The economic case will depend on many factors. The principles for estimating the lifetime extension costs are not different from those used for assessing the costs of plant upgrades, design changes, plant decommissioning or new construction. However, there may be some particular site-dependent factors that need to be taken into consideration.

Plants built during the early days of nuclear power production tended to be diverse; as lessons were learned, designs were improved, and unit capacities were increased. Thus, applications and therefore costs for life extension for early designs are likely to be design specific. This will be less so in the future as later current generation plants reach their design end of life and are considered for life extension.

Even for a given reactor design, there may be significant technical differences in the state of the important components. An obvious example is the state of the reactor pressure vessel. This will largely depend on the integrated neutron fluence experienced during its design life. This will be affected by the power level at which the plant has been operated, the effectiveness of the vessel protection radiation shield, whether the vessel has been annealed, etc. In short, the operating history will be needed to assess the state of the vessel.

Measures may be taken to upgrade certain components during initial planned life. These may be upgrades to improve performance, improve safety, or to improve or mitigate the effects of ageing. For example, steam generator replacement has been performed on a number of plants to improve reliability of operation. Component replacement during normal design life will clearly be beneficial to improving the chances of life extension for the particular plant. Further, higher costs incurred during normal operation could well mean reduced costs for life extension.

Other factors identified in IAEA-TECDOC-1084 (1999) include differences in costs due to differences in regulations between countries, resulting in differences in costs in the safety cases presented by the utilities. Another factor in assessing the comparative economics of lifetime extension costs rests with the proprietary nature of plant data. Utilities are reluctant to release information that might benefit another competitive utility.

Generic cost data for US plants have been published in IAEA-TECDOC-1084 (1999). These show lifetime extension costs relative to the building of new nuclear plant compared

Table 2.10. Lifetime extension versus new building costs ($US per kWe)

Lifetime extension 210- 840 Based on Surry 1 (PWR) and

Monticello (BWR)

New nuclear plants 0 2000 Building costs only

New combined cycle units 700-900 Building costs only

Data from IAEA-TECDOC-1084 (1999).

with combined cycle plants. There are considerable uncertainties on life extension costs but the conclusion is favourable for life extension at least on the basis of building costs of new plant (Table 2.10).

Also in IAEA-TECDOC-1084 (1999), an attempt is made to identify site/plant specific factors that have most influence on the cost and tend to push the cost estimates to either the higher pessimistic or to the lower optimistic figure.

With regard to plant design, steam generator replacement and reactor pressure vessel annealing were key factors for PWRs while replacement of pipes and reactor pressure vessel internals were the key factors for BWRs. For reasons explained earlier, newer plants would be expected to incur lower costs than older plants. However, this is somewhat obscured since the former will tend to have longer design life than the latter.

The schedule for implementing the various measures is important. If the intention is to continue the extended operation immediately beyond the end of design life, it is advantageous to begin lifetime extension measures during earlier scheduled outages. Other costs identified were the costs of replacement power to meet the demand during the intervening period. Such costs are clearly highly power system specific.

Finally, costs to meet the demands of the regulator and also possibly to overcome the concerns of the public also need to be factored into the cost balance.

It is concluded that plant-specific costs be required in order to make realistic cost estimations. To evaluate the competitiveness of life extension options, it will be necessary to compare with other power-producing options, including both nuclear and non-nuclear. However, while not possible to produce a generic economic case for life extension, it is clear that a number of utilities have addressed the issue for their own plants and have come out in favour of the benefits of life extension.

This section provides some examples of recent programmes that are being implemented to improve performance and reliability across a span of the reactors that are currently opera­ting, see IAEA-TECDOC-1175 (2000). In some cases, the improvements are also being incorporated for safety reasons; the resulting better performance is an additional bonus.

In Central Europe, various safety upgrades and reliability improvements have been made on the earlier VVER-440/230 reactors that are still operating. For example, in the Bohunice Units 1 and 2 in Slovakia, the emergency core cooling systems and electrical systems have been reconstructed to achieve better separation redundancy and independence. On each plant, the instrumentation and control (I & C) system has also been reconstructed and an emergency feed water system (EFWS) has been added. Other significant improvements include annealing of the reactor pressure vessels and better seismic qualification.

