In February 2003, the UK government published an Energy White Paper (Energy White Paper, 2003) to define an energy policy looking forward from today to 2020 and beyond as far as 2050. Many of the policies set out in the paper took as their starting point the Energy Review published by the Cabinet Office’s Performance and Innovation Unit (now the Strategy Unit) (The Energy Review, 2002) in February 2002 and the White Paper was produced after in-depth analysis of the various options. The review covered all forms of energy requirement, from heating and lighting to transport, industry and communications.

Regarding nuclear power for either electrical or non-electrical generation, a key safety issue concerns the management of nuclear waste. Supporters of nuclear energy argue that the technical problems associated with waste disposal are solved; opponents do not agree. There are other commercial and practical issues such as: capital cost, market price of nuclear electricity and energy, and the risks, including liabilities and availability of an adequate skill base. All these will impact any decision for new build.

By 2020, the existing fleets of UK nuclear power stations will all have almost reached the end of their working lives. The White Paper acknowledged that nuclear power was currently an important source of carbon-free electricity and remains an option for the future. However, it did not propose new build and stated that before any decision to proceed with the building of a new power station, there would need to be the fullest consultation and publication of a White Paper setting down Government’s proposals. The arguments for a delay were both on economic grounds and concerning the issue of waste disposal. These considerations are clearly relevant to all nuclear energy products (electrical and non-electrical) in general.

Nuclear power in the UK has in the past been used largely for electricity generation, but some reactor designs are suitable for either co-generation of heat or even dedicated nuclear heating applications. For example UK industry is showing a revived interest in high temperature reactors (HTRs). The UK is keeping abreast of a number of international initiatives, via participation in the Generation IV programme led by USDOE.

For many years, fast reactors have offered the attraction of a sustainable fuel supply based on a uranium-plutonium fuel cycle. There is now a current interest in exploring particular advantages of the fast reactor to consume plutonium, and reduce the stockpile of weapons fuel. Also the fast reactor can be used to irradiate minor actinides and fission products to reduce the toxicity of long-term wastes. Within this framework, the gas-cooled fast reactor (GCFR) has a number of potential advantages to offer. The UK is participating in EC initiatives in this area; e. g. an ongoing review of gas-cooled reactor concepts (Mitchell et al., 2001) within the 5th Framework programme.

The UK is also participating in the EC CAPRA (Consummation accrue de plutonium dans les reacteurs Rapides) project, which aims to utilise existing plutonium stocks arising from the operation of commercial thermal reactors (IAEA-TECDOC-1083, 1999).

Work is currently underway in the UK in the EC CAPRA/CADRA project to evaluate the potential for the transmutation of plutonium and minor actinides in a wide variety of reactor concepts including a GCFR or a HTR system (Smith et al., 2003). Participation in these various gas reactor programmes takes advantage of the UK long-standing experience of gas reactor technology.

The UK is also keeping abreast of other initiatives, including the application of proton particle accelerators in connection with sub-critical reactor systems.

The UK participates in fusion research and collaborative international programmes. During the 1990s, the Joint European Torus (JET) project has made progress in generating significant amounts of energy. For the next generation of Tokamaks, interested nations including the UK will participate in the International Tokamak experimental reactor (ITER) project. This technology is not likely to be available as a viable power generator until beyond 2030.

Gas-cooled reactors have been studied in various countries since the start of the nuclear power programme (Methnani, 2003; Mitchell et al., 2002). Future generation plants will benefit from this experience. In this section, attention is focussed on the high temperature thermal systems, in the following section, fast spectrum systems will be considered. The early gas reactors were natural uranium fuelled, graphite moderated and air cooled and used for military operations. Following on, in the UK, Magnox plants incorporated pressurised carbon dioxide cooling followed by advanced gas reactors with enriched uranium oxide fuel and higher pressure carbon dioxide as coolant.

