12.1. INTRODUCTION/OBJECTIVES

This chapter considers the innovative reactor designs that are being put forward as likely candidates for future generations of reactors. Higher efficiencies can be achieved for electricity generation by increasing the temperature of the reactor systems. Higher temperatures are a feature of many of the most promising future reactor designs.

There is also the potential of exploiting nuclear energy for more general energy applications than have been considered previously. These applications could be many and varied. They include the utilisation of fuel cycle systems to burn weapons grade plutonium or minor actinides from spent fuel. The possibility of utilising nuclear energy to generate hydrogen is an attractive option for transport. Efficient future reactors for electricity generation and for these additional applications are considered in this chapter.

Similar design requirements to those described for evolutionary plants in Chapter 7, relating to reliability, economics, safety and acceptability apply also to these type of systems, together with some additional requirements. General design requirements for these future reactors are described in this chapter.

It is worth noting at this stage that sub-critical reactors based on accelerator driven systems (ADS) are also attractive candidates for plutonium destruction and minor actinide conversion. These are considered separately in Chapter 13.

Some of the innovative reactor designs reviewed in this chapter are also being considered for heat applications. Heat and other novel applications for nuclear energy are considered in more detail in Chapter 14. In Chapter 12, the focus is on the innovative reactor designs, in Chapter 14 the focus is on novelties in the applications. Already in some countries, e. g. Russia, waste heat from electricity generators is being used for district heating. Most of the experience to date has been with low-temperature applications. Other low-temperature applications include desalination plants. In many cases, the proposed reactor designs and certainly already operating systems are based on established reactor designs; the novel aspects relate to the balance of plant configurations to achieve the desired goals.

Many of the most promising future reactor designs have been examined by the US instigated Generation IV Forum (GIF) programme that started a little over 2 years ago. There are a number of signatories from among the major nuclear plant operating countries, 10 countries have joined, Argentina, Brazil, Canada, France, Japan, South Africa, South Korea, Switzerland, UK and the US. Other European countries are participating through the EU, which is also a member.

The objective of Generation IV is to identify the most promising types of reactor design that will contribute to future generations of reactors and to put in place R&D to promote further understanding of the designs and their performance.

Initially over 100 different designs were considered under the simple title of future energy systems (not just nuclear). These were reduced to 19 designs and finally to the following 6 most promising designs, see Table 12.1.

There has also been a ‘Three Agency Study’ carried out by the International Energy Agency (IEA), the OECD Nuclear Energy Agency (NEA) and the International Atomic Energy Agency (IAEA) (IEA/OECD (NEA)/IAEA, 2002). There were 34 innovative designs considered. Of these, a total of 12 designs have been considered in some detail. Most of these are also included in the Table 12.1 categorisations.

13.8.1 BNL

There are three types of ADS under study at BNL (Takahashi, 1997).

The accelerator driven energy producer (ADEP) (Bonnaue et al., 1986) is intended for energy production, incineration of MA and LLFP. This concept uses a small power accelerator similar to that of a segmented cyclotron.

The concept is close to that of the conventional Pu fuelled fast reactor, but is run in slightly sub-critical conditions of keff equal to 0.98-0.99. The cyclotron with a few mA current and 3 GeV energy protons supplies a small spallation source.

The fuel in the ADEP core region is 9Pu + U, and MA in metal and oxide forms.

The reactor has thorium oxide in the blanket region. For transmuting the LLFP such as 99Tc and 129I (by neutron capture), the moderator region is installed between the outer core and blanket. To increase the production of 233U, the moderator region can be fuelled with thorium oxide.

The second approach, known as the Phoenix concept (Van Tuyle et al., 1993) (Figure 13.1) has the purpose of transmuting large amounts of MAs and LLFPs. It is based on modules of accelerator driven sub-critical lattices containing minor actinide fuel. From 1-8 modules serve as a target for an expanded proton beam of power 104 mA of 1.6 GeV protons. Each module of the core has a keff of 0.9 and is based on the fast flux test facility (FFTF) approach, e. g. oxide fuel elements and sodium cooling; with this specification the power generated would be 3600 MW.

