Switzerland has five units operating which in 2002 generated 40% of the country’s electricity requirement. There has been debate on whether Switzerland should exit from nuclear power but in April 2003 (Foratom e-Bulletin, 2003b), the Swiss parliament approved a new law providing for the nuclear option to be left open. This law extends the rights of Swiss citizens to take decisions on the future use on nuclear energy and also in the licensing of waste repositories. There is a recommendation of both government and parliament to vote against forthcoming referenda, which propose to phase out nuclear energy and to replace nuclear power plants by alternative energy sources over an unspecified period.

12.2.2.1 CANDU SCWR (Gen IV). The SCWR concept is also being considered as an evolution from CANDU reactor technology. As with the LWR systems, there is an aim to continually enhance the design and applications of the CANDU system. Thus, complementary to the SCWR loop concept described above, a channel design option with multi-stream products is also possible within the SCWR context. The CANDU SCWR concept is envisaged for flexibility of application, e. g. including electricity production, hydrogen generation (direct or indirect) and high-temperature process heat applications, depending on demand. It could also have desalination applications (Generation IV Seminar on Nuclear Energy Systems Research and Development, 2004; Duffey, 2004a, b). It is seen as part of the evolution towards the CANDU X system, sometimes referred to as a Generation V system. The CANDU X design could also be economically competitive. The main elements of the CANDU X system are described below.

12.2.2.2 CANDUX. The CANDU X concept is another pressure tube reactor in the CANDU family of reactors. The design is being put forward by AECL in Canada. It has a flexible generating capacity, in the range 350-1150 MWe. This depends on the number of fuel channels in the plant.

The innovative features of the Mark 1 model include supercritical heavy water for the reactor coolant and supercritical light water for the turbine generator. The utilisation of supercritical water results in a significant increase in system pressure and temperature compared with earlier generation CANDU plants.

CANDU X retains the use of two passive shutdown systems as in current generation plants. There is also passive decay heat removal even if the reactor system is empty of coolant.

The CANDU X reactor possesses a number of the attributes expected from future generation systems. It has high efficiency due to increased core outlet temperature. There is flexibility in reactor power scale available through extensive modularity in design.

Regarding its fuel cycle and waste management concerns, the option to use thorium fertile material and slightly enriched fuels is available to reduce the level of minor actinides produced.

As for supercritical light water systems, the main applications would be for electricity generation. However, the high core outlet temperature increases the number of process heat applications that are possible. The temperature is higher than can be achieved by current water reactors but lower than can be achieved for HTRs and LMFRs. The attractiveness for process heat applications is particularly true for the smallest 350 MWe version.

CERN have put forward a conceptual design for a fast neutron operated high-power energy amplifier (EA) (Figure 13.4). The principles are described in detail in Carminati et al. (1993), Rubbia et al. (1994) and Andriamonje et al. (1995). More recent optimised realisations are described in Rubbia et al. (1997). The EA can operate for an indefinite period in a closed cycle. The fuel load is discharged, apart from fission fragments, and then re-introduced into the sub-critical unit, but with natural thorium introduced to compensate for burnt fuel. Equilibrium is achieved between burning and incineration after several cycles. This represents an extremely efficient use of fuel.

The EA module includes a 1500 MWt unit with a 1.0 GeV proton accelerator of 12.5 mA. The accelerator is a modular cyclotron; i. e. a plant is made up of a number of modules.

Figure 13.4. Conceptual design concept of the diffuser driven energy amplifier. Source: Carminati et al. (1993).

A fast neutron EA is envisaged if the EA has power commensurate with the current of large pressurised water reactors. The proton beam is a novel element of the design; the current is lower by one order of magnitude than most LINAC designs. The anticipated efficiency, i. e. the beam power over the mains load, is of the order of 40%. The beam penetrates the EA through an evacuated tube and tungsten window, specially designed to withstand radiation damage and thermal stress. The electrical energy to operate the accelerator is about 5% of the primary energy production.

The coolant is molten natural lead at a temperature of 600-700°C; lead being chosen because of its high boiling point (1743°C), which combined with the negative void coefficient of the EA, enables very high operating temperatures to be reached. Heat can be removed by natural convection.

The EA can operate with various different fuels, for plutonium to be transformed into 233U, the EA would be initially loaded with actinide waste and thorium. Other actinides, e. g. americium, or neptunium could also be added. The EA mixture is sub-critical with keff in the range 0.96-0.98.

The Chinese are testing PBMR technology within the HTR-10 project. HTR-10 is a small test reactor of only 10 MWt and is operated by the Chinese Institute of Nuclear Energy Technology (INET). The reactor first went critical in late 2000. The fuel has been fabricated in China but is based on German fuel technology. The thermal-hydraulic cycle is being tested in several stages. Initially the steam/power cycle is being verified. This will be followed by testing of the gas turbine cycle.

The Institute of Nuclear Technology (INET) in Beijing, China, is developing several reactor systems for non-electrical applications (Sun et al., 1998). Technologies for water — cooled heating reactors and for modular high-temperature reactors are being developed. A 5-MW water-cooled test reactor was constructed in 1989 and feasibility studies for seawater desalination using the reactor as the power source are in progress. For high- temperature applications, a 10 MWt test reactor, HTR-10, is seen as a step towards developing a commercial HTGR demonstration plant.