High temperature gas-cooled reactor (HTGR) concepts have been studied in parallel with the carbon dioxide-cooled plants. Early experimental and prototype reactors included Dragon, AVR and Peach Bottom. The Dragon reactor operated at Winfrith and incorporated helium cooling and ceramic-coated particle fuel. This reactor included highly enriched uranium-thorium carbide fuel particles. The coolant operating outlet temperature was 750°C and much useful information on helium-based HTGR systems arose from the early Dragon programme. The AVR system operated in Julich in Germany. It had a higher temperature of 950°C and used 100,000 coated fuel spheres. This was the concept that is currently being considered for the Pebble Bed Modular Reactor (PBMR) design. In this design, the fuel spheres move downwards in the reactor core within a graphite reflector vessel. The first HTGR in the US was Peach Bottom Unit 1, rated at 40 MWe. Several fuel designs have been developed to overcome problems with cracked fuel.

Two main types of HTGR designs have emerged over the last 2 decades, through the operation of several prototypes. The German thorium high-temperature reactor (THTR — 300) was of a pebble bed type; The US Fort St. Vrain design was of the prismatic design.

Power ratings were raised to 300 MWe and there were various design features including a pre-stressed concrete reactor pressure vessel and a more advanced coated fuel particle design known as TRISO.

More recent designs have incorporated reduced power density, reduced overall power and more passive systems. The general atomics modular high temperature gas reactor (MHTGR) was rated at 350-450 MWt and the German HTR series design was rated at 200-300 MWt. These system designs were more modular. The direct cycle MHTGR design, utilising advanced gas turbine and high temperature turbine technology, could yield efficiencies up to 50%.

The IAEA has co-ordinated several safety-related research projects on the physics, heat removal aspects and fuel and fission product behaviour of HTGRs. A latest activity is concerned with benchmarking core physics and thermal-hydraulic methods against experimental data in order to evaluate HTGR performance.

The European Commission has recently supported a network R&D activity to address the major design issues associated with the core physics and fuel cycle, and the material and components issues. The project is also concerned with the safety and licensing issues associated with the HTGR design.

In respect of their reactor physics, HTGRs have a relatively low power density compared with light water reactors, of the order of 2-3MWm_3. They include a large volume of graphite as moderator that also implies a relatively large core size. The core is usually annular to give a flat radial power distribution. HTGRs typically include a central graphite reflector and radial and axial reflectors, and are designed such that the inner reflectors that absorb a large fluence are replaceable. HTGRs exhibit good neutron economy due to the low absorption of the graphite and negligible absorption by the helium coolant. Another desirable feature is a negative reactivity core temperature coefficient that increases in magnitude at higher burn-up and lower fuel enrichment.

In current PBMR designs, the control rods for both operation and safety purposes are situated outside the reflector region in order to limit exposure at high temperature. This means that they have reduced worth, which tends to imply smaller diameter annular cores are designed. The fuel inventory is relatively low due to the use of low enriched fuel, which means that safety is not compromised. The power can also be effectively managed by varying the helium inventory and taking advantage of the negative temperature coefficient in the 25-100% power range.

HTGR core physics tools have been validated by comparison with the HTR-10 reactor in China, the high temperature test reactor (HTTR) reactor in Japan and the Proteus critical facility in Switzerland. Reactor physics methods have been applied utilising methods ranging from detailed Monte Carlo methods to combinations of cell transport and core diffusion models. Benchmarks have shown that some of these codes predicted the core criticality loading to a good level of accuracy. Thus, there are adequate methods available for reactor physics calculations for low-enriched gas-cooled reactors.

Regarding their thermal design, the characteristics features of HTGRs include low power density, high core thermal capacity with very high core outlet temperatures as high as 950°C, much higher than other reactor types. Other geometric features include a large height to diameter annular core with a steel pressure vessel, which enable decay heat removal under normal and abnormal conditions.

Modern designs utilise helium gas enabling a direct Brayton cycle to improve thermal efficiency and economics. The coolant circuit is based on gas at high pressure in the core, moving upwards to a gas plenum, cooling the external reflector regions and the upper core structures before entering the core flowing downwards. The gas then exits at temperatures in the range 800-950°C. Efficiencies of up to 50% are the target. More ambitious future designs have even higher temperatures as described below.