The third concept is the accelerator driven particle fuel transmutator (ADPF) (Takahashi, 1990) which also transmutes MAs and LLFPs but at higher rate. This is achieved by means of a high neutron flux via the use of particle fuel.

Particle fuel can be used to generate high thermal power densities, because of its large heat transfer area (Takahashi, 1990). Helium is taken as the coolant because of the data available from conventional HTR technology.

The utilisation of nuclear heating for application to various oil refinery processes is being considered in Russia. Thermal power is required at different medium-range temperatures for different operations. These include low-temperature processes up to 400°C associated with initial reprocessing of oil products, e. g. hydrocracking, hydrocleaning. There are also middle-range temperature processes up to 600°C associated with secondary oil refining processes, reforming, cracking, etc. The VGM-P, HTR-10 and BN-600 systems, described below, are seen as possible candidates for this application.

The need for research into suitable materials for components that need to withstand the aggressive high temperature and corrosive environments of future generation plants has been mentioned earlier.

Key considerations are economics and how to reduce construction cost without compromising safety. Some particular areas in HTRs where cost savings could be made included the following. The heat recovery in high temperature gas designs incorporates a large tube-in-shell heat exchanger to recover helium heat discharged from the turbine, before it is recycled. The inclusion of an advanced plate type heat exchanger would result in much reduced size and, therefore cost.

Another area is the load imposed on the helium turbine bearings. Such large turbines have not been built or operated. One option might be to increase the speed of the power turbine, thus reducing its weight and size, and therefore possibly allowing the use of gas bearings.

Much work has been done of passive devices for innovative reactors. There needs to be a better understanding of the limits of the safety devices used on present day operating reactors. The issue is how to extend the existing devices to innovative reactor applications or, if necessary, how to develop new ones.

This chapter examines the incentives for future energy production from nuclear (carbon free) power generation in general. The potential economic incentives for the continued operation of current generation plant are also considered. These depend on whether the costs of maintaining and renewing the plant licence (which will in general increase with life) and other generation costs remain acceptable, compared with the revenue earned by the plant and perhaps other economic factors. Other broader incentives, e. g. environmental benefits are common to both continued operation of existing plant and the building of new plant. These are considered in more detail below.

World energy supply is dominated by fossil fuels (Table 2.1). It is generally accepted around the world that there is a need to reduce the emissions of greenhouse gases from the

burning of fossil fuels. These can be reduced somewhat by increased dependence on renewables and by energy savings, but a continued or possibly increased dependence on nuclear power is likely to be the only credible option to achieve the limitations in greenhouse emissions that are thought to be necessary.

There is general agreement that there will be an increase in the world’s requirement for electricity over the next few decades. The World Energy Council (WEC) (Blix, 1998) predicts that the expansion will increase by 50-70% between 1990 and 2020. The drivers are increase in world population, expansion of industry and improvements in standard of living particularly in the developing countries, e. g. Asia.

The present trend towards meeting this demand includes the building of fossil fuel plants, particularly combined cycle gas fired (CCGF) plants. There are at present no orders for new nuclear reactors in Europe or North America (Finland may place an order in the near future). There is some limited completion of plants in Eastern Europe. There are still a number of new reactors under construction in Eastern Asia. The consequences of this ‘little or no nuclear build’ strategy are increasingly greater emissions of carbon gases, together with other gases associated with the burning of fossil fuels (e. g. sulphur oxides).

The spiralling increase in greenhouse gas emissions has resulted in the setting of targets for many of the individual industrialised countries and international bodies concerned with nuclear energy. Although sound in principle, this approach has met with only limited success. Targets were set in Toronto (1988) to reduce emissions of carbon dioxide by 20% by 2005, in Rio (1992) to return to 1990 levels by 2000, by the UN General Assembly (1998) to achieve a 15% reduction of greenhouse gases by 2010 compared with 1990; a means of achieving constraints was put forward at the Kyoto conference (1997). In practice, however, emissions have significantly increased. From 1988 to 1998, carbon dioxide emissions have increased globally by about 16%. IAEA predict that emissions will be 36-50% higher by 2010 compared with 1990. Figure 2.1 (Energy Visions 2030 for Finland, 2003) shows past and projected carbon emissions in the industrialised and

developing countries for a future scenario based on a relatively robust market development with a fossil fuel-based economy. The Kyoto protocol limit (indicated by a dotted line and applied here to the CO2 from fossil fuels only) is also shown.

