The main areas that contribute to nuclear plant availability and reliability have been investigated in an IAEA study of six representative plants (IAEA-TECDOC-1098, 1999; Table 4.3). This involved case studies of plants from Western and Eastern Europe and the US. The plants chosen for the study had exhibited high-energy availability factors and also improvements over recent years in safety, availability and reliability. In addition, the plants

covered different types of water reactors including PWR, BWR, VVER and PHWR designs.

The IAEA study concluded there were a number of management practices that contributed to good plant performance, e. g. organisational structure, strategic planning and objectives, management involvement, internal communication, quality management, relationships with contractors and financial management.

Management philosophy should embody a wide range of core values, which should be conveyed to all its employees. These include diverse aspects including environmental respect, economic competitiveness, the engendering of team spirit and also the characteristics of trust and integrity. It should also include technical aspects such as adherence to ALARA principles and ageing.

Factors that had a direct influence on plant performance included personnel characteristics, the training and development of personnel and the behaviour and attitude of personnel. This conclusion was common for all staff in the workforce, including permanent employees and contractors.

Another conclusion concerned working practices. These included the monitoring of the plant states, the quality of operating procedures, maintenance policy, technical support and interaction and communication between different work groups.

There is general agreement across nuclear plant operating countries, on the necessary working practices for plant performance improvements. These relate to plant status control monitoring, the quality of operating procedures, maintenance policy, technical support and interaction between various informed working groups. The implementation of these practices inevitably varies from plant to plant, depending on local strategies. It is also widely recognised that there are benefits in utilising all levels of local, national and international experience to continue to improve performance.

The issue of public concern surrounding decommissioning is largely centred on the concerns of safe waste disposal (de la Ferte, 1996).

In reactor safety research, there has been a continued drive to improve the understanding of severe accidents in order to prevent significant releases of radioactivity under severe accident conditions. The work has been focussed on different levels corresponding to the defence-in-depth principle discussed earlier.

The first objective during a core melt accident is to maintain vessel integrity following attack by molten corium or debris from the higher up reactor core internals. The corium may or may not be in contact with water and even in the latter case may not coolable. Various research programmes have been carried out including investigation of early and late phase melting phenomena under different accident conditions and the energetics of corium/water interaction studies of heat transfer-related mechanisms from the debris to the vessel. Others include thermal-hydraulic cooling of debris beds, investigations of the structural response of the vessel and examination of the effectiveness of cooling of the vessel with water from the outside, etc. Theoretical programmes of work supported by experimental programmes have been carried out. The combined programmes have considered scaling effects and also how to extrapolate results obtained from simulant materials to reactor materials.

If the vessel is breached, then molten debris will be released in the cavity beneath the reactor. It is therefore important to understand the physical processes of the potential release of melt from the vessel, how it will spread over the concrete floor or how it might be impeded by other retention structures (in some of the newer designs). It is likely that there will also be interaction of corium with water from discharge of emergency core cooling systems (ECCS) and the need to quantify the load on the containment from any resulting steam explosions.

The EC programmes have covered experiments and theoretical studies on the thermochemistry of molten corium interactions with structures. They have included projects to determine the production of hydrogen and other non-condensable gases, e. g. carbon monoxide to establish the threat to containment from gas combustion. Another facet was to consider the retention of fission products in the melt with respect to the ‘Source Term’ (see below). Work items covered vessel failure and corium release modes, corium spreading effects and the consequent impact on direct containment heating. It also covered the interactions of corium with water and structures and generic studies on retention devices (e. g. core catchers).

An important aspect of severe accident research has been to quantify the Source Term. This is defined to be the quantity, timing and physical form of the radiological and chemical species release to the environment. It is dependent on the type of accident. The important inputs for determining the Source Term are fission product release from the fuel, and the transport in the primary circuit and the containment. Also important are the suspension, resuspension and condensation/revaporisation mechanisms within the reactor circuit and containment. Accident mitigation devices such as sprays and other measures have an important mitigation effect (Table 8.5). There are important large-scale integral tests and supporting separate effects tests to provide data to validate computer codes for analysis.