The power conversion unit converts the core thermal energy into mechanical and then electrical energy by means of various engineering components designed to achieve high efficiency. The gas turbine is connected to the generator, turbo-compressors to pressurise the helium, pre-cooler, inter-cooler and recuperator.

Different HTGR designers have proposed different direct and also indirect cycle designs. In the former case, the reactor vessel is connected by a cross-duct to the power conversion unit. In the latter case, primary and secondary circuits are interfaced by an intermediate heat exchanger (IHX). An advantage of the latter is to include an additional barrier against radioactive contamination of the turbo machinery. There have been considerable advances in turbo-machinery technology that have been achieved in parallel with the development of the Brayton cycle.

Below are briefly described some of the currently proposed designs of high-temperature thermal reactors. These are listed in Table 12.3.

The EFR design has been completed which aimed to encapsulate the combined experience of France, Germany and the UK for liquid metal reactor technology based on pool-type reactors. Although construction is not foreseen in the near future, there is a design now available based on established technology and with realistic cost estimates.

The EFR project was launched in 1988 by the European Fast Reactor Utilities Group (EFRUG) including EdF (France), ENEL (Italy), Nuclear Electric (UK), Bayernwerk, Preussen Elektra and RWE (Germany) and BNFL (UK) and UNESA (Spain) joined later in 1993. Other design and construction companies ‘EFR Associates’ were also involved together with R&D companies to perform supporting experimental and theoretical studies.

The design objectives’ lifetime were for high availability over a lifetime of 40 years. The technology was therefore based as far as possible on proven methods or methods that would be expected to be fully endorsed by appropriate R&D.

The reactor core consists of three radial core zones, with different plutonium contents with the inner, intermediate and outer zones with 207, 108 and 72 fuel assemblies, respectively, in a hexagonal lattice. Surrounding the core are 78 breeder subassemblies. Further, two options for the core design are possible, a homogeneous core and an axially heterogeneous core with axial breeder blankets. There are 24 control and shutdown rods and 9 diverse shutdown rods for fast shutdown.

Each fuel assembly has a bundle of 331 fuel pins and the breeder subassemblies have 169 pins. The fuel and the fertile material consist of pellets of UO2 and (U, Pu)O2, respectively. The control and shutdown rods are each retained in a hexagonal bundle of 37 absorber pins and the diverse shutdown rods each contain 55 absorber pins. These include B4C absorber material.

The reactor and its cooling systems were based on a six circuit sodium coolant design. The reactor unit is an evolution of the Superphenix design. Sodium is circulated through the core region by three primary pumps. The heat is transferred to the secondary sodium loop by six IHXs. Each secondary loop transfers heat to a steam generator unit.

The safety concept is based on the ‘defence-in-depth’ approach. The system is at low pressure and loss of coolant accidents are precluded within the design basis. The prevention is based on enhanced shutdown and removal of decay heat. Decay heat removal is normally via the steam/water plant; there are in addition two diverse decay heat removal systems. An objective is to choose a core height to minimise the sodium voiding positive reactivity effect. Reactor shutdown is assured via two independent and diverse shutdown systems.

The Institute of Nuclear Energy Technology (INET) in China designed and has operated a low-temperature heating plant from the early 1980s (Dazhong et al., 1996; Zheng, 1998). This pool-type plant provided district heating to nearby buildings. Following this activity, a 5-MW thermal experimental vessel type reactor, NHR-5 was started in 1986 at INET and went into operation in 1989.

The NHR-5 has a number of new and innovative features, including natural circulation, passive safety systems, integrated geometry and hydraulic control rod drive systems. To ensure protection of users from radioactive contamination, an intermediate circuit was added. The operating pressure of the intermediate circuit is higher than the primary side to prevent any radioactivity release to the network.

Headers and valves

YZZZZZZZZM

Water1 level

9000

Щ&Щ1

Pnmary heatexchanger

Control rod driver

4Q0Q

Concrete

Rock mass

Core

Leek momtonng device

Figure 14.1. RUTA 55 reactor. Source: Adamov and Romenkov (1996).

A commercial scale NHR (NHR-200) with a generating capacity of 200 MWt has been developed since 1990, taking full advantage of the technology developed for NHR-5. Approval was given for building in 1995 at Daqing in China. The intention is that this technology can be applied to district heating, air conditioning, seawater desalination and other industrial processes.