Ways have been proposed to reduce these increases by directly reducing the quantity of greenhouse gases produced, by such means as increased efficiency, via national economic constraints or by the setting of global limits that define national quotas, etc. Renewable sources of energy should not be ignored but there are technical limitations on the scales of operation that might be required, e. g. the size of wind farms. There are also issues of reliability and transmission; the wind does not blow every day and the power may be generated in remote areas or out at sea. Finally, new technologies have been proposed to convert harmful flue gases such as sulphur and nitrogen oxides to ammonium salts, by adding ammonia to flue gases and then irradiating with an electron beam produced by a nuclear accelerator. These techniques though do not apply to the carbon dioxide emissions associated with the burning of fossil fuels. Other long-term solutions, e. g. hydrogen and fusion do not provide viable alternatives on a timescale of the next few decades.

Many commentators, therefore, feel that the only viable alternative to fossil fuels is nuclear energy to reduce the rate of increase of greenhouse gases, particularly carbon dioxide.

Another incentive for nuclear power is to maintain diversity of supply. A national strategy limited to one particular form of energy (fuel) will be vulnerable to reductions of other fuel costs.

There are differences in view on the economic competitiveness of nuclear electricity compared with other fuels. Clearly, there are significant uncertainties in future costs, looking forward over a timescale of the life of a plant (at least several decades).

It has been recognised over many years that the ‘defence-in-depth’ principle is fundamental to the design of nuclear power reactors and other types of nuclear plant (Table 3.10). The important feature is the requirement that multiple barriers exist against the release of radioactivity to the environment. The defence-in-depth principle is generally assessed using either or both deterministic or probabilistic methods.

Various means of strengthening the defence-in-depth principle are being considered in current generation reactors and indeed implemented, with respect to accident prevention

(Hogberg, 1998). Additional levels of protection have been installed in many European and other reactors worldwide. In particular measures have been taken in a number of countries to improve the capability of existing components to withstand severe accident loads. The main objective is to mitigate the release of radioactive isotopes to the environment, particularly iodine and caesium. These measures have been complemented with the development of severe accident strategy improvements.

Clearly there are economic and technical constraints on back-fitting improvements in existing reactors. There are many types of design in operation and the feasibility of such improvements is design specific. Nevertheless significant improvements have been achieved at acceptable cost. Many of the desired measures have been identified in Periodic Safety Reviews, which are now a common-place regulatory requirement in most countries. They are being introduced within modernisation programmes, which may also be in place for other reasons, e. g. to replace out-of-date systems or instrumentation that has become too costly to maintain. There may be a requirement to improve the older operating plants to a standard commensurate with later models. If this is not achieved, it may be necessary to shut the older plants down.

It has been realised for many years that the defence-in-depth in many of the earlier Russian designed reactors only applies to a much more limited design basis than Western reactors. The safety of VVER and RBMK reactors has been extensively studied in a number of international projects over the last decade. Numerous safety recommendations have been made, including back-fitting of safety systems, etc. Some of these plants are operating in the EU Enlargement Countries, which will be joining the EU over the next few years. There is, therefore, a driver to accelerate the safety improvement process.

Table 3.11. Examples of back-fits on current plants

Availability of additional water-delivery sources Filtered venting

Hydrogen control with ignitors and catalytic recombiners

Improved safety valves

Reinforcement of containment penetrations

Sehgal.

A number of safety improvements have been recommended for the early VVER-440 designs in respect of control of the reactor pressure vessel embrittlement, improved emergency core cooling systems and, improved reliability of residual heat removal systems. Additionally there are recommendations for improved instrumentation and control systems, including the reactor protection and shutdown systems and improved capability of the confinement to limit radioactive releases. Safety improvement programmes are underway to address these concerns.

There were greater drives for immediate safety improvements of RBMK designed reactors in the wake of the Chernobyl accident. Some of the early plants have now been shutdown but a number of safety improvements have been implemented in the newer RBMK reactors still in operation.

