CAREM-25 is an integral type PWR in which all components of the primary circuit (NSSS, pressuriser, primary heat exchangers and coolant pumps) are included in a single pressure vessel (IEA/OECD (NEA)/IAEA, 2002). It is being applied in Argentina as a dual purpose plant for electricity and process heat for the extraction and purification of various minerals including sodium sulphate.

A concern for smaller reactors of traditional design is that the economies of scale militate against the economics. With an integral design, savings can be made by reducing the number of pressure loaded and load-bearing components. Considerable emphasis is placed on inherent safety features in the design, including passive shutdown systems and decay heat rejection.

In addition to the economic competitive features mentioned above, there is scope for design modularisation and factory assembly. The other characteristics of operation relating to proliferation, waste management, resource efficiency, flexibility of operation (process heat and co-generation) are similar to those of other current generation PWRs of comparable size and application.

The NEA also maintains a shielding and criticality database for the validation and benchmarking of methodologies for modelling different nuclear systems. For example, the database includes data from the integral shielding experiments (SINBAD). New editions of the database are being continually released. With regard to advanced technologies, radiation shielding for accelerator facilities has been the study of a recent workshop hosted by the Stanford Accelerator Centre, (NEA Annual Report, 2002). For criticality, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) database, is also being developed and contains several thousands of benchmark specifications for critical or near-critical configurations.

For small-scale power generation, it is anticipated that hydrogen fuel cells will be playing a greater part in the economy, initially in a static form in industry or as a means of storing energy. The hydrogen would be generated by non-carbon electricity. Hydrogen can be produced in many ways, e. g. renewable energy sources such as hydro, solar, wind power, electrolysis, biomass and by nuclear energy. Nuclear power could be used to provide electricity for electrolytic hydrogen production. Fuel cells could also be used to back up intermittent renewables. Fuel cells are an area of active research (N. B. in addition to hydrogen, it should be noted that biofuels are another possible option for fuel cells).

Transport is still a major contributor to air pollution and carbon dioxide emissions (about 30%). For transport, hydrogen could be increasingly used for fuelling public service vehicle fleets and utility vehicles and is, therefore required as a primary source. It could possibly be used in the car market where hybrid internal/combustion/electric vehicles would be commonplace in the car and light goods sectors. N. B. For these there is also likely to be a substantial and increasing use of low carbon biofuels. (It is worth noting that other innovative technologies are being investigated for transport, e. g. vehicles powered with batteries that can be charged by electromagnetic induction from metal plates buried in the road at selected stops.)

There is an increasing interest in hydrogen as an energy system, produced from a carbon dioxide free process (The Parliamentary Office of Science and Technology, &, Millbank, London, 2002). Hydrogen may have a number of widespread applications as a fuel for road transport, distributed heat and power generation and for energy storage. The most likely use for hydrogen in the UK and in other countries, is for transport, for fuelling fleet vehicles and buses. The Energy Saving Trust (EST) (The Parliamentary Office of Science and Technology, &, Millbank, London, 2002) refers to the use of hydrogen in fuel cell vehicles as ‘the most promising option for zero carbon road transport’. The Institute for Public Policy Research (IPPR), an UK think-tank and the Carbon Trust, a non-profit company set up by Government to take a lead on low carbon innovation in the UK, are supporting the case for a high-level strategic approach towards developing a hydrogen economy (The Parliamentary Office of Science and Technology, &, Millbank, London, 2002).

There are a number of international initiatives towards developing the hydrogen economy including IEA, EC and OECD activities (http://www. iea. org/workshop/2003/ hydrogen). There are major international activities in train, the EC has announced a large programme on hydrogen and renewable technologies, the US is supporting a five-year programme on hydrogen, fuel cells and related infrastructures and the Japanese have substantially increased their level of activity on hydrogen research since 1995 (http:// www. iea. org/workshop/2003/hydrogen).

The EC has set up a high-level Group to assess the prospects for using hydrogen and fuel cells in transport and overall energy policy (http://www. world-nuclear. org/news/2002/ wd_oct18.htm). The EU Clean Urban Transport for Europe programme aims to provide fuel cell buses in 10 European cities in the near future, including 3 in London (The Parliamentary Office of Science and Technology, &, Millbank, London, 2002). Also there is a European Integrated Hydrogen Project (EIHP) which aims to create a harmonisation of necessary legislation in the EU for hydrogen safety, infrastructure and standardisation (http://www. world-nuclear. org/news/2002/wd_oct18.htm).

In the UK, the Engineering and Physical Sciences Research Council (EPSRC) which funds a UK hydrogen energy network, also promotes hydrogen research (The Parliamentary Office of Science and Technology, &, Millbank, London, 2002). There are calls for a dedicated programme to co-ordinate and support UK research initiatives and support demonstration projects. The use of hydrogen as a fuel for buses is being pursued in the Cambridge Urban Solar Hydrogen Economy Realisation Project (The Parliamentary Office of Science and Technology, &, Millbank, London, 2002). Hydrogen fuel cells are being developed for local heating and energy supply applications.

Pressurised heavy water reactor (PHWR) concepts have been developed in a number of countries, including Canada, Japan, France, UK and others. Interest in heavy water as a moderator arose because it overcame the problem in LWRs of relatively high absorption of neutrons, enabling the reactor to operate at lower enrichment, or even with natural uranium.

However in the UK, the steam generating heavy water concept was not taken forward to commercial operation because the economies of scale were not favourable in comparison with alternatives.