The loads on equipment and structures in nuclear power plants due to water hammer phenomena are being examined as part of the EC 5th Framework Programme WAHALOADS (Giot et al., 2001). The main interest is in water hammer due to condensation or shock waves. This might be caused by the inflow of sub-cooled water into pipes or other components containing steam or two-phase steam-water mixtures. Pressure waves might be generated by valve operation or following pipe ruptures.

Water hammer data are being obtained from three different test facilities.

The UMSICHT facility in Oberhausen is being adapted to simulate pipes with supports, in a configuration that is prototypic of a nuclear power plant. Experiments are being conducted with the opening and closing of valves in two 230 m pipes at different elevations, the pipes have inner and outer diameters of 54 and 108 mm, respectively. Fluid dynamic loads, fluid structure interactions and global structural response will be investigated.

The Cold Water Hammer Test Facility in FZ-Rossendorf aims to generate water hammer by accelerating a water slug to impinge on a lid flange (bouncing plate). The facility consists of a pressurised water tank connected to a horizontal pipe section connecting through a 90-degree bend to a vertical pipe section with the lid flange. The total length of the pipe is 3 m with outer diameter 219 mm.

The water hammer test rig in the integral test facility PMK-2 at AEKI will be used to perform at system pressures up to 4 MPa. A horizontal pipe of 80 mm inner diameter is connected to the head of the steam generator on one side and the steam condenser of the facility on the other side. Water hammer is generated by displacing steam in the test pipe with cold water.

There are some features of evolutionary plants that require new models and extension of the codes that have been developed for present day plants.

Some integral codes have special models that have been developed for particular plants. They therefore cannot be applied or it is difficult to apply them to new plants. It is also difficult to apply them to new experiments for the purposes of code validation. The MAAP code developed by EPRI is an integrated severe accident code, which has specific models for specific plants and phenomena. A special version AP600-MAAP has been developed for evaluation of AP600 safety (IAEA-TECDOC-752, 1994).

In general, in advanced LWR designs, there is a requirement for much stronger thermal-hydraulic coupling between the primary circuit and containment. This has led to the coupling of some system thermal-hydraulic codes, e. g. RELAP5 with the containment code CONTAIN.

Many evolutionary passive designs have large pools as heat sinks and condensers. To be effective, these need to be well mixed and the effectiveness of these pools needs to be established. The system thermal-hydraulic codes do not have the required mechanistic mixing models and therefore need to be benchmarked against CFD codes.

The system codes have limitations in their modelling of condensation, particularly in the presence of non-condensables or 3D effects.

Finally, it has been established that the performance of the system codes in buoyancy-driven situations is less robust, than in their application to the modelling of high-pressure forced convection flows, the regimes for which they were originally developed. Much effort has been expended in improving the performance of these codes in low-pressure applications in current generation reactors, e. g. in the modelling of shutdown accidents. Generally, later versions of the system codes, e. g. RELAP5 are much more robust (compared with earlier versions in this respect).

The costs of safety upgrades have been considered in IAEA-TECDOC-1084 (1999). In this section, the costs for continued operation within the design life of the plant are considered specifically, costs associated with plant life extension (and decommissioning) are considered later in Chapter 6. It is also recognised that it is not usually possible to separate out from the available data, the costs associated with plant performance or for normal equipment replacement, against the costs associated with an actual safety upgrade.

In IAEA-TECDOC-1084 (1999), costs (levelised to 1997) associated with a range of water reactor types are reviewed. These include PWRs and BWRs from the US, Korea and Western Europe (Germany and The Netherlands) and Russian-designed VVER and RBMK plants in Central and Eastern Europe and the Russian Federation.

The cost estimates per unit of plant capacity and per year were considered for different categories of plant age (in 3-year period spans) for both PWRs and BWRs (Table 2.4). Costs over 5 years are also given to enable broad comparisons to be made against VVER and RBMK data covering costs of safety upgrades on these plants, carried out over the last few years. The average figure for BWRs was somewhat higher than PWRs (Table 2.4; IAEA-TECDOC-1084, 1999). However, it was concluded that the costs were not particularly reactor dependent.

