 * 109 years for U-238. Nevertheless, it does not mean at all that the exploitable reserves of thorium are two or three times larger than uranium, as many would assert. In fact, because of its limited uses so far, extensive prospecting of thorium has not yet been conducted so that reliable estimates of the world wide reserves of thorium are not currently available. The famous IAEA ‘red book’ on uranium resources, published periodically, included detailed data on thorium resources until its edition of 1981 but, since then, only global data have been provided. For example, in the last edition published in 2009, a figure of 6.038 tons is given for the total world thorium resource. However, the IAEA has launched recently a small programme specifically intended to estimate thorium resources in the world.

The largest source of thorium is the mineral monazite (phosphate), also a primary source of rare earth elements. It is also found in the mineral thorianite (thorium dioxide) and some has been recovered from igneous veins and igneous carbonate deposits called carbonatites. Significant deposits of thorium are found in Australia, Brazil, Canada, Greenland, India, South Africa and the United States. More generally, the world’s reasonably assured reserves (RAR) are known to be at least as large as those of uranium, and quite probably higher.

In any event, should a closed thorium cycle be deployed on a large industrial scale it must be underscored here that thorium reserves are not a real issue since, like U-238, it is a fertile isotope, that, when deployed with U-233 recycling, would be able to sustain nuclear energy development for a very long time. To provide an explanation of what we mean, let us suppose for example that thorium reserves are only those identified as easily available today U. S. Geological Society ([USGS]) (let us say between 1 and 2 million tons). If one transforms all these reserves into U-233 in nuclear reactors, the complete fission of this uranium-233 would be enough to produce energy equivalent to that produced annually by all the existing nuclear power plants for several thousands of years. Therefore, the problem is not that of the amount of available thorium reserves but that of the quantities of fissile materials necessary to initiate and then sustain a cycle with thorium. Exactly the same may be said of U-238 and the availability of plutonium or U-235.

N. BUTTERY, EDF Energy, UK

Abstract: This chapter discusses the various designs of light water reactors currently in operation and future developments of these, some of which are already under construction. Light water reactors currently account for almost 90% of the world’s installed nuclear capacity. The principal design features of both pressurised water reactors and boiling water reactor are discussed including their safety systems. Operational requirements are outlined and fuel design and performance are discussed. The chapter then outlines future developments in terms of both the development of advanced light water reactors and small and medium reactors.

Key words: light water reactors, boiling water reactors, pressurised water reactors, advanced light water reactors.

10.1 Introduction

According to figures from the International Atomic Energy Agency (IAEA) PRIS database (IAEA, 2011) in January 2012 more than 80% of the world’s nuclear power plants are Light Water Reactors (LWRs). These account for almost 90% of the installed nuclear generation capacity (see Table 10.1). A number of different designs of both Pressurised Water Reactors (PWRs) and Boiling Water Reactors (BWRs) are in operation throughout the world and are manufactured by a number of different companies. The earliest commercial reactors were produced in the USA and many of the modern plants were developed from these as the technology

was transferred to customers outside the US. In other cases indigenous designs developed based on similar principles but with detailed design differences.

In this chapter the main plant features of LWRs will be discussed together with their operational requirements, before outlining the different designs of PWRs and BWRs currently in operation. In Section 10.11 future developments, including plants, which are currently under construction will be discussed.

Fuel design

A diverse range of fuel elements was produced for the Magnox reactors in the UK alone (Fig. 12.2). Magnox reactor fuel channels contained typically eight elements,

although up to 13 were used per channel in Berkeley. Magnox fuel was of natural enrichment, other than in two of the Chapelcross reactors used for tritium production and in Oldbury towards the end of its life due to graphite corrosion having reduced the amount of moderation available.

The cladding alloy of Magnox fuel was principally magnesium with small levels of aluminium (AL80 alloy) or zirconium (ZR55 alloy) in the UK and magnesium/zirconium in French reactors. The Italian and Japanese reactors (Latina and Tokai Mura respectively) used UK-manufactured fuel whereas Vandellos 1 used French fuel. Other fuel element components used a number of different composition alloys. The various Magnox alloys all have low neutron absorption cross sections.

Most Magnox fuel produced in the UK fell into two categories, known as ‘herringbone’ and ‘polyzonal’ (also known as helical).