The ultimate Source Term to the environment depends on whether the containment is breached. There are different threats in the short term and long term.

Assuming the containment holds the Source Term will depend on the leak tightness of the containment. The short-term threat arises from corium/steam explosion, hydrogen

combustion, more particularly detonation, direct containment heating, secondary effects of missiles and containment isolation failure. The long-term threat comes from the build up of heat (and therefore pressure) due to failure of removal of decay heat or by failure of isolation devices, i. e. material failure. The containment strength is very design dependent and different mitigation systems will be feasible for different designs.

Within the EC framework programme, there have been generic experiments and theoretical studies on hydrogen combustion (deflagration and detonation), thermal — hydraulics, stratification and natural convection. Also included in the programme has been investigation of dynamic concrete behaviour at high impact velocity and studies of leakage of steam and aerosols through cracks and penetrations. These have been in conjunction with the identification of mitigation measures.

The dissipation of decay heat is accomplished in present generation water reactors via redundant and diverse emergency core cooling systems (ECCS). One approach in evolutionary reactor development, both ALWRs and AHWRs, is to utilise the best features of these present systems in an optimal way (Yadigaroglu et al., 1998), without significant recourse to new passive systems. Reactors based on this approach employ:

— improved system design with more redundancy, separation and diversity;

— increased pressure vessel water inventory;

— increased volume of pressuriser;

— direct in-vessel injection of cooling water;

— design features to reduce the risk of a LOCA, e. g. elimination of primary circuit piping;

— improved containment water storage tank facilities;

— introduction of cavity water flooding facilities

— automatic depressurisation of primary system followed by low pressure safety injection; and

— utilisation of a fire water system for containment sprays.

Plants in this class include EPR, ABWR, BWR 90, System 80 + and KNGR. There are in addition some CANDU and VVER designs. Table 11.1 summarises a few of the design highlights of these reactor types, which have been developed from optimisation of the best features of present generation plant.

Table 11.1. Classical evolutionary water reactor systems

Reactor Description

EPR Improved decay heat removal via active

systems, e. g. ECCS

System 80+

KNGR Greater redundancy, diversity, independence,

and separation of safety systems

ABWR

BWR90

VVER-1000

CANDU 6, 9 Improved containment cooling systems

Yadigaroglu et al. (1998).

12.6.2.1 BREST-300. Lead cooled reactor systems are under study at the Institute of Physics and Power Engineering (IPPE) and the Kurchatov Institute (IAEA-TEC- DOC-1289, 2002).

In the BREST-300 designs, developed by RDIPE and Kurchatov there is a two circuit design, there are four parallel loops including pumped lead flow removing heat from the reactor core. Lead inlet and outlet temperatures are 420 and 540°C, respectively. The design is integral with a supercritical pressure (24.5 MPa) steam water cycle. The uranium and plutonium nitride fuel implies low moderation and absorption of neutrons hence it is possible to achieve a core breeding ratio equal to one.

The BREST-300 reactor has various safety features such as negative void temperature coefficient; it operates with a breeding ratio of near unity with consequently minimal excess reactivity and there are no soluble poisons in the reactor coolant (IEA/OECD (NEA)/IAEA, 2002). Regarding the coolant, there is decay heat removal by passive systems, increased reactor coolant inertia, and the system pressure is low.

It has good thermodynamic efficiency due to high core outlet temperature, reduced number of components in the nuclear steam plant and reduced containment design requirements. The relatively small size implies reduced capital cost and this together with increased core outlet temperature means that the plant is also applicable to process heat applications.

