Ex-vessel core melt phenomena have been studied to ascertain the feasibility of mitigation by water flooding or other means. The EC 5th Framework Programme ECOSTAR (Steinwarz et al., 2001) is concentrating on three important areas in relation to: melt release from the RPV, ex-vessel corium transport and long-term corium cooling. This programme builds on earlier projects CSC (Cognet et al., 1999), COMAS (Steinwarz et al., 1999) and CIT (Adroguer et al., 1999) to enhance the understanding of complex ex-vessel core melt behaviour, especially dispersion processes and jet formation, and their consequences.

To date, melt dispersion experiments using water/nitrogen fluids have indicated that lateral failures of the lower head lead to less melt dispersal out of the reactor cavity than do failures at the central part of the lower head. The new programme will examine the impact of fluid density on this conclusion. The erosion of different concretes with jets of iron melt and also oxide jets has also been studied. These experiments show that a metallic jet eroded the base-mat more deeply but that the oxide jet eroded a greater amount of the concrete. Melt dispersion experiments have been carried out in the DISCO facility at FZK, Karlsruhe and jet erosion is being studied in the KAJET facility, also at FZK.

Ex-vessel transport has been studied in the COMAS facility at the CARLA plant of Siempelkamp. This focuses on the spreading and distribution of the melt under molten core coolant interaction (MCCI) conditions. In a representative test, approximately 350 kg of oxidic melt are spread over a flat surface of siliceous concrete.

Reactor material experiments are in progress in regard to long-term stabilisation of the melt. Experiments with simulants have shown that phase segregation may exist within oxidic corium. Simulant experiments have been performed in the VULCANO facility at CEA using ZrO2(Al2O3. Experiments have also been conducted in the ISABEL facility to determine the plane front solidification limits. Experiments are conducted to examine the efficiency of both top and bottom flooding as a means of cooling. Several series of experiments on dryout and quenching with different particulate beds have been conducted at KTK. A large-scale top flooding of a melt pool experiment has been carried out at Siempelkamp. Bottom reflooding is being examined at FZK in 1D and 3D.

The melt coolability and concrete interaction (MCCI) (NEA Annual Report, 2002) project at Argonne is managed by the USNRC and aims to provide experimental data on the spreading of molten debris over the base of the containment and the effectiveness of water cooling from the top. It also aims to provide information on the 2D interaction of the molten corium with the concrete structure of the containment, including the kinetics.

The role of nuclear energy for carbon-free power generation is recognised by a number of national and international bodies, e. g. as noted in the UK Energy Review White Paper (Energy White Paper, 2003). An EC green paper has also been published noting the contribution of nuclear energy in meeting Kyoto Protocol targets (NEA Annual Report, 2002). However, there remains doubt internationally whether nuclear energy is a sustainable energy source. This issue has recently been discussed at the World Summit on Sustainable Development (WSSD) in Johannesburg, South Africa and the Eighth Conference of UN Framework Convention on Climate Change (COP8) at New Delhi, India.

Looking forward, a much wider range of energy generation mix is anticipated compared with the present day. For some of the options, nuclear energy is a viable source of primary energy. Nuclear power could be used to electrolyse water and produce hydrogen, or indeed can and has already been used for a number of other heat applications.

In general there is increasing environmental awareness in all the major industrialised countries, not just on the issues associated with nuclear power. The population will become more aware of the challenge of climate change and the part they can play in reducing carbon emissions. The content of carbon in fuels will increasingly become a commercial differentiator if the cost of carbon is reflected in prices. This should promote more reliance on non-carbon producing energy generators.

1.2.1 Pressurised Water Reactors

The pressurised water reactor (PWR) owes its origin to nuclear submarine reactor technology. The first civil PWR was built at Shippingport in the US and it entered commercial operation in 1957. This was a 60 MW (Net) reactor utilising high enrichment uranium fuel. This was soon followed by the Yankee Rowe plant, which included uranium oxide fuel and then other plants commenced operation both in the US and in Europe. Subsequent plants were progressively increased in capacity, in respect of the size of

the components, the number of coolant loops (increasing from 1 to 4) and overall improvements in design. Large modern PWRs now generate typically up to 1300 MW (Net).

The basic components common to all PWRs are a reactor pressure vessel containing the core and the core barrel, primary circuit loops to convey the heat to steam generators, secondary loops to take steam to the turbine, together with a variety of other systems, e. g. control and safety systems. The primary side pressure is controlled by a pressurizer on one of the primary loops. The primary circuit is enclosed in a containment. There have

been various differences in the design of these major components across the various vendors but the fundamental principles are common. Figure 1.1 shows a schematic of the modern Sizewell B PWR.