The innovative designs pose new challenges for inspection and maintenance teams. Integral systems will be very compact; there will need to be non-intrusive monitoring techniques, continuous monitoring in confined space systems. The high temperatures will require more remote and possibly robotic systems.

There has been significant progress in advanced monitoring and control. The lessons learned for the current and evolutionary systems will provide valuable input into future innovative reactor design.

Fuel purchase costs for nuclear plants are generally low in comparison with other energy producers (Table 2.3). These costs are an important differentiator to the competitiveness of nuclear vs. non-nuclear plant. Nevertheless, fuel purchase costs are still high and can significantly affect the economics of plant operation.

For a nuclear power plant, over half of the generating costs relate to the initial capital investment. Fuel accounts for less than 25% of the total generation cost and in recent years, fuel cycle costs have decreased significantly in all countries. Conversely for coal and gas, fuel costs are the most dominant, representing 40-80%, respectively, of the total generation cost. Regarding other costs, operating and maintenance (O & M) costs represent only a small part of the total generating costs of nuclear power plants. These O & M costs relate mainly to the technical performance of the plants, safety regulations and staff costs. Decommissioning costs’ issues are discussed later in Chapter 6.

The security of nuclear installations has become an issue following the concerns of global terrorism in recent years. Draft guidelines are being drawn up by the IAEA to enable utilities to carry out self-assessment of the safety and security of their installations. There is a general impetus to promote interaction between staff from safety and security backgrounds and to harmonise terminology. The safety and security of radioactive sources has been identified as an issue by IAEA, following the accidental overexposure of individuals from ‘orphan’ sources and from concerns arising out of September 11, 2001.

An IAEA international conference on the security of radioactive sources addressed this issue in 2003 (IAEA International Conference, 2003b).

National regulators govern the waste management programmes in their countries. Nevertheless, there is a considerable harmonisation of policy and principles in regard to waste disposal safety. The UK approach is considered by way of example (Cmnd 2919, 1995). There are also however differences, particularly in regard to the high — level waste disposal issue, already discussed earlier.

In the UK, the same legislative framework exists for waste management and decommissioning, as exists for operating nuclear power plant. Activities are governed by the Health and Safety at Work Act, 1974 and the associated statutory provisions of the Nuclear Installations Act, 1965. More details are given in Chapter 8.

The UK national policy was reviewed in the 1995 White Paper; the conclusions are given in Cmnd 2919 (1995). The UK Health and Safety Executive (HSE) has defined 10 policy issues. These are summarised in Table 6.2.

A key requirement is the need for strategic planning. Where disposal routes exist, the general principle is to move towards long-term storage with the waste in a passive safe form, rather than an approach that requires frequent monitoring.

The fundamental legislation for nuclear energy in France is based on the Decree on Nuclear Installations issued in December 1963, together with further decrees in 1970, 1974 and 1984 (EUR 20055 EN, 2001). The regulatory body is formed within the Ministry for Industry and administered by the Ministry for Environment. It is represented by DSIN (Direction de la Sdrete des Installations Nucleaires) which is responsible for regulation and inspection of the plants.

The regulatory regime is not prescriptive; no particular design codes are prescribed by DSIN. However, Basic Guidelines for Safety RFS (Regles Fondamentales de Srnete) are defined by DSIN. In practice, American design codes were used (ASME) but later, French design codes (RRC: Regles de Conception et de Construction) have been developed by the French industry that meet the requirements of the safety authority.

In 1989, France and Germany agreed to harmonise their safety approach for future reactors. In 1990, the safety authorities of both countries formed the DFD (Deutsche — Franzosische Direktion) forming close links between DSIN on the French side and BMU on the German side (Bundesministerium ffir Umwelt, Naturschutz und Reaktorsicherheit). In 1992, the DFD agreed that the DSIN and BMU would establish a common safety approach for future reactors.

There are six units operating in Slovakia that provide a large share of the country’s electricity requirement, 73% in 2002 (Foratom e-Bulletin, 2003c). Two of these are the first generation VVER-440/230 design, Bohunice 1 & 2. These have been extensively refurbished in 2003, including modernised control systems and replacement ECCS. Nevertheless, these units are due to close by 2006 and 2008, respectively (World Nuclear Association, 2003). There are also four second generation VVER-440/213 plants in operation, Bohunice 3 & 4 and the newest plants, Mochovce 1 & 2. Upgrading projects for Mochovce, e. g. replacement of the I & C system have been implemented. A similar upgrade is scheduled for Bohunice 3 & 4 (EUR 20056 EN, 2001).

6.3.7 Slovenia

Slovenia has one operating PWR unit, which in 2002 generated 41% of the country’s energy requirement (Foratom e-Bulletin, 2003c). The Krsko nuclear power plant is jointly owned with Croatia. The two countries now have an agreement on the status of the plant and also for the management of radioactive waste. It also provides for a joint decommissioning programme.