There are, however, some penalties in using lead. There is a much greater pressure drop (about 7 times greater than sodium) across the core for otherwise similar conditions of reactor power, coolant flow cross section area in the core and fuel element length. This is caused by the lower thermal capacity of lead compared with sodium. The higher density of lead compared with sodium does not compensate. Lead cooled reactors therefore need to have a reduction in fuel fraction and increase in core diameter to reduce the hydraulic resistance. This implies that the core dimensions of the BREST reactors are large.

12.6.2.2 BREST-600. The plant has been scaled up to 600 MWe by RDIPE in co-operation with RRC Kurchatov (IAEA-TECDOC-1289, 2002). The characteristics of BREST-300 and BREST-600 are similar.

There is also an active programme on lead cooled reactors in Japan.

12.6.2.3 LCFR. Design studies of lead cooled fast reactors (LFRs) with nitride have been performed by the Japanese (IAEA-TECDOC-1289, 2002) as part of their programme to improve uranium resource utilisation and for the transmutation of high-level waste nuclides. The Japanese studied the impact of plant size on seismic issues and ways of developing more compacted and integrated plant designs.

The LCFR has negative void reactivity but a high breeding ratio of 1.26. The design is integral with the core, support structure and primary heat exchange systems situated within the reactor vessel. On the secondary side, the once through steam generator and its helical tubes are situated around the core and core diagrid. Regarding safety characteristics, the design is to reduce the propensity for lead-steam interaction.

12.6.2.4 General. Lead cooled systems have both advantages and disadvantages compared with sodium systems. Lead is much less reactive with air and water compared with sodium. In terms of other advantages in regard to minimum reactivity excess, transmutation of old actinides and fission products, proliferation safeguards consider­ations, safety in accident situations and economic competitiveness, lead and sodium systems have comparable properties.

There are however some negative aspects of lead associated with its corrosiveness; it may freeze in the steam generator in the case of feed-water heater failure. Repair and maintenance and remote re-fuelling operations are carried out at high lead temperatures of over 400°C. There is a potential for fuel subassembly blockage caused by lead/water/steam interactions.

A number of small medium-size nuclear power plants have been developed by RDIPE in Russia for district heating which can also be used for seawater desalination.

RUTA-TE is a pool-type thermal reactor that can be used for the cogeneration of electricity and power. The RUTA concept has already been described earlier in the book and more details are given in (RDIPE, 1994; Grechko et al., 1998). It can be used in conjunction with an RO process or together with a distillation process. The latter is expected to be the more economic with this power source.

14.4.1 NHR-200/Desalination Plant

This concept is being considered by the Institute of Nuclear Energy, Tsinghua University, Beijing, China (Zhang et al., 1998). The NHR-200 is an integral light water reactor

(introduced in the previous section). The proposed options for the desalination system are based on the steam generator and MED process for water production as a single process or with a steam generator and MED process for co-generation of water and electricity (Duo et al., 1995).

The main areas for additional research on in-vessel related response, following a core melt event, relate to the timing and influence of reflood, both early and late (Krugmann, 2001). There are uncertainties in the late core degradation mechanisms and how these affect the pressure vessel failure mode. There have been experimental programmes at Sandia

National Laboratories, USA, the Paul Scherrer Institute (PSI), Switzerland and the Royal Institute of Technology (RIT), Sweden, addressing these issues. Particular EC projects in relation to vessel failure are the EC funded ARVI project and the FOREVER experiments at RIT. It has been shown that the pressure vessel failure mode impacts the integrity of the vessel supports, corium dispersal, missile generation and direct containment heating risk. In regard to outstanding issues, there is a research need to consider the hydrogen production rate in the event of delayed depressurisation as this impacts the hydrogen management control system. The composition and temperature of the gas discharge will depend on the response of the primary system discharge valves. At high pressure, the integrity of the SG tubes may also be an issue.

Activities currently in progress within the EC 5th Framework Programme include the following.

