Every Magnox plant had a unique design of refuelling machine, reflecting the development of reactor technology over two decades. However, all had a number of common features.

In the early Windscale Piles, the fuel was inserted into horizontal channels, from which it could be pushed out into a cooling pool. The early French reactors also had horizontal fuel channels (Marcoule G1-G3), but these were followed by more conventional layout plants with vertical channels, as with the UK’s early commercial reactors. The first eight UK Magnox plutonium production reactors were designed and operated only with off-load refuelling but all subsequent reactors were designed for on-load refuelling to maximise time at power, although at the cost of considerable extra complexity both in plant and in safety case.

The large number of fuel channels in a Magnox reactor (several thousand) and the need to keep the number of pressure vessel penetrations to a minimum led to each refuelling standpipe servicing typically ~32 channels (with the potential in some cases to reach more remote channels). The fuelling machines are complex devices, with a number of functional requirements (in particular, for the majority of reactors in which on-load refuelling was undertaken):

• acting as a pressure boundary for the coolant circuit

• opening the standpipe closures

• loading and holding a reserve of new fuel

• having the means to locate any one of several tens of fuel channels from an individual standpipe

• lowering a fuel grab into each channel to pick up a fuel element

• storing used fuel elements in a carousel

• cooling discharged fuel

• shielding discharged fuel

• lowering new fuel into a channel

• discharging used fuel towards the fuel store

In addition, some fuel machines were designed to perform servicing activities on thermocouples, closures or control rods.

Innovative fuel technologies are necessary for almost all the Generation IV systems discussed in this chapter. Fuels for fast reactor systems will probably contain significant quantities of TRUs, which require remote manufacturing operations. The vast majority of the world’s nuclear fuel experience thus far has been obtained with ceramic fuel pellets composed of uranium dioxide or mixed (U, Pu) oxides. Nitride or carbide fuels are more likely to be suitable for Generation IV systems given their high melting point, density and thermal conductivity. They might be supplied as composite fuels in which a fissile component is combined with a higher conductivity inert matrix. Although a certain amount of experimental data are available for these fuels, far more information will be necessary in order to fully assess their suitability. Fuel microstructure plays a decisive role in in-reactor performance and changes in microstructure (e. g. grain size, porosity distribution, etc.) and chemistry can cause marked changes in behaviour, particularly with the elevated temperatures and higher burn-ups associated with these technologies.

Since the beginning of nuclear energy in the 1960s, the strategies for spent fuel management have varied significantly. This variation was mostly caused by proliferation concerns and public opinion of the nuclear industry but variations in uranium prices have also had a part. At the beginning it was believed that spent fuel would be mostly reprocessed and recycled as MOX fuel. The proliferation concerns, primarily, made the US abandon the idea of reprocessing while some other countries continued to pursue it (France, UK, Russia, Japan, China, India and Germany). Under public pressure some countries like Germany later abandoned or slowed down their reprocessing activities. The UK abandoned their reprocessing activities due to technical problems. In the meantime several countries have contracted and built nuclear power plants without any plans to build enrichment and reprocessing capability of their own. The situation in the nuclear technology-developing countries was reflected in the technology-receiving countries as they started considering the disposal of spent fuel. The US strategy with plans for disposal of spent fuel in Yucca Mountain was adopted by many smaller countries and several projects for nuclear fuel disposal were started. Nevertheless, public opposition to the disposal of fuel grew stronger, particularly in Europe, and many projects stalled or slowed down. The at-reactor spent fuel storage capacities were often built with reprocessing in mind and actions had to be taken to enhance the capacity for spent fuel storage. As a consequence, the understanding that spent fuel is a resource has changed into it being considered a liability as high-level radioactive waste. This situation has continued until the end of twentieth century. At the beginning of the second millennium, in the first decade, the recycling of fuel began to attract interest again. It became obvious that many spent fuel disposal projects will be delayed or stalled partly because of a new approach in nuclear projects, which requires engagement of the public in the licensing process. The public acceptance of nuclear energy is, in most democratic countries, often reflected in the politics of the day, which influences the continuity of the licensing process and related research into safety cases related to disposal. Many technology-receiving countries realized that long-term storage of spent fuel is becoming a necessity. Some international projects were started to investigate spent fuel performance during storage,5 which was later extended to long-term storage (50-100 years and there is even mention of a 300 year storage period). France, Russia and countries in Asia (China, India, Japan and Korea) continued their R&D work in advanced reactor technologies and novel fuel cycles in combination with fast reactors. This resulted in the adoption of the ‘wait and see strategy’ or decision postponement in many countries. This strategy involves the long-term storage of spent fuel with continued monitoring of the R&D developments of new nuclear reactors and advanced fuel cycles in technology developing countries. There are several international initiatives that stem from proliferation concerns and at the same time to ensure access to nuclear energy to all countries that have a need for it. Those initiatives (like GIF, INPRO, IFNEC, the Russian Initiative, etc.) are looking at development of generation IV of nuclear fuel cycle facilities as well as possibilities to establish nuclear technology centres for reprocessing fuel in countries that have the required technological basis. As some high-level waste disposal capability will be required regardless of the path chosen, there are also regional initiatives for regional disposal facilities for this type of waste.