For example the Ignalina power plant in Lithuania (an EU Candidate country) has recently undergone international peer review and various short-term safety improvements have been recommended. These relate to control and protection system reliability, the structural integrity of the major primary circuit components and the confinement function, improved emergency operating procedures, and the need to address fire hazards that could impact safety systems.

The large amount of severe accident phenomenological research carried out for Western water reactors has led to various mitigation measures being introduced and back-fits to be implemented (Table 3.11). The safety of current generation plants has been substantially improved by the development of this knowledge base. Containment research has been supported by experimental programmes on the removal of aerosols with sprays and on the modes of hydrogen combustion using igniters. The results of research for present day reactors are also benefiting the designs for future plants. Many of these plants include severe accident mitigation concepts in their design. Advanced designs include measures for improved in-vessel coolability of debris and ex-vessel debris coolability and retention. These are discussed in the subsequent chapters.

6.2.1 Scale of the Problem

The quantity of radioactive waste produced from all sources is a very small fraction of the overall waste produced. In France for example, which has the highest fraction of its power generated by nuclear power, about 84,000 t of the 650,000,000 t of total waste produced annually is radioactive (Rosen, 1999) The latter figures include 200,000,0001 of hazardous industrial waste, yielding a percentage of radioactive waste in the hazardous waste of 0.015%. US figures are comparable.

The solid wastes produced from diverse energy sources for a 1000 MWe power plant are shown in Figure 6.1. A coal plant produces annually about 320,000 t of ash, containing about 4001 of hazardous heavy metals such as vanadium, mercury, and others. Additionally without abatement, a further 44,000 t of sulphur oxides and 22,000 t of

nitrous oxides go into the atmosphere and further waste is produced from mining and transportation. By comparison, a corresponding nuclear plant produces annually about 30 t of high-level waste (spent fuel) and about 800 t of intermediate — and low-level waste with virtually no release of noxious or greenhouse gases. Additionally, the waste quantities for fossil power plants are significantly increased by modern abatement techniques; e. g. sulphur abatement procedures for coal plants produce about 500,000 t of solid waste.

The management of radioactive waste is largely through confinement, since the quantities are extremely small. This contrasts the approach for large quantities of other toxic waste, which are dispersed in the environment to a level that is considered to be safe. Because of the large quantities involved, this is the only practical solution, yet clearly there may be safety concerns with this strategy.

Radioactive waste is typically characterised at three levels, low, medium and high. The levels of activity are categorised in different ways, but generally low level waste is deemed to be at a sufficiently low level of activity that shielding is not necessary apart from simple protective measures for handling. At intermediate level, shielding would be required; at high level, thick shielding and certainly remote handling facilities would be necessary.

For the purposes of waste disposal, the timescale of decay of the various isotopes will be an important factor, determining the time of confinement, and the facilities that are required for confinement.

Radioactive waste can come from many sources in the modern world. Most intermediate — and all high-level waste arises from civil nuclear power and military operations. In nuclear power activities, such waste arises from all stages of the fuel cycle; the significant waste problems arise from spent fuel and in waste from reprocessing operations.

Table 6.1 indicates the typical quantities and levels of waste arising from a 1000 MWe nuclear power plant.

In terms of worldwide production, the total volume of low-level waste is ~ 100,000 m3 per annum, compared with about 4000 m3 per annum of high-level waste.

The present strategies for waste management depend on the relative levels of activity.

Low-level waste is usually stored in steel drums and disposed of in surface trenches above the local groundwater level. Since many of the isotopes in low-level waste have half-lives of only a few decades, the timescale for the waste to no longer pose a radiological hazard may be the order of only 100 years. The containment has to be sufficiently robust to resist corrosion and leaching of material for only a relatively short period.

Intermediate waste is encased in cement, inside steel drums. These are disposed of in relatively near surface repositories in a number of countries. Many repositories are already in operation and further facilities are expected in the future.

For high-level waste, an initial period is required to allow some decrease of its radioactivity and for residual heat to dissipate, before it is practical to consider long­term storage. Storage of spent fuel is usually under water initially at the site of production; in the longer term, dry storage may be possible. There is an intention in many countries to store high-level waste in deep underground repositories but as stated earlier this is not yet realised in most countries (Finland has recently given permission). To reduce volume, there is also the intention to vitrify high-level liquid waste to facilitate storage before disposal.