Polyzonal fuel cladding was ribbed in a helical pattern around the fuel element to increase heat transfer between the uranium and the coolant. The helical pattern resulted in a rotational force being applied to the fuel elements, and so in many designs a spring-loaded arm was added to the top fitting to hold the element stationary in the channel (using friction). The spring was made of a Nimonic alloy containing high levels of stable cobalt: as a result, the Nimonic springs became highly activated with Co-60 in the neutron flux.

Herringbone fuel had ribs running diagonally, with each quadrant of the fuel having ribs running in opposite directions. Thus, there was no net rotational force and no requirement to prevent rotation.

Rib height varied between 7.6 mm and 11.7 mm.

Fuel was located centrally within the channel in a variety of ways:

• Berkeley fuel had bridge pieces located at top and bottom holding a pair of graphite struts running the length of the fuel element.

• Hunterston, Tokai Mura and some Chapelcross fuel was loaded inside a graphite sleeve.

• All fuel was equipped with either ‘splitters’ (four longitudinal Magnox alloy strips held in place by a small number of circumferential braces) or, in the case of herringbone fuel, ‘lugs’ (five or seven longitudinal raised sections on each quadrant of the fuel element, integral with the rest of the cladding). These served to keep the fuel centred in the channel or in the graphite sleeve, as well as providing a means of breaking the flow and increasing heat transfer.

French Magnox fuel was principally of herringbone design.

A variety of end fittings were used to enable fuel elements to be grabbed for refuelling purposes. Fuel elements were individually handled in each channel, i. e. they did not latch onto each other. All fuel elements were fitted with internal ceramic insulating discs at the top and bottom of the uranium rod to protect the cladding from the hot inner regions of the fuel.

The bottom of each fuel channel was fitted with a gag and fuel support unit, incorporating a shock-absorber to protect dropped fuel. The gag was pre-set to restrict flow in each channel, and was not capable of subsequent adjustment. Hunterston A had a separate cast-iron fuel element support member at the bottom of each channel, which was replaced during each refuelling, resulting from the reactors being charged from beneath rather than the conventional layout of pile-cap refuelling.

The low melting point of the metallic fuel and cladding resulted in low irradiation temperatures. The maximum cladding temperature was nominally in the range 400 °C to 470 °C, with maximum fuel temperatures in the range 500 — 600 °C. Maximum thermal ratings varied from approximately 3.5 to 5 MW/te(U).

The mass of Magnox fuel elements was also highly variable according to design, with uranium metal content varying between approximately 5 kg and 12 kg, and gross element weights in the range 7 kg to 20 kg.

The gas-cooled fast reactor (GFR), which has been studied since the 1970s, is a high-temperature fast spectrum reactor capable of using a closed all-FR fuel cycle (NEA, 2006b) (Fig. 13.25) . It combines a more sustainable use of uranium

resources and waste minimization with high efficiency electricity generation. If He is used as a coolant, the outlet temperature reaches around 850 °C, requiring an entirely ceramic core. However, this also means the co-generation of high-quality process heat. As well as being identified by GIF as one of the designs to develop, the GFR is also one of three fast reactor designs selected for development to the demonstration stage within the European Sustainable Nuclear Industry Initiative (ESNII), see Table 13.16. Unlike the SFR design, there is less operational experience with GFR design, which means it will take longer to develop. GIF plans to complete a viability assessment by 2012. GFRs will therefore need to be introduced gradually with the move to a fully fast reactor cycle only probable after the turn of the next century.

Table 13.16 GFR characteristics defined by GIF (Foley and Knight, 2009)

The use of a gas coolant (such as He in GFRs) has several advantages (van Rooijen, 2009):

• chemical compatibility with water, obviating the need for an intermediate coolant loop

• good chemical compatibility with structural materials

• negligible activation of coolant

• since gas coolants are transparent, fuel shuffling operations and inspection are easier

• since gas coolants cannot change phase in the core, the potential of reactivity swings in the case of an accident is reduced

• significant reduction of the void coefficient in comparison with SFR systems

• a harder neutron spectrum, which increases the breeding potential of the reactor

• the potential for a larger coolant fraction in the core without an unacceptable increase in parasitic capture

There are also some disadvantages, resulting from the very poor specific heat capacity of gases in comparison to liquid coolants (Bomboni, 2009):

• the need for artificial roughening of the cladding to maintain acceptable cladding temperature, resulting in an increased pressure drop over the core, and necessitating a higher pumping power