In conclusion, spent fuel management strategies have varied from the perception of spent fuel as resource to the perception of it as a liability as radioactive waste. This division of perception continues until 2010 until the increased interest for future nuclear reactors and fuel recycling technology. This scenario involves the long-term storage of fuel in many countries and research effort into long-term fuel storage technologies, fuel integrity and related safety issues. The post Fukushima accident’s impact on spent fuel management strategies remains yet to be seen.

The risks of external exposure of personnel are linked to the treatment in the plant of products having a significant в, у and/or neutron activity. Radiation exposure of operations personnel or the public must be below the regulatory constraint and as low as reasonably achievable (ALARA). This is achieved by:

• provision of radiation shielding

• monitoring of radiation fields both within and around the plant

• monitoring of normal occupational exposure of staff

• limiting radiation exposure of staff by use of additional protective equipment and careful health physics monitoring for specific situations (maintenance or exceptional working conditions)

The accessible parts of the plant are categorized into different classes so that the working conditions may be adapted to the radiological conditions.

488 Nuclear fuel cycle science and engineering

As indicated in Section 17.3, two modes are possible: a) homogeneous recycling of all the TRU elements without further separation or b) heterogeneous recycling of separated MAs (possibly recombined with some TRUs) as targets in specific assemblies, e. g. at the periphery of a fast reactor core. As indicated in 17.3.3, one significant consequence of TRU or MA recycling is an increase (compared to

Pu-only multi-recycling) in spontaneous neutron production after irradiation, which causes difficulties in subsequent fuel fabrication. Homogeneous and heterogeneous recycling should be compared in terms of their respective impact on the fuel cycle.

In 1959, the CETDG recognized the necessity to coordinate with the IAEA in the drafting of any recommendations relating to the transport of radioactive materials for incorporation into the UN Model Regulations. Thus, ECOSOC requested the United Nations Secretary-General to inform the IAEA of ECOSOC’s desire that the IAEA be entrusted with the drafting of recommendations on the transport of radioactive materials, on the understanding that the recommendations would be consistent with the principles adopted by the CETDG and would be formulated in consultation with the United Nations and the relevant specialized agencies. This has led to continuing cooperation between the CETDG, the IAEA, the relevant specialized agencies (particularly ICAO, IMO and the Universal Postal Union) and various other United Nations bodies, including the UN ECE, which oversees the development of two of the mode-specific, regional land-transport regulatory documents.

The IAEA’s founding statute authorizes it to perform certain functions, including in Article III. A.6 ‘to establish or adopt, in consultation and, where appropriate, in collaboration with the competent organs of the United Nations and with the specialized agencies concerned, standards of safety for protection of health and minimization of danger to life and property’. Consequently, the ECOSOC request complemented the IAEA’s statutory functions in the establishment of safety standards (IAEA, 1998).

Following the ECOSOC decision, the IAEA established and first published in 1961 its Regulations for the Safe Transport of Radioactive Materials (identified at that time as Safety Series No. 6), for application to both the national and international carriage of radioactive materials by all modes of transport. Subsequent reviews — conducted by the IAEA’s Secretariat in full consultation with IAEA member states, the relevant specialized agencies and various other United Nations bodies — have resulted in five comprehensively revised versions (published in 1964, 1967, 1973, 1985 and 1996) and several minor revisions. The latest revision was issued in 2009 as IAEA TS-R-1 (IAEA, 2009a). All versions of the Regulations have struck a balance between the need to take account of technical advances, operational experience and the latest radiation protection principles while maintaining a stable framework of regulatory requirements.