The core loss during a severe accident (COLOSS) (Adroguer et al., 2001) programme considers various issues concerning core degradation phenomenology. For both PWR and VVER rods, it includes the impact of UO2 and ZrO2 dissolution by molten Zircaloy on core geometry degradation. The objective is to examine the consequences on hydrogen production, melt generation and the source term. It also addresses how the burn-up effect affects the dissolution of UO2 and MOX fuel by molten Zircaloy for PWR rods.

The experimental programme considers how the oxidation of U-O-Zr mixtures contributes to the peak hydrogen production during core reflood. Separate effects tests are carried out using a number of different composition U-O-Zr alloys. The results show that the oxidation of mixtures contributes to significant hydrogen release during degraded core quench.

Several large-scale tests are included to examine the B4C effects, from absorber rods, on core degradation and melt progression. These include a large-scale VVER-1000 bundle test with a central B4C rod, carried out in AEKI, Hungary, and a similar test with a B4C rod carried out at FZK, Karlsruhe in Germany. Results show large escalation of oxidation and hydrogen during the final steam cooling phase, this phenomenon had not previously been observed.

The programme is also examining whether the oxidation of B4C rods can induce volatile organic iodine production.

Some of these issues have been examined in earlier EC 4th FP projects, CIT (Adroguer et al., 1999) and COBE (Shepherd et al., 1999), which, respectively, were concerned with core material interactions and quench effects during core degradation.

Within the NEA collaborative programme, the MASCA (NEA Annual Report, 2002) project has also investigated the consequences of core melt within a severe accident. Experiments are being carried out in the Kurchatov Institute in which prototypical corium compositions are used. The experiments address the uncertainties on heat load to the reactor vessel and, therefore, on the uncertainties of vessel failure.

In this section, recent past in taken to infer the last decade.

There has generally been an improvement of performance at many plants. This has been evident from a number of performance measures, e. g. from WANO indexes and in the US, the Institute of Nuclear power Operations (INPO) (Sinco, 2003). This has been driven by better leadership and improved plant management.

Another driver for improved performance has been the move towards deregulation of the electricity industries in some countries, e. g. the US and the UK. This has resulted in competition in the electricity markets between all providers, nuclear and non-nuclear.

The last decade has seen the shutdown of some nuclear plants, for both safety and for economic reasons. For example, first generation VVER plants operating in former Eastern Germany were shutdown, following re-unification, because of safety concerns. On the other hand, business decisions on whether to shutdown some plants prematurely have depended on the scale of cost liabilities being carried.

There have thankfully been no major accidents over the past decade but there have been several incidents that have not helped the cause of the industry. The finding of boric acid corrosion in the reactor vessel head in the Davis-Besse plant has resulted in increased inspection, longer outages, etc. Although not on a reactor, the Tokai-mura incident in a fuel handling plant in Japan has also caused some concern.

Despite some of these more negative aspects, there has been a decade of safe and reliable operation. Building of new power plant in Asia has continued. Particularly in the last five years there has been an increase in confidence in some countries in which the industry was beginning to stagnate and the possibility of new build is now under consideration. This is true for Finland and France in Europe and also in the US.

J. N. Lillington

During the last century, nuclear power has been established as a reliable source of energy in the major industrialised countries. It has a potentially important role in the future since it does not contribute to the production of ‘Greenhouse’ gases; a growing concern of continued fossil fuel power generation. The time is now appropriate to review the issues surrounding the future operation of current generation nuclear reactors and consider the potential offered by the new advanced reactor designs that have been proposed for the new century. The main purpose of the book is to present in a single volume the main issues of future civil nuclear power plant operation including the justification and incentives for future continuation, safety considerations and existing national strategies. The survey covers the entire major designs and their associated research programmes.

The evolution of the civil nuclear energy programme has seen the development of different generations of nuclear plant. In the US, the different generations have been designated as follows:

I — early prototype reactors in the 1950s & 1960s;

II — commercial power reactors in the 1970s & 1980s;

III — advanced light water reactor designs developed and certified in the 1990s, and;

IV — future generation nuclear energy systems.