• the need to keep the coolant at high pressure compared to liquid coolants (e. g. 7 MPa is needed for He-based coolant systems)

• the risk of significant vibration of the fuel pins due to the high coolant flow velocity

• difficulty in extracting the decay heat from the high power density core, particularly following a depressurization event (an essential element in passive safety systems identified by GIF)

He is the most promising gas coolant. As with the VHTR design, He allows the potential use of a direct Brayton cycle with high efficiency (around 50%). As a backup option, an indirect cycle using a secondary circuit with supercritical CO2 (25 MPa, 650 °C) could be used (Fig. 13.25b), achieving a cycle with a similar efficiency (van Rooijen, 2009; Hejzlar et al., 2006).

Although no definitive design has yet been agreed, a good example of a GFR system could be the plate-type GCFR 2400 MWft ‘E’ proposed by CEA (Richard et al., 2006). The main design parameters are summarized in Table 13.17. An overview of the core layout is shown in Fig. 13.26 , the geometry of fuel assembly shown in Fig. 13.27 and Fig. 13.28, and the main fuel plate characteristics set out in Table 13.18. The materials composing the core (except for HMs) are Si, C, He and Zr, which minimize parasitic absorptions. A small fraction (1.5% by volume) of the core is composed of a liner, which functions as a sort of catcher for

Table 13.17 Main core parameters of GCFR 2400 ‘E’ (Richard et al., 2006; Girardin et al., 2006)

Thermal power (MWth) 2400

Power density (kW/l) 100

Specific power (W/gHM) 40

Height/diameter ratio 0.63

Theoretical breeding gain 0.0

Fissile height (mm) 2300

No of fuel assemblies 162+120

No of control rods 24

No of reflector assemblies (mixture of Zr3Si2, SiC and He) 168

No of Nominal coolant pressure (MPa) 7.0

Helium inlet temperature (°C) 480

Helium outlet temperature (°C) 850

Maximum clad temperature (°C) 985

Maximum fuel temperature (°C) 1860

Coolant volumetric fraction (%) 30.8

Structural material volumetric fraction (%) 20.8

Helium pressure drop through the core (bar) 1.6

Average coolant speed through the core (m/s) 85

# Inert 1 central

О Fi ssile 162 zone 1 О Fi ssile 120 zone 2

# Control / AU 6 + 12 « AU SAC 6

О Reflector 168 О Neutronic shielding 164

13.26 GCFR 2400 MWth ‘E’ core (Girardin et al., 2006).

the volatile FPs (Girardin et al., 2006). A definitive choice of liner materials still needs to be made.

An important feature is the high height to diameter (H/D) ratio in comparison with typical FR values (‘pancake cores’) and the ‘zero’ breeding gain. A higher H/D ratio reduces leakages and improves the neutron economy (Bomboni et al.,

2008a). This allows for very high irradiation levels, a relatively small fissile inventory, large flexibility in the choice of fuel composition and the option of inserting dedicated targets for transmutation without significant reduction of core performance. Since the GFR design aims at a ‘self-sustainable’ cycle, i. e. a

production of fissile material that is equal to its consumption, an optimal H/D has to be established, which is sufficiently high to sustain the cycle without external addition of fissile material and/or the presence of fertile blankets but, at the same time, is not too high for thermal-fluid-dynamics reasons (van Rooijen, 2009). A breeding blanket of depleted uranium (DU) is not envisaged, because it could pose proliferation risks.

There are several cladding properties that may become significant in the management of spent fuel. Typical changes are rod growth and clad hardening. During reactor operation, the cladding undergoes corrosion in water resulting in the formation of hydrogen. Some of this hydrogen is taken up by the cladding and contributes to the reduction of cladding ductility.

The mechanisms of corrosion in PWRs and BWRs are somewhat different but the results may be similar. The corrosion of cladding can lead to an increased driving stress intensity factor (due to changes in the local chemical potential and local geometry) for cladding breach due to delayed hydride cracking (DHC) (ref 1410). Hydriding can affect ductility and can lead to a cladding breach during an accident. Another feature of the cladding that affects spent fuel management is crud deposition on the rod surface. Crud can be found on the cladding in very different amounts dependent on the individual reactor but it also seems to be dependent on the fuel burnup. The experience shows that for some unknown reason crud is not a problem in Russian WWER reactors or in some PWRs in Europe.