In 1964, when approving the first revised version, the IAEA’s Board of Governors authorized the Director General of the IAEA to recommend that the Regulations ‘be taken as a basis for relevant national regulations and be applied to international transport’ (IAEA, 1998). Through the establishment of the international regulatory regime described in Sections 19.2.1 and 19.2.2, the IAEA Regulations (TS-R-1) have now indeed become the basis for national regulations and are applied in international transport.

Worldwide application of the IAEA’s Transport Regulations for all modes of transport has resulted in a high standard of safety, as was recognized in IAEA General Conference Resolution GC(42)/RES/13, which stated that ‘compliance with regulations which take account of the IAEA’s Transport Regulations is providing a high level of safety’ (IAEA, 1998). TS-R-1 is a ‘stand alone’ document that is providing all the requirements for radioactive materials transport safety.

Following PCGE, we take total decommissioning costs for nuclear to be about 15% of the overnight cost. We assume that these are met by a charge against electricity revenue that is paid into a decommissioning fund that earns interest. Whether 15% is likely to be an under — or an over estimate is difficult to predict. On the one hand it seems that the cost of decommissioning old nuclear liabilities often turns out to be more difficult and more expensive than expected.9 On the other, as experience is gathered with decommissioning, changes are being made to the design and operation to reduce ‘back-end charges’, especially through waste avoidance and simpler dismantlement. On balance, it seems that 15% is likely to be an overestimate for future nuclear plant. In any event, its impact is small: using the figures just given, the contribution of nuclear decommissioning to the LCOE is only $0.6 per MWh(e). Decommissioning charges for non-nuclear technologies are likely to be even smaller and they are ignored here.

Operating and maintenance (O&M) costs

PCGE presents O&M costs per MWh(e) for each power plant submitted by the individual countries. The variability in the nuclear O&M costs is relatively high even when restricted to Europe and North America. Overall average values for these countries (rounded to the nearest whole number) were used and are shown in Table 5.2 . This shows that onshore wind has the highest O&M costs. Next highest is nuclear closely followed by coal.

Many mines apply international environmental management standards such as ISO 14001 to their operations. Furthermore, there is now emerging an industry­wide audit framework. This is being developed in collaboration with consumers of uranium, especially utilities, which are sensitive to sustainable development principles. Historically some electric utilities such as Vattenfall and EdF have applied Life Cycle Analysis to include audits of the mines and other fuel cycle facilities supplying them so that they are confident of and can vouch for the standards applying to those activities, both environmentally and socially (especially in relation to indigenous peoples).

The World Nuclear Association (WNA) has developed a framework for internationally standardised reporting on the sustainable development performance of uranium mining and processing sites. This has been agreed to by the main mining companies and developed in close collaboration with utilities so that they are in a position to report to their stakeholders. WNA is working towards implementation of a common audit program to be used worldwide by utilities and mines. There are moves to involve government regulators in this, since it complements their role, and national mining associations. The data supplied by mines will be subject to a verification process.

8.1.4 Non-proliferation

The International Fuel Cycle Evaluation (INFCE) study (1978-1980) summarized thorium fuel activities world wide and considered particular issues related to the technical barriers to proliferation. It was shown that the technical characteristics that would inhibit proliferation for thorium cycles with up to 20% of fissile material were similar to those of uranium-plutonium cycles.

Depending on the design, it takes between 5 and 15 kg of U-233 to make a nuclear weapon, which is not very different from plutonium. Thus, the U-233 bare sphere critical mass is 16 kg, compared to 10 kg for Pu-239 and 48 kg for U-235. Moreover, like U-235, a simple bomb made of U-233 is easier to fabricate than one made of plutonium because there are very few spontaneous neutrons emitted (only 1 neutron/sec/kg). It is therefore possible to design and fabricate a ‘gun — type’ weapon (in which the assembly comes together with the speed of a rifle bullet as opposed to an order-of-magnitude greater speed using high explosive). This is not possible with plutonium, because neutrons emitted by its even mass number isotopes (Pu-238, Pu-240 and Pu-242), always present at some quantity, require the manufacture of a more sophisticated implosion device. In this regard, it must be remembered that ‘civil’ plutonium contains a large proportion of these isotopes, making the manufacture of such a weapon very difficult. Another important feature of U-233 regarding proliferation is that it generates less heat than the even mass number isotopes of plutonium (but more than U-235). This property makes U-233 potentially less troublesome when fabricating a nuclear weapon. In fact, according to some experts and unclassified documents, the USA conducted a test of a U-233 bomb core in 1957 (the ‘Teapot test’) and has since conducted a number of other tests using this isotope.

