Current enrichment technologies are based on either gaseous diffusion or centrifuge methods. Other methods based on curved nozzle separation and laser enrichment have also been explored (Leclercq, 1986). The capital costs of enrichment plants are relatively high; around 6% of the total generation cost (Bertel and Wilmer, 2003).

Gaseous diffusion is a widely used method of enrichment in many countries. In this process, uranium hexafluoride is enriched by diffusion through porous barriers. The process is repeated through a large number of stages until the required enrichment is obtained. Capital costs are therefore high. The engineering issues relate to constructing corrosion resistant and efficient barriers to prevent blocking of the pores. For the process to work efficiently, feed streams need to be compressed and then the heat of compression removed to maintain the gases at the correct temperature and pressure. This is an energy­intensive process and therefore gaseous diffusion plants have high operating costs because of their large electricity requirements.

In centrifuge plants the uranium hexafluoride gas is spun in a vertical centrifuge and the U-235 concentrated near the axis. The high rotational speeds place limits on centrifuge capacity and therefore many (thousands) of the identical centrifuges are required. Centrifuge plants are, therefore, also capital intensive. However, operating costs are less than for gaseous diffusion. The gas must be successively centrifuged in stages but the number of stages is approximately 10 times less. Also the operating energy requirements are 10 times less than for gaseous diffusion plants.

At the present time, there is a surplus of enrichment capacity in operation in the world, despite the fact that few companies operate enrichment plants and the number of plants is very small (Bertel and Wilmer, 2003). However some plants have been in operation for 25 years and will need to be replaced. The next generation of enrichment plants is likely to be based on centrifuge technology.

7.5.1 Availability

Improved performance of current plants has been discussed earlier in this book. This is being achieved by better ways of processing information on the plant condition, e. g. components, better surveillance and diagnostics. The causes of reduced level of performance can be determined by analysing the better data obtained and improved management techniques can be implemented. Clearly these types of practices equally apply to advanced as for current generation plants.

Potential improvement in performance of evolutionary plants can be established in the design phase as indicated in Table 7.1. It may also be possible to take advantage of specific improved technology, e. g. the use of high burn-up fuel to enable longer length of cycles, more advanced computer-based systems, and simpler hydrogen control systems, which require less testing during outages and thereby reduce outage time.

Other technological improvements, some of which have already been tested on current plants, concern the utilisation of better materials. For example, Inconel 690 has better corrosion resistance compared with Inconel 600 in a steam generator environment. This improved material can be used for SG replacement in current plant as well as being used for new advanced plants.

Another way, which will reduce operating costs, is to reduce the number of welds, using better forging techniques. This reduces the need for weld inspection in areas of high — radiation fluence.

Future designs should achieve improved energy availability; targets of 87% for average energy availability factor have been put forward (Juhn, 1999) for future plants. Values of high 70s% are being achieved on current plant. These figures for advanced plants can be achieved by incorporating, at the design stage, the experience gained from currently operating plant.

7.5.2 Man-Machine Interface

Over the past few decades there has been very considerable progress in instrumentation and control (I&C) including the man-machine interface (Wahlstrom et al., 1999) (Table 7.2). New digital instrumentation has been developed; bringing both benefits and some difficulties. This new technology has been rapidly assimilated into conventional industry but has been incorporated to a lesser extent into the nuclear industry. The partial reason for this has been a significant downturn in the building of new plants in the last two decades of the 20th century. Other reasons are the lack of drive to replace proven old systems by new systems and in a similar vein, the conservatism of the nuclear industry and its regulators.

Nevertheless new technologies have been implemented in modernisation projects and good experience has been obtained. For new reactors, the new technology will be incorporated at the design stage. It will cover instrumentation, cabling, signal conditioning, many aspects of control, process computers and all aspects of an efficient man-machine interface. Developments relate to hardware, software, the development of information networks, interfacing and back-fitting with older systems (in the case of existing plants) and management of these aspects.