The improvements for these earlier VVER-440/230s are to ensure safe and economic operation for only a relatively short period of operation; most will be decommissioned within the next few years. Improvements are being carried out on the newer VVER — 440/213 reactors to ensure operation for one or possibly two decades into the future.

There have been major modernisation and upgrading programmes on the Dukovany plants in the Czech Republic. Achievement of improved economics by increasing the power rating of each unit by as much as 20 MWe may be possible via an improved evaluation of the operating margins. Modernisation activities include better fire protection, improved I & C systems, modification of the EFWS and better hydrogen control under accident conditions. The Paks plants in Hungary are undergoing similar enhancements.

In Japan, there has been an active programme to establish the necessary inspection and maintenance activities to be done as countermeasures against ageing. There are ambitious targets to extend the life of some of the older plants out to 60 years.

The most modern plants are already incorporating technical features in their design for better practice, improved operation and better maintenance. The latest advanced boiling water reactor (ABWR) plants (Kashiwazaki-Kariwa Units 6 & 7) that entered operation in Japan in the last decade have the largest capacity, yet shortest outage times. This is seen to be due to national component testing programmes to verify their performance for Japanese ABWR operation, even if previous international experience exists elsewhere. Further, during the initial outages, there were rigorous overhauls and inspections of new design features (reactor pumps, advanced control drive mechanisms, high-efficiency steam turbines). There have also been full-scale training programmes for reactor maintenance staff.

7.1. INTRODUCTION/OBJECTIVES

This chapter describes advanced reactor design requirements and the status of international activities. The assumption is made that nuclear power will continue to provide a reliable and sustainable energy source, while complementing that produced from other technologies, e. g. fossil fuel, renewables, etc. It discusses general design objectives, primarily from a utility and vendor requirements perspective. Advanced reactors are often classified into two categories namely, ‘evolutionary’ and ‘innovative’. In this context ‘evolutionary reactor’ refers to the class of reactors with relatively small modifications from existing designs. By contrast, ‘innovative reactors’ incorporate substantially new designs, which would require significant investment to develop. Potential regulator requirements for advanced plants are considered in the next chapter.

The primary raison d’etre behind the design of current generation plants was that they should be able to provide a reliable and safe base-load electricity supply. The same requirement holds true today but with increased emphasis on economic viability, increased safety characteristics and improved public acceptance. These considerations are paramount in advanced reactor design specification. The main focus of this chapter is to describe international developments in design philosophy in advanced evolutionary reactors. The characteristics of a number of more innovative advanced reactor designs that are being proposed, including design requirements, are considered in more detail in the subsequent chapters.

1.5.1 Existing Plants

Many LWRs operating in the Western world were designed according to US safety criteria and philosophy based on the defence-in-depth principle in design. This also includes the construction, maintenance and inspection and operational practices that were developed according to the US model. Some countries introduced country-specific regulations, e. g. associated with higher density populations or the requirement to withstand military aircraft crash, i. e. as in Germany. Nevertheless, the US historical influence has tended to encourage a process of harmonisation in the regulation of LWRs.

Within the Western world, there has always existed openness in communication at the level of research. This has had the result that significant differences in practices have been discovered. Regulators have been informed and the most advantageous common approaches adopted in many cases.

There are some areas where there has been a smaller level of harmonisation due to differences in safety philosophy because of redundancy requirements, levels of conservatism, etc. It is also difficult to develop a harmonised approach to safety criteria because of differences in plant design. Two particular areas are in the fields of fuel safety criteria and PSA.

In general, there is a greater degree of harmonisation in operational safety, in the requirements for Non-destructive Testing (NDT), on environmental qualification, the benefits of periodic safety reviews and the merits of risk based service inspection.

Some of the benefits of harmonisation for future plant are given below and also in Table 8.7.

1.5.2 Future Plants

There should be greater scope for a harmonised approach to licensing new designs. Already, this is happening for evolutionary reactors. Many of these have been designed against URs.