Principal PWR vendors included Westinghouse, Babcock and Willcox, Combustion Engineering in the US; in Europe, Framatome in France and Kraftwerk Union (KWU) in Germany.

Modern PWR cores comprise assemblies containing fuel rods and absorber rods in a vertical bundle. The rods are arranged in a lattice of 17 X 17 positions. Of these, about 264 positions are occupied by Zircaloy-4 clad fuel rods of about 3% enriched U-235, the remainder of positions are occupied by absorber rods.

The vessel contains light water at a sufficiently high pressure to prevent boiling. The discharge temperature and pressure are about 320°C and 15.7 MPa, respectively. Reactivity is controlled by positioning of the control rods and by managing an appropriate concentration of boron in the coolant. Water is pumped to the steam generators, from which heat is transferred to the secondary side operating at a pressure in the region of 6-8 MPa. Steam produced is passed through moisture separators and dryers before entering the turbine generator. It is subsequently condensed, reheated and returned to the steam generators and the cycle is repeated. There are some differences in detail between different designs.

Typical features of some of the principal designs are as follows. In the Westinghouse PWR for example, the steam generators consist of inverted U tubes immersed in water within the secondary side loop. Other designs, e. g. Babcock and Willcox incorporate once through steam generators, which enable the steam to be slightly superheated.

Other designs exhibit different distinctive features, e. g. in the KWU reactor there are no penetrations in the lower head of the reactor vessel. The KWU design also incorporates a

spherical (as opposed to a cylindrical) containment principle. It includes a steel containment structure encompassing the primary system, which is itself enclosed in a reinforced concrete building.

Framatome have introduced boron carbide control rods in contrast to the silver-indium- cadmium rods of other designs to enable greater flexibility of control. The company has also pioneered further improvements in respect of extended fuel cycles and the use of MOX fuel.

PWRs have operated very successfully over many years. A wealth of experience has therefore built up that has resulted in improved operational, cost effectiveness and safety. PWRs are the most widely used plants in operation in the world today, both in terms of the number of units, the quantity of electricity produced and in their distribution worldwide. Table 1.2 indicates that by the end of the 1990s, PWRs dominated the generating capacity of nuclear reactors worldwide; there are about 204 units producing a gross capacity of 203,228 MWe in 15 different countries. This trend continues today.

PWRs are refuelled off-load. During refuelling, a third of the spent fuel is removed, the remaining two-thirds is relocated to different parts of the core and new fuel is loaded. The core is arranged to provide optimal performance. A disadvantage of the PWR is that it can only be fuelled off-load, which means that the reactor has to be down for 4-6 weeks. During the outage, maintenance operations can be carried out. Typically, once every 3 years, the pressure vessel and internals are inspected, which means that all the fuel has to be removed and this outage might take up to 3 months.

In terms of running costs, these reactors along with most other current plants require some degree of uranium enrichment, and therefore fuel costs are relatively high. Against this they utilise abundantly available water both as moderator and coolant — the cost of these being low. Overall PWRs can compete economically with fossil fuel plants over many years.

PWRs have a relatively complex technology requiring diverse safety systems to guard against major loss of coolant accidents. Modern PWR designers have recognised this weakness and have attempted to simplify the complexity (and hence reduce capital costs) in new proposed designs. These are discussed in later chapters in the book.

Much experience has been gained from around 50 years of successful and largely safe operation of nuclear power plant around the world. Undoubtedly the operational safety of most reactor systems has been improved by the collective knowledge acquired from all types of reactors. Many of the principles for safe operation relate to plant management and other generic factors and are not specific to a particular type of plant.

In the first instance, safety must be built into the plant design. This is usually referred to as engineered safety. Good design can prevent significant accidents through the intervention of good safety systems. Conversely there are examples where less good design has resulted in very significant major accidents. New designs will benefit from previous operating experience.

Operational safety has generally come to relate to the performance of plant personnel and the management of plant safety at the plant. The performance of management and staff can be judged against a number of performance indicators. Recent WANO data for collective radiation exposure and industrial accident rate are shown in Figures 3.1 and 3.2, respectively. These show steadily improving trends.