It was also found that costs of upgrades depended on the age of the unit. In the first few years, costs were relatively high associated with bringing units up to latest regulations; this was followed by a period of lower costs; costs then started to rise as ageing factors start to become an issue.

Assessments for the Russian-designed VVER series were also carried out (IAEA — TECDOC-1084, 1999); reference data are shown in Table 2.5. The VVER-440/230 design

has recognised deficiencies in relation to the integrity of the reactor vessel, the confinement pressurisation limit, and the limited scope of design basis accidents. The VVER-440/213 contains safety enhancements compared with the 440/230, particularly in terms of enhanced confinement capability, extended design basis for pipe breaks and more safety system equipment redundancy. This is reflected in the costs of the safety upgrades for VVER-440/213s, being 2-3 times lower than those for VVER-440/230s.

The VVER-1000s are better equipped again with a stronger containment. Additional enhancements have also been identified to achieve improvement of core behaviour and measures introduced to protect the integrity of the steam generators. Thus for VVER- 1000s, the costs are still relatively high. The Russian Federation estimates were much lower than the corresponding Bulgarian and Ukrainian estimates. It is worth noting that safety enhancements were implemented earlier in the VVER-440s because of the perceived urgency. The modifications of the VVER-1000s were of lower priority.

Important areas for safety enhancements of RBMKs have been identified, e. g. reduction of positive steam reactivity coefficient and improvement of the scram systems. It is clear that RBMK reactors still require investment of at least the same order as other plants, although a large part of the investments has already been made. Data are shown in Table 2.6.

The OECD study concluded that a new nuclear plant is unlikely to be the cheapest option, but that existing nuclear power plants provided they were operated and well managed can have a clear economic advantage, because of their low marginal costs. An important factor for the economic equation is whether a plant can operate reliably and in a stable condition, i. e. achieve a high load factor. Efficiency and performance are considered in Chapter 4. The load factors of nuclear plants tend to be less than those of fossil fuel plants.

The economic benefits of continued operation of the Magnox plants in the UK have been published (Mortin, 2000). The UK electricity market is de-regulated and the Magnox stations have to compete with the other electricity generators. The price for electricity in the UK market has also reduced in recent years. Nevertheless Magnox stations have continued to operate for many years and some will continue to do so for the next decade. In the main, Magnox stations have achieved very respectable load factors. Nevertheless, the marginal contribution from individual Magnox stations of the fuel purchase costs is substantial and stations are closing. Fuel costs are increasing as there is a diminishing requirement for metallic fuel, as more stations are shut down (Smitton, 1999, 2000; May, 2003).

However, in summary, there are economic benefits in continuing to operate many existing nuclear plants. At the present time, these benefits are being further realised with the help of good management providing efficient, cost effective measures for running and ensuring the safety of the plant.

4.1. INTRODUCTION/OBJECTIVES

The needs for operational efficiency and reliable performance are clearly important to the continued operation of current generation plants. This is particularly so if the operating utility has to compete in an open electricity market. This chapter concerns the issues of power plant operation in relation to performance. In general, a significant factor in achieving maximum plant performance is to operate as close to the operating margins (within the constraints of safety limits) as possible. Another factor is to ensure the plant is on load for optimal periods, i. e. plant trips, maintenance and refuelling outages are minimised. These items are covered in detail in this chapter.

Optimised fuels are being developed in order to generate more power within the operating margins, to deliver improved fuel performance, and to extend the duration of individual fuel cycles. These fuels are being developed for both currently operating and evolutionary plants. Nuclear fuel cycles are covered in Chapter 5; a whole chapter is devoted to these topics in view of their importance to operational matters.

Deregulation of the electricity industries in many countries is an important driver for utilities to improve operational efficiency and performance and reduce costs. This is likely to be an increasing global trend in the future. Deregulation took place in the UK some years ago following the break-up of the Central Electricity Generating Board. Deregulation of the US electricity industry has occurred over the past decade, following the 1992 US Energy Policy Act.