Table 8.7. Potential benefits of a harmonised approach to licensing

Achievement of a common licensing position across a number of countries would increase the common market

Common international standards for plant design would facilitate the licensing of plant

Harmonisation of design requirements, enabling design certification would benefit vendors

A harmonised regulatory approach would benefit utilities by reducing uncertainties in the licensing process

EUR 20055 EN (2001).

For example, designs have been specified by utility companies in Europe in consultation with regulatory authorities. A good example of this approach is the EPR French-German co-operation. Another co-operation involved the Westinghouse 1000 MWe passive plant reactor development programme, the Siemens 1000 MWe BWR and the Westinghouse Atom BWR90 +.

Harmonisation of EURs provides a focus for utilities and is more important than harmonisation in international working groups, which may be less focussed on utilities’ specific requirements.

Harmonisation can result from a design certification process, as adopted for AP600 and which is in progress for AP1000.

Another means to improve harmonisation would be in the common development of standards but this is not currently the situation.

Harmonisation of approach has resulted in increased consideration of severe accidents at the design stage.

In an intact circuit accident, the heat sink is no longer available, e. g. to the steam generator secondary side or to the turbine.

In the AP1000/600 designs, under accident conditions, heat is transferred to the in­containment refuelling water storage tank (IRWST) via a passive residual heat removal heat exchanger (PRHR HX). This is connected to the reactor cooling system forming a full pressure, closed, natural circulation cooling loop (Hochreiter, 1992).

In, for example, a loss of normal feedwater scenario, the PRHR can remove sufficient heat to prevent operation of the pressurizer safety valves. The PRHR HX is activated following reactor trip and loss of power. If the pumps are operating, the flow through the passive RHR heat exchanger will be forced convection from the higher pressure cold leg to the hot leg. However, if the pumps are not operating, the flow direction will be reversed and by natural circulation from the hot leg to the top of the PRHR heat exchanger to the cold leg.

The EP 1000 incorporates a similar system (Yadigaroglu et al., 1998).

Other designs are summarised in Yadigaroglu et al. (1998).

In the SWR 1000 design, there are Emergency Condensers connected to the RPV without valves and immersed in the core flooding pool.

In all the above cases, a further step is required to transfer heat from the pools to the ambient. These are described later in the containment section.

Other designs, e. g. KNGR Chang et al. (1997), some CANDU systems and some Siemens systems utilise cooling of the secondary side via a condenser.

The VVER-1000 and AC-600 systems make use of condensers outside the containment via a natural circulation air-cooled system.

Finally, the ESBWR and the Indian heavy water moderated light boiling water cooled AHWR utilise isolation condensers condensing steam from the RPV.

The passive cooling of the moderator in CANDU reactors employs a similar approach.

The USR reference 625 MWe design developed by ORNL builds on earlier experience from the laboratory (IEA/OECD (NEA)/IAEA, 2002).

12.7.2 MSR-NC

The MSR-NC 470 MWe reference design has been put forward by the Kurchatov Institute (RRC-KI) in Russia.

12.7.3 FUJI

The FUJI reactor developed by ITHMSO in Japan is a low-pressure vessel loop style reactor with a graphite moderator and a molten salt coolant. It built on the ORNL technology and has an electrical output of 100 MWe. It includes inherent and passive features in that no moderating materials are located near the reactor vessel. Thus the reactor cannot achieve criticality outside of the core in the event of molten salt leakage.

In MSRs such as FUJI, the fuel (uranium and thorium) is dissolved in the molten salt. The salt is 7LiF-BeF2 and it can contain fissile material, 233UF4, and fertile material, 232ThF4. The temperature reactivity coefficient is strongly negative with increasing temperature due to the presence of the graphite moderator and reduction in molten salt density.

The FUJI reactor has other important safety features, decay heat can be passed passively to the environment, on-line fuelling ensuring that reactivity is minimum at all times, the pressure is low and the vessel is designed against high fluence embrittlement. There are no soluble poisons in the molten salt coolant.