Although there may be differences in detail, performance indicators utilised by different utilities have much in common. For example, in the UK, BNFL/Magnox Generation, in a recently published review of station performance, consider indicators

such as collective radiation dose, lost workday rates, the number and severity of events and the number of automatic trips (Marchese, 2000). Selective data are shown in Table 3.2 and these also indicate an improving trend.

There are a number of international standards which industry can use in assessing and improving performance. The internal safety system (ISRS) of Det Norske Veritas consists of different elements of performance. For example, element 1 relates to ‘leadership’ in putting emphasis on safety and reliability in achieving high standards of performance. The plant in question is then rated at a particular level. The system was used by BNFL/Magnox Generation in the review referred to above.

WANO performance objectives and criteria also include management performance. There is an increasing realisation that improvement in company business performance is commensurate with improvement of safety. A company needs safe reliable operation in order to be competitive. As inferred above then, both business and safety performance depend heavily on plant personnel.

The European for Quality Management (EFQM) model has been developed to provide a method for reviewing how management processes are actually working in practice. One of the features of EFQM is that both business and safety objectives and standards

should be implemented at all levels through the company. Regarding personnel development, the UK Investors in People (IIP) standard has been adopted by many companies in promoting the well being and development of their staff.

A good review of current and future fuel cycle options for LWRs and HWRs (heavy water reactors) is given in IAEA-TECDOC-1122 (1998) (Table 5.3).

5.6.1 Light Water Reactors

5.6.1.1 Present. LWRs are the most widely operating type of reactor in the world and LWR fuel optimisation is of international interest. There is intense competition between fuel vendors and there are many different designs offering different performance advantages. However, there has been extensive experience amassed on fuel performance, and fuel designs based on a conventional uranium cycle are well optimised. Thus the

differences in such designs are relatively small. There has been even some measure of collaboration between fuel vendors arising from the need to share costs associated with expensive research programmes. The common drivers in fuel design are to achieve greater reliability, to reduce fuel failures, to move towards higher burn-up and to reduce fuel cycle costs. These fuel performance issues are considered in the Section 5.6.1.2.

In addition to the conventional uranium fuel cycle for LWRs, MOX fuel has also been used and is well established. In the MOX fuel cycle, plutonium oxide is mixed with uranium dioxide for use as fuel in LWRs. MOX fuel is used in France, Germany and Japan. It was first used in Europe and the US in the mid-1960s and since then hundreds of tonnes of MOX fuel have been burnt in commercial LWRs. The success of burning plutonium in MOX fuel demonstrates that plutonium is an asset that can be used for civil nuclear power generation. Further this has been realised by the development and safe operation of large-scale plutonium recycling facilities in France and most recently in the UK, now that the BNFL Sellafield MOX plant has become operational. The IAEA have put in place controls to ensure adequate safeguarding of materials.

5.6.1.2 Future

MOX. MOX fuels represent the most significant developments in LWR fuel technology, particularly in Europe (IAEA-TECDOC-1122, 1998). MOX fuel up to 30% loading can be used in LWRs within current operating and safety margins; higher percentage loadings would require control rod changes to maintain current margins. MOX fuel costs are higher than UO2 fuel costs but this largely reflects reduced production at the present time. The potential of advanced MOX fuel is being studied in France and Japan.

CEA are investigating advanced plutonium fuel assemblies to overcome the problems of multiple plutonium recycling in PWR MOX assemblies. As MOX assemblies are irradiated, the isotonic quality of the plutonium is reduced (Groullier, 2001). CEA are working on high moderation plutonium fuels (Youniou et al., 1998). In conventional MOX assemblies, the moderator/fuel volume of MOX is the same as in UO2 assemblies and new designs are being investigated to increase this ratio which gives a more complete thermal flux and reduces the initial plutonium content. Conversely, the Japanese are looking to lower moderation fuels to achieve plutonium breeding, see for example Tochihara et al. (1998).

Thoria Fuel. The development of thoria fuel has been overshadowed by the emphasis and investment in uranium-based fuel. Nevertheless, thorium is about three to four times more abundant than uranium and represents a good long-term nuclear fuel supply. The cycle produces fissile U-233, thereby enabling breeding potential in a thermal reactor, good in-core behaviour and lower excess reactivity requirements. A disadvantage is that thorium ore does not contain a fissile isotope and so U-235 or Pu must initially be used in conjunction. Thorium fuel is attractive for various reasons. There is very little production of plutonium or transuranics, which reduces the radiotoxicity burden and, therefore, there is a benefit from the point of view of proliferation (Hesketh, 2003). Thoria fuel has been successfully demonstrated in power reactors.