As noted above, I&C systems bring both benefits and some disadvantages. Digital systems are more flexible than analogue systems, which are limited in both practical and financial constraints. Storage capacity is not limited by physical constraints, ease of duplication of signals, better functionality of the control room, better reliability, etc. Other beneficial features are that new functions can easily be included; computers can be embedded into different components. Nevertheless digital systems are more unpredictable than analogue systems, because the software may be complex. A disadvantage of digital systems is their lack of robustness to different environmental factors such as temperature, moisture and radiation. However, commercial off-the-shelf systems can be designed to apply to the nuclear as well as the non-nuclear sector. This ensures better validation for application in some of the more challenging environments existing in nuclear plant.

Modernisation projects have been in progress in various countries — Finland, Germany, Netherlands and Sweden. Different strategies for establishing a mix between new and

existing I&C systems have been developed. In Korea, for example, upgrades of the Korea Standard Nuclear Plant (KSNP) are proposed which will be implemented into the new Ulchin Units 5&6 under construction.

The I&C systems for new plant designs clearly build on the experience gained from modernisation projects on current plants. However, for new reactor designs, a more generic approach to I&C systems is being adopted. The approaches being put forward for evolutionary plant though, do not vary substantially from the more developed systems already in place on the newer present generation plants. In both cases, I&C systems are based on digital distributed systems. Control room layouts follow the approach of compactness with information displayed on visual display units (VDUs). The main future developments are likely to be simplifications in regard to redundancy and physical independence; these have been put forward in some of the more innovative designs of the future.

Differences across the reactor vendors are relatively small. The KNSR design (a typical design) implements the utility requirements of the EPRI URD, including three redundant consoles, a separated console, large display panels and additional monitoring consoles. This concept relies on the 2/4 redundancy principle. The man-made interface incorporates computerised operating procedures and the I&C design is a plant-wide digital system. The plant protection and safety control system are four-channel programmable logic controller-based systems. Non-safety controls are implemented in a two-channel system with diverse processors; similarly plant monitoring has two independent diverse systems (Wahlstrom et al., 1999).

Germany has 13 pressurised water reactors and 6 boiling water reactors currently in operation, contributing to about one-third of the country’s electricity generation (World Nuclear Association, 2003). Following the reunification of Germany in 1990, all the Russian-designed VVER plants in the East were shutdown for safety reasons.

Following the formation of a coalition government in 1998 between the Social Democratic Party and the Green Party, it was agreed by both parties to introduce legislation to eventually phase out nuclear power. However, a consensus was agreed between the utilities and the government in mid-2000, which would allow the continued operation of the nuclear plants for some years ahead. There was also a government commitment to allow present reprocessing practices and waste disposal operations to continue. In particular, this allowed for reprocessing in France and the UK and the maintenance of two repository projects in Germany.

In mid-2001, an agreement was eventually signed between the energy companies and the coalition government that limited the operational lives of the reactors to an average of 32 years. In practice, some of the less economic plants are likely to be shutdown sooner. The construction of any new nuclear power plants however remains prohibited at present. An additional principle in the agreement is the storage of fuel on-site.

There is some evidence that German public opinion has moved more towards supporting nuclear energy. It remains unclear whether the country’s goals for greenhouse emissions can be achieved without nuclear energy. There is still strong support for Franco — German co-operation in some areas, e. g. in the development of the EPR and in securing the improved safety of Russian-designed reactors via technology transfer.

There are designs proposed that are radically different from current generation technology and these would require substantial development and investment before building and licensing. Examples are given in Table 11.4.

Integral type pressurised water reactors such as PIUS, VPBER-600, SPWR and ISIS (PIUS, 1997; VPBER-600, 1997; SPWR, 1997; ISIS, 1997) are completely immersed in a large pool. Many of these concepts have been described as inherently safe, i. e. they depend entirely on the forces of gravity and natural circulation for operation. Typical design objectives are that they should be ‘operator forgiving’ and should incorporate simple safety principles (which should therefore imply increased reliability). For flexibility of supply and operation they should be available in small — or moderate-size units, which could be coupled if necessary. Such designs and other revolutionary approaches are considered in this chapter.