Economically, there is good thermodynamic efficiency because of the high core outlet temperature. This also makes the reactor a good candidate for combined heat and power applications compared with present generation water reactors. The reactor has a smaller number of components, reduced containment requirements and is of small size so capital costs are kept down. It has on-line refuelling so refuelling outages are eliminated.

Environmentally, this fuel cycle has some attractions; the presence of thorium implies that a smaller number of higher actinides are produced. Over the operating life of the reactor, the fuel is not removed, so no fission products are removed from the fuel/coolant. It also operates as a near breeder (breeding ratio near unity); thus uranium resource requirements are reduced.

There is a proposal to set up a nuclear desalination demonstration plant at the Madras atomic power station (MAPS) in India. It will have a capacity of 6300 m3 per day and would be based on the MSF and seawater RO processes (Hanra and Misra, 1998). The power would be drawn from a 200 MWe PHWR operating in co-generation mode.

Since the Rasmussen study in 1975, various potential containment failure modes giving rise to radioactive releases have been examined. Source terms have been identified at various stages according to the delay for containment failure and the potential for delayed release through some pathway with some possibility of retention, see for example WASH 1400 (1975).

In this case, the source terms have been classified into releases associated with:

— an early containment failure with a pathway for direct release;

— a delayed containment failure (24 h) with a pathway for direct release;

— a delayed release through a pathway including some radionuclide retention.

Experimental studies are being undertaken within the EC 5th Framework programme to quantify fission product and core materials released from molten corium during the late phase of a severe accident. This would be at a time when the integrity of the containment vessel might be threatened. This work has been carried out within the PHEBUS programme. A schematic of the facility is shown in Figure 15.2.

In addition to promoting understanding, another important objective of the PHEBUS programme is to provide well-instrumented data for the validation of integral severe accident computer codes. The main processes that effect the degradation of fuel

and control rods, the release of fission products and aerosols, their transport in the primary circuit and the source term to the containment, are all included within the scope of the experiments.

Generally a good understanding has been gained of the releases of the more volatile fission products from intact fuel. However, the database for the release of less volatile fission products, core materials, aerosols, etc. from a degraded core is much less complete. This is particularly true for releases from a molten core. Previous experiments (Benson et al., 1999) have partially improved the database for the behaviour of metallic and ceramic pools. Additional data are required on the effects of fission product release in sparging and on the formation of crusts.

A current EC experimental programme (Bechta et al., 2001) is underway to examine such behaviour of fission product release from metallic and oxidic melts. The experiments will provide more understanding of species chemistry during the late phase of an accident. There are also tests aimed at examining the long-term behaviour of previously liquefied melt with an overlying water pool.

Metallic melt experiments are providing a better understanding of the important mechanisms affecting fission product and core materials releases up to 2000°C. They cover

the effects of temperature, oxygen potential, sparging, slug formation, two phase pools and composition of melt.

Oxidic melt experiments are in progress, concerned with studying the volatilisation of uranium oxides, fission products and boron oxide from melts of different compositions of UO2/ZrO2/SiO2/FeOx. These experiments are for both air and inert atmospheres, and with different temperatures of corium.

The main chemistry interests of the work concern tellurium, ruthenium, barium and strontium and the influence of steam on the volatility of the refractory fission products and actinides. The experiments focus on the generation of these elements at both high temperature, 1000°C and low temperature 25 °C.

Immersed core experiments are being carried out to determine the leaching and suspension rates of solid melts immersed in water under prototypical accident conditions. These use prototypical materials composed of UO2 and ZrO2 and other oxides.

Electricity generation is by far the most important civil nuclear energy application. This is likely to remain the case in the future, although some additional applications are envisaged, as discussed below.

There are marked differences across the major industrialised sectors in regard to future trends for nuclear power electricity generation. In Asia, modest expansion can be expected, in Europe, Finland is preparing for new build, but other European countries, e. g. Belgium and Germany are pursuing phase out policies. Nuclear power potential is being reconsidered in the US. Table 17.2 shows a relatively pessimistic scenario for nuclear power whereby no new power plants are built, beyond those already being built or firmly planned, together with the retirement of old plants.