The status of selective national regulatory frameworks in relation to the design and safety of future NPPs is reviewed below (IAEA-TECDOC-905, 1996).

8.4.1 UK

In the UK for example (EUR 20055 EN, 2001), safety is governed by the Nuclear Installations Acts 1965 and 1969 (NII Acts) and by the ‘Health and Safety at Work Act 1974 (HSW Act)’. These are supplemented by the Nuclear Installation Regulations, the Ionising Radiation Regulations and other Licensing Conditions. Regulatory Guides (non­mandatory) include the Tolerability of Risk (TOR) for Nuclear Power Stations (HSE, 1992) and the Safety Assessment Principles (SAPs) (Harbison, 1992). These latter two documents are to provide guidance to NII Assessors in assessing Licensees’ safety cases but are not legislative.

The Health and Safety Commission (HSC) is responsible for preparing proposals for safety laws and standards approved by the Secretary of State for Environment (DOE). It advises the Department of Trade and Industry (DTI) regarding regulatory matters in England and Wales and the Secretary of State for Scotland. The HSC is advised by the

Advisory Committee on Nuclear Installations (ACSNI) and the National Radiological Protection Board (NRPB). The HSW Act is enforced by the independent UK Government Health and Safety Executive (HSE), under the HSC. The HSE is responsible for granting nuclear licences and the enforcement of the Health and Safety Laws. Licences and Inspections are administered by the UK NII who have the authority to withhold licences for nuclear plant operation.

6.3.3 Armenia

Armenia has one early VVER-270 unit operable.

6.3.4 Bulgaria

Until recently, Bulgaria had four VVER-440/230 units operating at Kozloduy generating 47% of the country’s electricity in 2002 (International Atomic Energy Agency, 2002). Units 1 & 2 closed at the end of December 2002 (Foratom e-Bulletin, 2003a). Units 3 & 4 are continuing in operation, due to a pending government decision on their safety, following the outcome of several international reviews (including European Commission and WANO) (Foratom e-Bulletin, 2003e). Further upgrading of Units 5 & 6 is likely to be carried out (World Nuclear Association, 2003).

There are plans for the construction of the Belene nuclear plant to be resumed in 2004, starting up by 2008. A feasibility study is ongoing in 2003. Results are expected in October 2003 after which time the decision to resume or otherwise will be taken (International Atomic Energy Agency, 2002).

The GCFR programme had taken place in the US from early in the country’s history on nuclear power. It was considered in parallel to the US LMFBR programme, being perceived to offer a number of advantages. The conceptual GCFR was in principle simpler to operate compared with the sodium-cooled LMFBR and, if required, has the potential for higher breeding gain. Considerable experience had also accrued from the operation of the HTR Peach Bottom and Fort St Vrain reactors.

Regarding the fuel and core design, the GCFR was based on the Liquid Metal Fast Reactor (LMFR) design, incorporating niobium stabilised stainless steel pins and wrappers. The coolant was helium and the pin design was based on pressure equalised vented pins. This has the advantage for pin design and performance at the expense of a removal of a fission product release barrier. Reactivity was controlled by two independent and diverse shutdown systems.

The primary heat removal system was an upward flow system through the core, driven by active circulators. The design also included an independent and redundant decay heat removal system. The reactor vessel was engineered from pre-stressed concrete. On the helium side, it was insulated and there was water cooling on the outside.

The containment building incorporated a molten fuel containment system, below the bottom of the vessel.

The GCFR programme was halted by the USDOE in 1981, because Liquid Metal Fast Breeder Reactor (LMFBR) technology had progressed sufficiently to become a credible option. The GCFR had no proliferation advantages over the LMFBR design. There were also some economic, safety and technical factors that affected the decision to go forward. The multi-cavity pre-stressed concrete posed problems for manufacture and inspection and also problems for extrapolation. From a safety perspective, vented pins implied that the first barrier for fission product release was lost. Finally, there were concerns over spent fuel assembly cooling. Nevertheless, the initial work over the first few decades was viewed as providing a positive story in terms of technical design development and the safety and licensing activities that were carried out.