Nevertheless, a specific technical hurdle does exist in the case of U-233. This is due to the small quantities of U-232 always mixed with U-233 and its associated strong gamma emitters (See section 8.1.2), which create a substantial difficulty in handling purified U-233 during weapon fabrication. In fact, after U-233 containing U-232 is processed, over a few years Th-228 ingrows to a nearly constant level, balanced by its own decay so that the gamma emissions increase and then stabilize. A 10 kg sphere of weapons grade U-233 (with U-232 as low as 5 ppm) could be expected to reach 0.11 mSv/hr at 1 metre after one month, 1.1 mSv/hr after one year, and 2 mSv/hr after two years. Because weapons are usually assembled and disassembled in unshielded glove-boxes, the build-up of the U-232 daughters would quickly create difficulties in complying with limits on the radiation exposure of workers. Terrorist groups, of course, may be less scrupulous about observing such limits.

To some extent, these radiation problems can be overcome by a ‘quick’ processing of U-233 after its separation and/or by the use of appropriate remote handling equipment. Alternatively, it is possible to reduce the concentration of U-232 by taking advantage of the fact that only very energetic neutrons (E > 6 MeV) can bring about the (n,2n) nuclear reaction responsible for its production. According to the French CEA calculations, a concentration of U-232 as low as 5 ppm can be reached if thorium is irradiated in the blanket of a fast reactor where the number of high energy neutrons is relatively low.

The presence of gamma emitters in a U-233 device would be useful to the extent that they provide a radioactive ‘tag’, which can help in the detection and prevention of covert diversion attempts. Once a U-233 weapon is fully assembled, of course, the various neutron-absorbing materials surrounding the fissile core such as neutron reflectors, would reduce the level of external radiation although enough would still penetrate to provide a distinctive signature that can be used to detect and track the weapons from a distance.

Another deterrent to the diversion of U-233 for weapons usage may be obtained by its dilution with U-238. This may be easily performed by mixing thorium with natural or depleted uranium in the fresh fuel (this is the so-called ‘denatured thorium cycle’, mentioned above). However, this option would lead to plutonium production (through U-238) and, therefore, would also raise proliferation concerns (because plutonium can be separated chemically). Another option would be isotopic dilution, mixing U-233 with uranium (natural or depleted) in the course of reprocessing thorium fuel. This option would be ineffective with plutonium because, unlike uranium, all of its isotopes have sufficiently small bare-sphere critical masses to potentially permit their use in nuclear explosives. The drawback with isotopic mixing of U-238 and U-233 is that recycling of the latter would be much less attractive.

Another potential difficulty in using U-233 to make a nuclear weapon results from the high alpha activity of U-232. Indeed because of (alpha, n) nuclear reactions on light element contaminants in the fissile material, neutron emissions would also occur. However, this process produces much fewer neutrons in uranium metal than spontaneous fission of Pu-240 contaminant in plutonium. Furthermore, a high degree of purification would allow the virtual elimination of this potentially disturbing neutron source.

To sum up, U-233 is clearly a material that can be used to make a nuclear weapon but several routes can be implemented to ‘denature’ this material easily enough. Thus, should a uranium-thorium cycle be developed, it would likely offer a degree of proliferation resistance equivalent to that of the LEU cycle, provided that uranium mixed with thorium is not used in conjunction with HEU (enrichment > 90%12).

The core of a BWR is similar to that of a PWR being on a square pitch but the individual fuel assemblies are surrounded by a Zircaloy channel box. Thus the core, hydraulically, consists of a series of parallel channels rather than the open structure of a PWR core. The control assemblies are made up of absorber rods held in a cruciform stainless steel sheath. Each absorber rod consists of a stainless steel tube containing boron carbide absorber pellets. These assemblies (or control blades) are positioned in the spaces between the fuel assemblies as shown in Fig. 10.13. The water gaps between the assemblies also increase the moderation.