The PIUS reactor (PIUS, 1997) is immersed in a large pool where the boron concentration is controlled by several ‘density lock’ arrangements (Figure 11.4). There are no control rods and the required reactivity is maintained by control of the boron concentration and moderator temperature. In the event of an accident, a natural circulation loop through the core is established, resulting in reactor shutdown and core cooling.

The VPBER-600 (VPBER-600, 1997) is an integral PWR, located in a guard vessel. The design basis was taken from the AST-500 heating reactor, which was designed in the early 1980s. VPBER-600 includes passive safety systems and diverse operation principles with significant redundancy and self-actuation.

The SPWR (SPWR, 1997) is based on an integral design with the complete primary circuit including the core, MCPs, pressuriser and the SG encompassed within the reactor pressure vessel. It employs passive systems for shutdown and decay heat removal under normal operation and also accident mitigation. Highly borated water is used for shutdown in place of control rods.

ISIS (ISIS, 1997) is also an integral PWR reactor, which is completely immersed in cold borated water. It is similar to the PIUS concept except that the reactor components are derived from proven technology.

Unlike the integral designs described above, the JPSR is a passive two-loop PWR design (JPSR, 1997), adopting a boron-free concept to increase reactivity sensitivity to changes in moderator density. As a result, reactor power can be controlled by adjusting the steam generator feedwater flow rate. This simplification in design results in a reduction in manpower for operation and maintenance.

Neutron transport in the fission range of heavy metal energies has been studied for many years within the nuclear reactor industry. In ADS, the situation is more complicated than in conventional nuclear reactors. In this case, there is dual transport modelling required, the transport of medium energy charged particles in the energy range 1-3 GeV in the spallation target, and the transport of neutrons down to low — energy range.

A two-step process of spallation and evaporation of the residual nucleus occurs when medium-energy protons collide with a nucleus. If the residual nucleus has high mass and moderately high excitation energy, it might undergo fission in competition with the evaporation reaction.

In regard to presently developed methodologies, the nuclear cascade processes can be calculated by the NMTC (Coleman and Armstrong, 1970) and HETC (Radiation Shielding Information Centre, 1977) codes using two-body collision theory, which is valid until a particle slows down and its wavelength becomes longer than the average distance between the nuclei. In this regime, an optical potential model can be used, based on quantum mechanics. These codes have been developed to calculate high-energy fission, for targets with high atomic number such as uranium and the actinides, by various laboratories including JAERI (NMTC) (Nakahara and Tsutsui, 1982), BNL (NMTC) (Takahashi, 1984), LANL (LAHET) (Prael and Lichtenatein, 1989). Other nuclear cascade codes FLUKA (Ranft et al., 1985) and CASIM (VanGinnekin et al., 1971) have been developed by the international community.

Two areas of microscopic nuclear physics have been studied by OECD/NEA, using data from a thin target benchmark and transport modelling using thick target physics (IAEA-TECDOC-985, 1997a).

The applications discussed previously in this section relate mainly to low-temperature applications. There are a number of interesting medium and higher temperature process

heat-related applications associated with oil refining and liquid fuel production from coal and hydrogen production. A particular interest is in hydrogen production for future generation reactors; this has already been discussed in Chapter 12.

Nuclear heat applications at medium and high temperature have not yet been developed at industrial scale. They are, however, being researched at smaller laboratory scale. The most promising systems are the high-temperature gas cooled reactors. The present near­term designs (GT-MHR, PBMR) are evolutions from smaller prototype reactors that operated in the UK (Dragon), Germany (AVR & THTR-300), and the US (Peach Bottom and Fort St Vrain).