Crud is the only source of radioactive particulates that can be released without a fuel cladding breach and it presents a dose risk if the fuel storage containment is breached.

Within the statutory framework that regulates the nuclear industry, there is an overriding requirement to demonstrate, through an adequate and appropriate safety case, that all hazards associated with operations on a licensed site are understood, effectively managed and controlled. The risks directly linked to the presence of radioactive material are contamination, irradiation, and criticality, and measures to limit or eliminate these are discussed below.

Containment is provided through several barriers between the radioactive material and the environment, dynamic tightness by ventilation and specific design features. Irradiation is controlled by using proper shielding (concrete, lead and lead glass). Criticality requires the use of safe geometry and redundant control.

External events and their consequences have to be taken into account: earthquakes, loss of utilities, fire and explosion.

Operations must be reliable. This is achieved through the use of maintenance — free equipment whenever possible, the duplication of units or equipment subject to failure, strict quality control, the use of specific devices for remote maintenance and a layout per design allowing direct maintenance after decontamination of equipment.

Instrumentation and control are, as well, key in a reprocessing plant. A recycling plant is equipped with several control systems to ensure the monitoring of processes and also to protect workers and the environment. This includes sampling benches for indirect sampling of an active apparatus, standard sensors (temperature, pressure, level, . . .) and nuclear sensors (activity a, yp, neutrons).

16.4.1 Containment

Contamination is produced when radioactive material escapes into the workplace or the environment. Control of contamination relies on the installation of containment systems, which generally include:

• A first static barrier that is an integral part of the equipment being used or some other envelope that is in direct contact with the radioactive material. This first barrier provides protection for the environment and the operations personnel and is designed to provide containment that is as total as possible.

• A second static barrier whose role is to limit dispersion of the radioactive material in case of a leak or breach of the first barrier. This barrier may consist of the walls of the cells or the containment building.

These static barriers are complemented by a dynamic containment system using forced draft ventilation:

• A ventilation system connected to the first static barrier keeps a depression (a negative pressure difference) between the process equipment and the cells or containment rooms.

• A ventilation system connected to the second static barrier keeps a depression between the secondary containment system and the surrounding environment (which may be a building or the open air).

These ventilation systems are designed to ensure that any flow of gas (e. g. air) is towards the areas where the level of radioactivity is highest (Fig. 16.23).

Containment system

• Ventilation

system г—Й-СТ-

Depending upon the level of radioactivity in the cells or containment areas to which they are connected, ventilation systems are characterized by the numbers, the nature and the arrangement of filters mounted on the blowing or extraction circuits.

The secondary containment system may include areas where the operations personnel have to work. It is a complementary barrier for the environment in case of failure of the first system. The areas forming the second containment are adjacent to the cells or to the containment areas; they are at a level of pressure intermediate between the cells and the external environment.

As discussed in Section 17.3, the choice of the appropriate core conversion ratio is related to the chosen strategy, i. e. a ‘high’ conversion ratio (CR>1) corresponds to a strategy of sustainable nuclear energy with waste minimization in a repository and stabilization of the MA inventory contained within the fuel cycle. A ‘low’ conversion ratio (CR -0.4-0.6) corresponds to a strategy of rapid TRU or MA inventory reduction (see Figs 17.5 and 17.6). ADSs correspond to cores with CR = 0 if an inert matrix fuel (IMF) is used as a driver.

17.1.1 Reactor type

The ADS in the schemes of Figs 17.5 and 17.6 can be considered as a ‘dedicated’ transmuter reactor. The ‘critical FR’ in Fig. 17.9 could be, for example, a FR that is progressively loaded with MA-containing fuel, with sodium as the coolant. The choice of the CR for the critical fast reactor determines whether the FR is to be

means ‘option’

Where

77.9 Scheme showing the different P&T options within the seven technical domains (on the left hand side) for the three main fuel cycle scenarios of Section 17.4 (across the top of the figure).

regarded as part of a sustainable system, as shown in Fig. 17.4, or as a transmuter reactor and an alternative to ADS (see Section 17.3). In fact, this is true regardless of the choice of coolant, so that gas or a liquid lead alloy could be substituted for sodium with no impact on the P&T implementation strategy, at least in an initial phase. No firm choices have been made at the present time and three specific prototypes are envisaged in Europe with missions related to transmutation:

In France, a Generation IV and actinide incineration demonstration prototype should be commissioned in the 2020 decade. This prototype, called ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration55 will probably replace the Phenix fast reactor and provide a means of testing homogeneous and heterogeneous minor actinide recycling modes, as described in Section 17.3.