Although this terminology has been introduced mainly in the context of US designs, it will be used more generally in this book in referring to the different generations of reactor systems in question.

The first part of the book reviews the commercial plants currently in operation (Generation II) and focuses on the issues concerning the future operation of these plants.

In the main, nuclear power plants have operated very successfully since the 1950s. Water reactors are the predominant type in the world today, mainly pressurised but there is also a significant fraction of boiling and heavy water reactors. The UK is an exception, where gas reactors are predominant. A brief survey of present day reactors is given in the first chapter.

There are wide ranging issues associated with the future of nuclear power. There are also very different perceptions of the benefits compared with the risks. There have been only a very small number of significant accidents, e. g. Three Mile Island and Chernobyl but these have had a major impact in limiting the expansion of nuclear power. The safety of plants for all aspects of operation, including the management of waste, is a public concern that needs to be addressed.

By far the most important pre-requisite for the continued operation of nuclear power plants is that they should remain safe and reliable. Improved safety has resulted from extensive evaluations of the few accidents that have happened together with a general improvement in all aspects of plant management.

Operational efficiency and reliable performance must be achieved to ensure competitiveness in the world market. Operating margins are being optimised, subject to safety limits, to enable maximum power output. Outage times for maintenance and refuelling are being minimised to produce high load factors.

The drive for improved safety and reliability is leading to improved maintenance operations and better monitoring techniques. The contribution of reactor diagnostics, through noise analysis, to both the safety and performance of operating reactors is increasingly recognised. The development of this technology for application to future and/or advanced plants is considered later in the chapters on experimental and theoretical research.

Modernisation programmes are in progress to improve the safety and performance of the older plants in operation. Some of these activities are being carried out in support of life extension, if there is an economic incentive to extend the life of current plants subject to meeting safety constraints. Many of the older VVER reactors are also being modernised. These include replacement of components to improve station performance, but also the back fitting of safety systems, in some cases to extend the design basis accident envelope.

Improvement of the fuel cycle is an important area of current attention. Holistic approaches are being considered to reduce costs over the whole fuel cycle. There is an increasing trend towards the use of high burn-up fuel. Mixed Oxide (MOX) fuels are also loaded into some present day plant. More is likely to be loaded into future reactors, beneficial as a means of reducing plutonium stocks. Advanced fuel cycles based on a thorium cycle could also be considered in place or uranium and plutonium cycles.

Technical solutions have been put forward for the management of waste and spent fuel. The high level waste component is largely contained in temporary on-site storage and some action will need to be taken to ensure continued safe containment. Further, there is a significant number of plants reaching the end of life over the next decade. This will increase the volume of decommissioning activity and the volume of material waste that will need to be managed.

Advanced reactor issues are considered in the second half of the book.

Design objectives are discussed for advanced reactors. There has been a range of different approaches adopted in the development of new advanced reactor designs to simplify the design and hence reduce cost. Evolutionary designs are being proposed, which represent relatively small perturbations from current technology. Other more innovative or revolutionary designs are also being considered that are substantially different from existing technology and these require major development investment.

Advanced reactors will need to meet continued demands for increased safety. Regulatory issues in regard to the potential licensing of advanced plant will be covered in the book. There are likely to be moves towards more harmonised approaches in licensing, perhaps enabled by an increasing tendency from vendors to seek design certification, as has been the trend recently in the US.

There are significant differences in the rates of nuclear power expansion and contraction across the different ‘nuclear’ countries across the world. A snapshot will be provided of the current status of nuclear industry in these individual countries in respect to their position on potential ‘new build’ or otherwise, likely preferred reactor systems, regulatory and political climate etc.

By taking account of operating experience and safety evaluations of current generation reactors, new advanced reactor designs have been proposed which are competitive and more economic than existing designs, incorporating standardised and simpler components. Increased reliability of safety systems will be a further requirement; an objective in many designs is to include a high degree of inherent safety, by taking advantage of the natural forces, e. g. gravity and natural circulation.