Decommissioning and waste management form part of the deeper problems that nuclear power has to face in today’s society. It is argued in Wilkie (1996) that the issues are neither technical nor just financial. Although technical issues remain, much progress has been made, and the remaining problems may be solved with sufficient financial support. This may be costly from a financial perspective, since many of the nuclear facilities that require to be decommissioned were not designed to take this into account. Nevertheless, this is not the whole problem. The problem is that the liabilities of decommissioning, e. g. spent fuel and other radioactive wastes remain hazardous over very long time scales, for hundreds of years. To deal with this problem requires stable national frameworks to sustain an appropriate nuclear industry with the required technical skills for a similar period.

The nuclear electricity generating industry involves the availability of various facilities. Facilities involved in a thermal reactor fuel cycle of the type required to support the UK reactor programme are described in Gordelier (1997). For such a cycle, the stages involve the mining of uranium ore, followed by an appropriate chemical treatment plant to produce the required uranium fuel. Following this, fuel is fabricated in a fabrication plant ready for loading into the reactor. If the fuel is to be recycled, it would be sent to a reprocessing factory where the uranium would be recovered for future use. Recovered plutonium might be for the production of MOX fuel or for utilisation in a fast reactor fuel cycle, perhaps in the future. Residual radioactive material from the fuel would be sent for appropriate storage, treatment and waste disposal. If the fuel is not reprocessed, then it would be treated as waste and sent for high-level waste storage. There are therefore many and diverse facilities associated with the fuel cycle that at some stage will come forward for decommissioning.

Different facilities pose different problems. For example, chemical treatment and fabrication plants that handle first pass fuel are relatively easy to decommission since they only handle low radioactivity materials. On the other hand, facilities that handle recycled uranium and particularly plutonium, e. g. reprocessing plants, pose a much greater challenge. The decommissioning of the power reactors themselves is also a major challenge.

There are various issues that need to be considered in the decommissioning of nuclear facilities. Many of these relate to the timing of decommissioning and dismantling and the factors that determine strategy. Some general principles are set out later, see for example Twidale (1999).

Clearly the safety of radioactive and other hazardous materials is of paramount importance. The safety of the facility will have been assured by its safety case for operation. Operations for its decommissioning phase will need to be covered in an on­going safety case consistent with the relevant national government legislation. In the UK, the policy for decommissioning is to systematically reduce the hazards until the site can be freed from licensing constraints. This is set down in the UK Government’s waste management policy (Bolton, 1996).

After shut-down, defuelling and all other clean-up operations need to be managed to minimise any risk to the general public, the workforce and the environment. The defuelling process removes a high percentage ~ 99.9% of the radioactive inventory.

In general, decommissioning should commence as soon after cessation of operations as is reasonably practicable. In particular, post-operations clean-out (POCO) should be carried out early in the decommissioning process to reduce any radioactive contamination within the plant.

The management of waste must be consistent with long-term disposal plans and no action should be taken that might prevent these plans being carried out. The quantities of waste should be minimised, e. g. waste and fuel handing operations should avoid double­handling operations if possible.

The timing of the above operations will be dependent on the existence of facilities to retrieve waste, and waste disposal routes need to become available. This could impact on the timescale for dismantling the plant, consistent with the maintenance of safety. Processing plants need to be in place together with the long-term storage facilities.

The main legislation for nuclear safety in Hungary is the Act on Atomic Energy (Act No. CXVI of 1996 on nuclear energy), which became law in 1997 (EUR 20055 EN, 2001). For implementation of the act, there are a number of regulations, 12 Government Decrees and 33 Ministerial Decrees. These are issued by various ministries: the Hungarian Ministries of Interior, Health, Agriculture, Economic Affairs, Transport and Water Management and Environment. The primary regulatory body is the Hungarian Atomic Energy Authority (HAEA).

The Hungarian safety regulations are generally non-prescriptive. There are no particular design codes defined by the Authority. However, certain requirements are set, e. g. in regard to the validation of methodologies used, etc.

The nuclear safety regulations have been set fairly recently over the past decade. At the present time, they are considered to be appropriate for future reactors as well.