Uranium Free Fuels. The incineration of plutonium from weapons programmes and from reprocessed LWR fuel is under consideration in many countries. Another pressing issue to the nuclear countries is how to burn actinides as part of a waste management strategy. Research programmes are underway in Switzerland, Japan, France and Canada. The idea is to burn the plutonium (or actinides) in a non-fissile inert carrier matrix. Various fuel matrices are being examined, e. g. zirconium oxide in Switzerland, fluorite and spinel in Japan, ceramic (spinel, magnesia) or metallic matrices in France and silicon carbide (SiC) in Canada. Other materials may also be required, burnable poisons (e. g. erbium) for control of reactivity and addition of thorium or uranium to enhance negative temperature coefficient. To date, fuels have largely been irradiated with accelerators; initial results are good for SiC and zirconia. Some in-reactor irradiations have taken place. The main issues relate to materials performance that are not yet resolved and inert matrix fuels are unlikely to enter LWR fuel cycles in the near future.

Already, a considerable degree of harmonisation has been achieved within the international community, on the principles of safety for commercially operating reactors. The implementation of these principles may be achieved at different levels across the countries operating nuclear plant but considerable progress has been made. Further, international safety standards will become increasingly stringent. This means that future reactor designs are likely to have to demonstrated even higher standards of safety than at present, to meet more demanding national regulatory requirements and international safety standards.

In order to do this, design principles will need to be considered for future plant (Carnino, 1999), which build on the principles already established for present generation plant. These are discussed below.

There needs to be assurance that all technical safety needs are complied with in design. The following safety design principles are now accepted in most countries, operating nuclear plant (Table 7.4). Many of these have been put forward by the IAEA and are included in the IAEA list of 25 safety principles, listed in the next chapter.

The design must be such that plant operation is reliable, stable and manageable. Prevention of accidents is the prime goal. For many new evolutionary designs, the goal has been extended to provide better protection against severe accidents (Table 7.5).

Table 7.4. Safety fundamentals in design

Design must ensure the nuclear installation is suited for reliable, stable and easily manageable operation Design must include appropriate defence-in-depth principle Technology must be proven or qualified by experience or testing or both Man-machine interface and human factors must be included in the design and in the development of operational requirements Radiation exposures to site personnel and releases to the environment must meet ALARA principles

A comprehensive safety assessment and independent verification must confirm that the design meets the safety objectives before the operator completes his submission to the licensing authority

The ‘defence-in-depth’ principle that a number of levels of protection and multiple barriers are included to prevent radioactive release is well accepted. This ensures that the combinations of failures that could occur that could lead to a significant release are of very low probability. In advanced designs, the tendency is to increase the robustness of this principle by appropriate design.

An important requirement is to ensure that the design technology is proven. Advantage should be taken of experience, if relevant, if not by further testing or possibly a combination of both.

Man-machine interfaces and human factors must be considered in the design and must be incorporated into the development of operational requirements. A key objective of newer designs is to reduce human errors.

The ALARA principle should be adopted in the design in respect of staff exposure on site and in the releases of radioactive materials to the environment. A reduction of exposures is the goal in newer designs.

Confirmation of the design via a comprehensive safety assessment and independent verification should be carried out to ensure that safety requirements are met prior to submission of the case to the regulating body.

The case must show that the risk to workers and the public is continually decreasing and demonstrate that operation is environmentally friendly.

This can be achieved by a suitable containment, which is designed to reduce the frequency of large releases to very low levels. This needs to be demonstrated via appro­priate analysis (probably via deterministic and probabilistic means in addition to improved defence-in-depth).

In general, the protection of the workers and the public impacts must be demonstrated in the design, operational procedures and environmental assessments.

Development of a transparent and stable process for the licensing of plant.

A well-established and stable generating framework is an important requirement with good interfacing between the licensee and the regulatory body. The process can be enhanced via a rigorous self-assessment process coupled with independent assessment.

Need to gain public acceptance on the benefits of the proposed new design.

Harmonisation of regulatory approach, which may be more possible for new designs, is a good means of increasing public understanding and acceptance of nuclear safety.

An extremely important requirement is that there should be no serious accidents on current plants and that the nuclear industry is seen to act with integrity.

Safety requirements can be met while still maintaining costs at a level for nuclear plant to remain competitive with other generators.

The economics of nuclear power generation is improved by longer fuel cycles and by longer life (including life extension on current plants). This will clearly remain true for new designs as well.