Evolutionary water reactor designs (Generation III) are being designed against these objectives and are most likely to be chosen for any ‘new build’ initiatives at least in the short term (e. g. 2005-2015). The book summarises the most likely candidates in a separate chapter.

A major feature of many evolutionary water reactors is a much greater adherence to inherently (passive) safe design principles. Because of their importance in some water reactors, these principles are discussed in a separate chapter. Some of these principles are also a characteristic of more revolutionary water reactor designs. Inherently safe principles are also adopted in some other (non-water) reactors.

Future generation reactors are covered towards the end of the book. These include both medium (e. g. 2015-2025) and long term (2025 onwards) deployment options. The medium term options include some evolutionary designs from reactor systems that have already been prototyped e. g. high temperature reactors). The long term options encompass the Generation IV systems referred to above. A review is provided on these advanced designs including super-critical water reactors, high temperature thermal and gas cooled fast reactors, liquid metal cooled fast reactors (sodium and lead) and molten salt reactors. These reactor systems collectively provide a capability for a wide range of applications, including electricity generation, plutonium and actinide management, heat applications and hydrogen production.

A discussion of Accelerator Driven Systems (ADS) is included in the book. These provide an alternative to the future generation critical reactors described above since they can be used in similar applications. These utilise spallation neutrons, generated from a proton beam incident on a target, in conjunction with a sub-critical reactor. Designs are being considered for electricity applications and particularly for the incineration of plutonium and the transmutation of waste.

Nuclear heat applications reactors, other than for power generation, are also briefly reviewed in a separate chapter. Non-power producing reactors for low temperature applications such as district heating and desalination are already in operation. High temperature applications for hydrogen production and for the chemical and process plant industries are not yet developed commercially but are seen as potentially important in the future. The future generation reactors, referred to above, would be candidates for these applications.

Several chapters towards the end of the book describe the extensive research programmes (experimental and theoretical) that are currently in progress for the purposes of ensuring the safety and reliable operation of current plant and design certification and safety assessment of advanced plant. These include reference to the available published material from reactor vendors and utilities and the more widely available research published by research institutes. Much of present research focuses on present day plant but much is also relevant to the needs of new reactor design developments.

Thus the book considers the significant designs over the range of different advanced evolutionary reactors through to the more exotic reactor designs being proposed, including fluidised bed and burn-up wave type reactors.

The book concludes with a discussion of likely longer term future requirements of a more general nature. This includes such topics as anticipated future energy and electricity requirements. It describes how new nuclear power producing plant could meet the requirements. The book finishes with a brief summary of non-nuclear power options in relation to the projection of possible overall nuclear development strategies in the next few decades.

J. N. Lillington

3.1. INTRODUCTION/OBJECTIVES

This chapter addresses the issues of operational safety for existing nuclear power plants. It concerns safety throughout all aspects of power plant operation and a number of topics are covered. There are diverse issues, some of which have already been introduced in the previous chapter. The management of radioactive waste is a particular concern and must be resolved, although there is positive progress on this issue in some countries. It is now recognised that human factor considerations can play an important role in maintaining plant safety. Increasing attention is being paid to improved operator training and other means of reducing the risk of human error. Another goal is to achieve closer collaboration between regulators and utilities. The regulation of an increasing number of privatised utilities is a present day issue. Another topic included is how the experience of many reactor years of operation can be utilised to improve future safety.

Operational safety is of paramount importance for the continuation of nuclear power, both nationally and internationally. The Chernobyl accident demonstrated all too clearly that the consequences of a major accident cannot be confined within national boundaries. Further, were such an accident to occur in the future, the nuclear industry would be unlikely to survive in most countries. Regarding continuous improvement, additional safety systems have been back-fitted to some of the older operating reactors. Extended accident management procedures have also been developed. Both of these are intended to extend the plant safety envelope.