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. It is worth noting that some experts assert that the capital cost of modern nuclear plant is no higher than that of new coal plant. There are also predictions that the total cost of nuclear electricity of Generation IV reactors will be less than that of gas plant.

In order to improve on energy efficiency, there is likely to be increased interest in CHP. For example, in the UK, about 9 GW of nuclear plant will be decommissioned over the next two decades, and by 2010 the UK is planning to install about 10 GWe of CHP plant (http://www-tec. open. ac. uk/eeru/naatta/renewonline/rol39/11.htm). This is a commitment in the Energy White Paper (Energy White Paper, 2003). Currently heat produced in electricity generation is largely wasted. CHP plants could be made to produce heat as well as electricity in approximately equal proportions. Supporters of non-nuclear energy generation argue that the adoption of gas-fired CHP plants would release gas currently used for heating, for use in electricity generation without leading to increase in carbon emissions. However, if nuclear plant provides the CHP energy source, then carbon emissions are quantitatively reduced.

Boiling water reactors (BWRs) were first developed in the US by the General Electric Company. The first commercial BWR, Dresden, sold to the Commonwealth Edison Company, was a 200 MW plant commissioned in 1960. This was followed by subsequent orders in the US, Europe and Japan. Ratings were increased up to the 1300 MW plants in operation today. Other vendors developed designs, independently of the US, notably Asea — Atom, later ABB Atom, in Sweden, Figure 1.2.

The characteristic feature of BWRs compared with PWRs is that boiling occurs within the core. Due to the axially changing void fraction, the axial flux becomes asymmetric. After drying in moisture separators (as in a PWR), the steam is passed directly to the turbine. The loop is completed by condensing the steam; the condensate is then returned to the reactor vessel. The Forsmark 3 BWR loop is shown in Figure 1.2.

BWRs burn uranium oxide fuel at a typical enrichment of around 2%. Fuel rods are grouped in a square lattice of 6 X 6 up to 8 X rods, the full assembly being smaller than in the PWR. The enrichment within the rods depends on their position in the fuel assembly, the reason for this being to correct for the effects of water spaces between the fuel assemblies. Reactor control is achieved with control rods inserted from the bottom of the core. The absorber material in the rods is boron carbide.

There have been changes in the main recirculation systems employed in BWRs during their evolution. For example, in some of the older BWR designs, the water is circulated by external pumps, one pump on each loop external to the vessel. In the more recent designs, the tendency is to utilise internal pumps, to avoid the risk of loss of coolant in the event of an external line break. General Electric employed an intermediate system with both external and internal pumps. Reactor power can be controlled by altering the flowrate

since this affects the core water temperature and steam bubble level, thereby affecting the neutron moderation.

BWRs operate at a lower pressure than PWRs, typically 7-8 MPa. BWR vessels are generally larger than PWRs, which is a disadvantage, despite having the advantage of a single cycle system. The turbine area in a BWR has to be monitored to ensure that health physics regulations are satisfied. Radioactive products can be transported in the steam, from a failed fuel rod for example.

BWRs are constructed with a leak tight containment, which is designed to withstand the load from a large break in the coolant or steam system. Safety systems have the provision to inject water directly into the reactor vessel to cool the fuel. Containment pressure increase is relieved via condensation in water filled areas. There is an additional cooling system to spray the chamber surrounding the reactor vessel.

The design of containment has evolved through the years, mainly in relation to the designs of the dry well that surrounds the reactor and the wet well that contains the water for pressure suppression in the event of a reactor vessel penetration failure. For example, General Electric developed the Mark I, II and III design containments, the principal driver being to simplify design and increase capacity. Six different models, BWR 1 -6, have been developed, incorporating different pump configurations, increased fuel assembly arrays and power density.

BWRs exhibit many of the advantages and disadvantages associated with PWRs. They have also been operated very successfully over a long period of time and much experience in operation has been accumulated. They are fuelled off line, utilise similar fuel coolant and moderator and have relatively complex technology, albeit that the single cycle system of the BWR is clearly a simplification of the two loop cycle of the PWR (and hence capital costs tend to be somewhat lower). Comparative data for the PWR and BWR and also for the other reactor designs are given in Table 1.3.