The fuel assemblies are more complex than PWR ones and have evolved over time. They are based on a square lattice with pin geometries ranging from 6 x 6 to 10 x 10. Modern 10 x 10 fuel assemblies contain mostly full length fuel rods but also have a small number of part length fuel rods distributed through the bundle. Because the core operates with a two-phase mixture in the upper part of the core, the removal of the upper part of some of the rods increases the moderator to fuel ratio in this region, partly offsetting the reduction due to boiling. It also reduces

1. Top fuel guide

2. Channel fastener

3. Upper tie plate

4. Expansion spring

5. Locking tab

6. Channel

7. Control rod

8. Fuel rod

9. Spacer

10. Core plate assembly

11. Lower tie plate

12. Fuel support piece

13. Fuel pellets

14. End plug

15. Channel spacer

16. Plenum spring

the two-phase pressure drop in the upper bundle, which improves core and channel stability. In addition some of the central fuel rods are replaced with large ‘water rods’ (Zircaloy tubes containing water), which increases the moderation.

Reactivity is controlled in a BWR by means of both control rods and by varying the core flow rate. Because it is a boiling system the use of dissolved absorber is not practicable and so reactivity compensation for burnup effects must be undertaken using either control rods or integral burnable poisons. Since the
control rods enter the core from below they are generally inserted hydraulically but later designs use electro-hydraulic fine motion control rods, which give better control in normal operation as well as increased protection against inadvertent control rod withdrawal or insertion. A series of local power range detectors are distributed throughout the core in positions between the fuel assembly boxes to provide inputs to the power control scheme. The hydraulic systems are such that there is a balance between the hydraulic forces that would insert the rods and those holding them out. Rapid insertion is achieved by venting the pressure holding the rods out.

The recirculation system provides increased flow through the core to allow higher power levels to be achieved but it also provides a means of controlling the power. Increasing the flow reduces the average voidage by sweeping the two — phase mixture more quickly through the core, which increases the moderation and power output. Variations in power of about 25% can be achieved using flow control alone; larger changes will require control rod movement as well.

10.3 Safety features and issues

The fundamental safety functions (IAEA, 2000) required for any reactor are:

• control of reactivity

• core heat removal

• confinement of radioactive material and control of operational discharges as well as limitation of accidental releases

The safety features provided are based on these safety functions. Under accident or incident conditions the reactor must be safely shutdown, decay heat removed from the core and radiation confined. This is traditionally achieved using the principles of defence in depth (IAEA, 1996). This provides multiple administrative and physical barriers to ensure the fundamental objective (IAEA, 2006) of the protection of the public against the effects of ionising radiation is met.

The reactors must be protected against all faults which may be expected to occur. In defining what should be designed against the nuclear industry has always been very conservative in its definition of what should be considered in the design of a reactor and over time this has become even more stringent so that new plants are designed to cope with severe accidents involving multiple failures of systems. Thus the safety features use the principles of redundancy (to increase reliability) and diversity (to provide protection against common mode or common cause failure).

Faults which affect the core arise as a result of a mismatch between core power generation and heat removal. This can be caused by either changes in core power (e. g. a reactivity insertion due to control rod withdrawal) or a change in the heat removal capacity (e. g. failure of a coolant circulation pump). Because LWRs operate at pressure, failures in the pressure boundary will lead to depressurisation and loss of coolant. Emergency core cooling systems (ECCS) are provided to both replace any fluid lost and to provide heat removal from the core under these circumstances.

Confinement of radioactivity is based on the provision of multiple barriers and the defence of these barriers. For LWRs there are four main barriers to radioactive release to the environment. These are:

• the fuel matrix

• the fuel cladding

• the primary circuit

• the reactor containment buildings

The first two barriers are sometimes combined into a single barrier, the fuel rod, but should really be treated separately. In normal operation the vast majority of the radioactive fission products are held within the fuel itself. The fuel operates at relatively low temperatures and so only a small proportion of the volatile fission products are released to the fuel clad gap or the fission gas plenum. For these the fuel clad provides containment but for the majority of the fission products the fuel itself will confine them, provided that it is kept cool. In some faults the cladding may fail (e. g. due to rapid depressurisation following a hypothetical rupture of a major coolant pipe) as a result of the initiating event. Confinement of reactivity by containment building systems will be discussed in 10.9.3.