The Myrrha project 56 is initially a 57 MW ADS with a sub-critical fast reactor core. The coolant is a liquid lead-bismuth alloy (Pb-Bi). The spallation target is also Pb-Bi. A schematic illustration is shown in Fig. 17.10.

An experimental gas-cooled fast neutron reactor ALLEGRO57 is also planned with a 2020 target date for the start of operation. Detailed design studies should start in 2013.

To facilitate the integration of the IAEA’s TS-R-1 regulations into the international standards, the IAEA and members of its Transport Safety Standards Committee (TRANSSC) work closely with the CETDG to incorporate the TS-R-1 safety requirements into the UN Model Regulations. As a result, the UN Model Regulations now provide a complete set of UN-recommended requirements for all classes of dangerous goods. This approach results in having the TS-R-1 requirements serve as the universal basis for worldwide radioactive materials transport safety.

In addition, the CETDG, IAEA, ICAO, IMO and UN ECE have worked closely to develop an efficient approach to keeping the dangerous goods regulations up to date and closely coordinated. A two-year revision cycle is used by the CETDG to keep the Model Regulations current and avoid a backlog of issues. Similarly, the ICAO, IMO, UN ECE (secretariat for ADR and ADN) and the secretariat for RID all follow two-year revision cycles that commence with the completion of each CETDG revision cycle.

This closely coordinated set of revision cycles by the various international organizations defines the external environment with which the IAEA Transport Regulations need to effectively integrate. In order to facilitate harmonization, the IAEA works to provide two-year inputs to the CETDG revision cycle. Initially, the IAEA Regulations were updated in detail on an approximate ten-year cycle. With a view to global harmonization, it was noted that moving to a two-year cycle did not require that the IAEA Transport Regulations be revised every two years and it was agreed it could be considered a review cycle rather than a revision cycle. This approach allows for the frequent publication of as amended versions (containing no comprehensive changes), issued when needed; as well as revised versions (containing comprehensive changes), issued when needed. It allows two-yearly input to the UN Model Regulations. Consequently, the two-year review cycle was adopted by the IAEA. Thus, the radioactive material regulations are kept up to date and as consistent as possible with the UN Model Regulations.

The capital cost contribution to the LCOE (CAP in Eq. 5.3) is calculated from the overnight cost, the financing costs (represented in Eq. 5.3 by the parameter I), the capital recovery factor (represented in Eq. 5.3 by the parameter R) and the plant availability.

PCGE8 provides overnight cost data (INVEST in Eq. 5.2) that include a contingency of 15% (usually) for nuclear and 5% for the other technologies. The data show considerable variability — it is not unusual for overnight costs to vary by a factor of two or three between the highest and lowest cost plant. Very low costs, for all technologies, are quoted by China and South Korea. Restricting the data to OECD countries in Europe and North America produces higher values that are also more homogeneous. Table 5.1 presents mean values for this reduced dataset. The highest overnight costs are associated with coal generation when this includes carbon capture (coal+CC) and nuclear. The lowest values occur for CCGT, which is more than four times lower than both coal+CC and nuclear.

Table 5.1 shows that the cost of finance is greatest for nuclear (35%) and least for wind (7.5%). This is a direct consequence of the length of the respective construction times.

PCGE uses an availability figure of 85% for nuclear, coal and gas. This is reasonable for nuclear but is rather greater than is found in practice for gas and (to a lesser extent) coal where owners may have a strategy of avoiding baseload operation so as to take advantage of higher prices. The use of a higher availability is an attempt to offset this: in effect it represents an actual maximisation of operating profit by a hypothetical maximisation of generation. As a first approximation, we adopt it here (Table 5.1). There is further discussion of this in Section 5.2.7. PCGE provides availability figures for onshore wind that range between 22 and 41% with a mean value of 27%, which is the figure used here.

Based on Eq. 5.3, Table 5.1 calculates the capital cost component of the LCOE. The highest value recorded is the one for onshore wind despite the fact that the overnight cost is relatively modest. This is a consequence of its low availability. The lowest capital costs are found for gas and coal while nuclear and coal+CC are similar.