Some particular reactor concepts that are being considered in the development of future high — and medium-temperature applications are summarised in Table 14.8. These relate to on-going programmes in China, Japan and Russia. These reactor systems are described briefly below.

15.12.1 Technical Issues

A feature of many of the innovative future designs is their relatively higher outlet temperatures compared with current generation plant. The safety envelopes of many of the component materials may not extend to these temperatures in which case new materials will need to be developed and qualified. Additionally many of the coolants may erode or corrode the surrounding materials, particularly in the high-temperature environments that are exhibited (IEA/OECD (NEA)/IAEA, 2002). There will also need to be materials developments in process systems associated with the applications of innovative reactors, e. g. hydrogen generation.

The high-temperature gas reactors may have outlet temperatures that could be as high as 1500°C. These will require significant advances in high-temperature materials, alloys, ceramics and composite materials. Future water reactors, including supercritical systems, liquid lead and molten salt systems, will also require substantial material developments to withstand both corrosive and high-temperature environments (The US Generation IV Implementation Strategy, 2003).

Anticipated areas of research could include the performance of various material compositions in these environments, the development of protective coatings and research into particular materials for specific applications.

2.1. INTRODUCTION/OBJECTIVES

There are approximately 440 nuclear reactors in operation in about 30 countries worldwide. For the continuation of nuclear power, the most important requirement is the safe and efficient operation of these reactors. This chapter summarises the principal issues associated with the operation of current generation nuclear power plant. These relate to the incentives for continued nuclear generation (including its benefits as a carbon free generator), international policy, economics, safety, extension of plant life and public safety concerns. These issues are covered in more detail in separate succeeding chapters.

At the time of writing, the main focus of the nuclear industry in most countries is the continued operation of existing plant rather than on the building of new plant. This is particularly true in Europe and the US. However, some building is continuing in the Asian nuclear power states. The anticipated nuclear generating capacity at least until 2010 is expected to be comprised mainly of generation from plants in operation today (Chamberlain, 1997).

The main criteria for continued plant operation are that the plants must remain safe to the satisfaction of the regulators but also economically viable to meet the requirements of the utilities and the stakeholders. Other pre-conditions that are likely to apply to continued civil nuclear power generation in general, including new build, are separation from weapons programmes, openness and good communication of the issues and effective waste management.

Within present day generation plants, there is a continuing requirement to improve safety and performance. There has been a major investment in experimental and theoretical programmes of work to support this objective. Component research has been carried out to back-fit safety systems on older reactors. Better understanding has led to the development of additional accident prevention and mitigation guidelines. It is also leading to proposals for new systems, e. g. to reduce releases in severe accidents.

There has been a decade of safety upgrades and improvements in nuclear power plants in the EU Accession countries. The main design safety issues are associated with re­licensing, plant life extension, ageing and periodic safety reviews. Another issue is the completion of nuclear power plants that have been left partially built for a number of years (IAEA/NSR/2002, 2003).

6.1. INTRODUCTION/OBJECTIVES

There are still major issues associated with the disposal of nuclear waste. There are bodies of opinion within the nuclear industry, regulators and many experts that believe solutions exist for all stages, but there is considerable public mistrust. This is fuelled since in many countries, there is no position on the final disposal strategy for long lived high-level waste, i. e. only temporary solutions are in place and there is no long-term policy. However, forward progress is happening in the US, Finland and Sweden where repositories are now being considered. Good progress has already been made towards the incarceration of low and intermediate waste in final long-term repositories.

The nuclear industry in common with all other industries has facilities that eventually come to the end of their productive life. Decommissioning of these facilities is then required, which involves the safe disposal of various hazardous materials. Such activities are carried out as a normal practice in an on-going nuclear energy programme and much experience has already been gained from a programme that has already been in operation for over 50 years. However, many present day reactors built in the 60s and 70s are now approaching the end of their design life and therefore decommissioning activities will increase over the next few years. This chapter also considers the key issues of decommissioning, including a review of different options that are being adopted and the impact on costs.