The main objective in the management of radioactive waste is to ensure the protection of the public, workers and environment by isolating all hazardous material from the biosphere. Technologies are required for the various stages of waste management, i. e. the handling, temporary storage and long-term disposal. To ensure safety, the handling, storage and disposal of all waste materials arising from all plant operations is carefully managed according to the degree of hazard.

Wastes are obtained from all stages of the fuel cycle, from uranium extraction, refining, reactor operation and decommissioning. Waste can arise in gaseous, liquid or solid forms. Some of these waste products are radioactive with varying degrees of activity. They include low-level radioactive spoil from uranium mining residues and uranium and plutonium residues from fuel fabrication.

Reprocessing plant operations produce waste of medium level of radioactivity, arising from various waste streams, and hulls from residual cladding and support materials from fuel elements. There are other low-level wastes from reprocessing operations, including miscellaneous items such as gloves, containers, etc.

The reactors produce highly radioactive waste in the form of spent fuel; this issue is discussed further in the next section. They also produce inert gas (waste) from fractured fuel elements, liquid waste in the form of tritiated water and solid waste, e. g. filters and resins. The latter arise from water treatment plants; these resins are used to clean up primary system fission or corrosion products.

Decommissioning also produces mildly radioactive structural materials.

Drainage water from reactor support systems, fluids from decontamination operations and ion-exchange resins in the form of liquid effluents are collected in tanks. These effluents are then categorised and distributed to various sub-systems depending on their activity or impurity content.

Low-level effluents are discharged under controlled conditions back into the reactor plant. Solid low-level wastes may be compacted and then encased in stainless steel drums. Intermediate effluents may be treated with ion-exchange filters or evaporated, with the purified water returned to the reactor coolant system. Active resins or concentrated active liquids are stored in tanks, to allow for decay of the shorter lived isotopes, before being sent for waste treatment. Solid wastes such as filter resins evaporation residues are treated and then cast in concrete.

More details on the scale of the global nuclear waste management issue and present and possibly future containment practices are given in Chapter 6.

The development of fast reactors for electricity generation is largely in abeyance except in Japan and Russia. Work is, however, progressing on the use of fast reactors for consumption of excess plutonium and the destruction of minor actinides (MAs) and long — lived fission products (LLFPs). However, a number of fuel cycle options are being investigated, e. g. in the CAPRA-CADRA programme (Hesketh, 2003).

5.6.4.1 Plutonium Burning Fuel Cycles. Fuel cycles based on a fast reactor plutonium breeding cycle are being investigated for possible adaptation to plutonium consumption. In the CAPRA-CADRA project the European fast reactor (EFR) concept is being considered whereby the plutonium content of a MOX fuel assembly is increased at the expense of U-238; this, therefore, results in net plutonium destruction. Inert matrix assemblies have also been considered based on a plutonium nitride cycle.

5.6.4.2 Minor Actinide Target Fuels. Fuel cycles that produce and burn equal amounts of plutonium and which can destroy MAs and LLFPs are the subject of advanced fuel research. To do this, conventional fast reactors are possible but accelerator driven systems (ADS) may have some advantages. The latter are considered later in the book. From the view point of target fuel assemblies both designs are similar. There are essentially two approaches to fuel design:

homogeneous — where the MA or LLFP is mixed with the fuel, and

heterogeneous — where there is a separate target assembly.

The target materials being considered include oxide, nitride and cermet (ceramic/metal) fuels; of these, nitride fuels are the most promising. There are many options that are being researched, see for example (Hesketh, 2003).

There are a number of different evolutionary water reactor designs that are at different stages of maturity. One distinctive difference between these designs is that some incorporate established active safety systems whereas others rely on passive systems to provide some safety functions such as long-term heat removal (Yadigaroglu et al., 1999). More details of some of these different approaches are described in IAEA-TECDOC-1117 (1999) and Yadigaroglu et al. (1999).

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 ITEP, Moscow, together with a number of other Institutes (Shvedov et al., 1994; Chuvillo and Kiselev, 1997), have conducted investigations in the use of ADS.

The main objectives are as previously discussed including a means of utilising large amounts of weapons-grade and commercial plutonium, and for waste management, etc. These developments are being considered against possible future scenarios for the Russian nuclear power industry, e. g. continuing the development of new generation NPPs, which would include improved VVER type reactors and possibly BN-800 designed fast reactors.

Different modes of operation are being considered including transmutation with and without power utilisation, the production of new fissionable materials and long lived radioactive waste transmutation, and the utilisation of NPP spent fuel assemblies as nuclear fuel. Within these modes, fuel cycles include uranium, plutonium, uranium-thorium, plutonium-thorium and other actinide fuel cycles.

Both solid and liquid fuels are being considered. The use of oxide fuel and zirconium cladding would enable advantage to be taken of existing fuel production experience. MOX fuels with plutonium, cermet and nitride fuels and actinide addition into MOX fuel may be future options.

ITEP have been investigating fluoride molten salts of the type Li-BeF2-ThF4-Pu4, which have some advantages in reduced radiation damage, and reduction in the amount of fission products. However, the existing knowledge base on the performance of such fuels is more limited.

Different designs of blanket are being considered (IAEA-TECDOC-985, 1997b). These include blankets of solid fuel, blankets for liquid fuel and a modular channel blanket and a design for liquid fuel with a homogeneous blanket.

Conceptual targets include solid tungsten and other materials for proton currents up to 30 mA and liquid targets made of lead and lead-bismuth eutectic for high values of proton current. Figure 13.5 shows an example of a design with a lead-bismuth target with both fast and thermal blankets of Pu and Th oxides.

Experimental studies are being carried out at ITEP to verify the various concepts. These relate to both ADS design and to the selection of appropriate materials, e. g. in relation to the target and blanket.