There are nine nuclear reactors currently operating in Spain. In 2002, 26% of the country’s electricity was generated by nuclear energy. The previous year the figure was 27%; during this period the electricity consumption grew overall by 2.7% (Foratom e-Bulletin, 2003b). In 2002, the average load factor was in excess of 90%.

However, Spain announced a moratorium against building of new nuclear plant as early as 1984 (European Commission, 2000). Currently there continue to be no plans for building a new nuclear plant in the near future.

Ways in which to substantially enhance the efficiency of LWRs have been studied for some time. Efficiencies as high as 44% are possible by operating in a thermodynamically supercritical regime. Supercritical high performance reactors are one of the candidates of the Generation IV initiative (The US Generation IV Implementation Strategy, 2003) for medium term deployment. The European Commission is also currently assessing the merits and feasibility of such an approach in a project involving European institutes and industry in collaboration with the University of Tokyo (Squarer et al., 2001). A review of supercritical reactors has been carried out by Oka (Proceedings of the First International Symposium on Supercritical Water-Cooled Reactors, 2000) and the EC project is assessing the available technology against a reference design (Dobashi et al., 1998). There have been considerable advances in this technology in Japan.

Supercritical water reactor (SCWR) systems are principally aimed at electricity production. Their high thermal efficiency offers a potential for improved economics compared with current generation LWRs. An important issue in regard to these systems is the need to develop materials and structures that can serve in the high temperature and supercritical pressure regimes of these plants. A sample of designs currently under consideration is given in Table 12.2.

The concept is based on a once-through cycle, operating in excess of the water critical pressure of 22.1 MPa. Water enters the reactor core and then exits without change of phase. This system has the advantage that no steam-water separation is necessary, which in principle leads to a simplified (and therefore more economic design). Heat is removed from the system via a coolant of very high temperature and because the system is single phase, the turbines are driven directly by the primary coolant. Typically, water enters the core at about 280°C and exits at 500°C or higher, yielding efficiencies of the order of about 44%.

Supercritical systems have been considered at various times over the past 50 years, initially by Westinghouse and GE and in the last decade by Kurchatov Institute and AECL, based on a CANDU system. The early Westinghouse and GE designs were light water cooled. The Kurchatov and AECL designs were graphite moderated and heavy water cooled respectively; however, these required larger reactor volume and complicated systems. This resulted in less favourable economics. The Russian design, based on an integrated supercritical PWR design, was cooled via natural circulation but was more limited in scale and power. Heavy water super critical systems are considered below.

There have been various other types of supercritical reactor designs considered, including fossil plant systems, the GE nuclear super-heater, a steam cooled FBR (FZK), a B&W design, and a University of Tokyo steam-cooled FBR.

A national programme OMEGA started in 1988 for R&D in new technologies for the partitioning and transmutation of high level waste (Takizuka, 1997). The OMEGA programme consists of two areas of research, the separation of elements from high-level waste based on their physical and chemical properties and the transmutation of MAs and LLFPs into short lived or stable nuclides. A conceptual design programme has been put

together including code systems (Nakahara and Tsutsui, 1982; Nishida et al., 1990) and integral experiments (Takada et al., 1992) to investigate two concepts, a solid system and a molten salt system.

For the solid system, the design is based on a sodium-cooled fast reactor. The accelerator injects a 1.5 GeV proton beam onto a tungsten target, surrounded by a sub­critical blanket of actinide alloy fuel. The target blanket is at a total thermal power of 820 MW cooled by downward flowing sodium. The remainder of the heat transfer cycle is based on a tertiary cycle system (Figure 13.2).

The other design study is based on a molten salt target/blanket system, generating 800 MWt. The molten salt acts as a fuel and target and also as a coolant. The beam is

1.5 GeV. The latter concept is based on future generation reactor technology.

As noted in the previous section, there is continued interest in synthetic liquid fuel (SLF) production in China and Russia. The energy consumption and production in China is

dominated by coal and there is a shortage of liquid fuel supply. The application of high — temperature reactors to convert coal to liquid fuel is, therefore, of interest.

There is also interest in Russia in converting low-grade brown coal to motor fuel. The VGM-P, HTR-10 and BN-600 systems, as described in the previous section, are seen as possible candidates for refining of coal, which requires still higher temperatures than are needed for oil refinement. The high-temperature gas reactors could also be used in the production of hydrogen, ammonia and mineral fertilisers by, e. g. methane steam conversion or other means.