The strategy in most decommissioning activities is to reduce the radiological hazard in a systematic way, until the delicensing condition for the site is reached. As noted earlier, after operations have ceased in a reactor, the removal of fuel reduces the hazard by a significant degree. After that the POCO will result in a further reduction in level.

In general, radioactive decay will result in a reduction of radioactivity and deferral of operations may be of benefit. However, there may be an issue if radioactive daughter chains exist producing isotopes that present greater problems than with the parent isotope. For example (Twidale, 1999), the 241Pu isotope primarily emits Beta radiation but it has 241Am as a gamma emitter. Furthermore, the parent isotope has a half-life of 12 years but the daughter has a half-life of 432 years.

The activation of the construction materials of the reactor and the presence of gamma-emitting isotopes are a significant problem in decommissioning. The areas concerned are the core internals, the biological shielding and the pressure vessel. The most problematic isotopes are 60Co, 108Ag and 94Nb, which have half-lives of 5.27, 418 and 20,000 years, respectively. It is therefore possible to achieve reductions in activity from, e. g. 60Co after a timescale of several decades, but the other problematical radioisotopes will remain.

The quantity of activated components varies considerably with the type of reactor and the size of the vessel. A large PWR has a reactor vessel of diameter 4.5 m and total weight 600 Te compared with a Magnox reactor that has a vessel of about 20 m of weight 5000 Te. The PWR vessel can be moved as a single item but this would not be possible for a Magnox reactor (Twidale, 1999).

Many of the principles for design basis assessment have been established for present-day reactors over many years. These include the ‘defence-in-depth’ principle, needs for diversity and redundancy, Safety Analysis Report (SAR) assessments and so on. In this section, the principal design basis approaches that are likely to apply to reactor licensing in the future are reviewed.

Most of the licensing submittals for present-day plant have been submitted using the conservative evaluation model (EM) methodology (Table 8.3). Conservative modelling was required to overcome lack of detailed knowledge of the phenomena. This methodology is commonly referred to as ‘Appendix K’ referring to the relevant appendix in the US CFRs (10 CFR 50).

BE methods are likely to be a common goal for licensing in the future, including those for future reactors. BE methods have been accepted by the USNRC (and other regulators) c. f. Appendix K Revision in 1988.

The acceptance criteria for fault studies are established by good understanding of the physics of present designs. However, as designs evolve, these criteria may need to be re-evaluated. The additional changes may also be required to accommodate extensions in, e. g. mode of operation.

There are likely to be increased requirements for Probabilistic Safety Assessment (PSA) studies (Table 8.4) to underpin deterministic studies and to help estimate doses to the population, source term of release, etc. Probabilistic targets are likely to become more

Table 8.4. Safety approach for severe accidents

PSA approach Selection of most probable sequences leading to a core melt

Provision of preventative or mitigative measures Wide coverage of possible sequences

Good quantification of the benefits from proposed measures Depends on the status of PSA accident analysis

Deterministic safety Definition of containment challenges from core melt behaviour

analysis approach Assurance of containment integrity by design measures

stringent, e. g. on core damage frequency or on containment limits. The current trend is to use best estimate methods for the frequencies and probabilities in PSAs.

The move towards BE methods is being supported by regulators and utilities because more realistic margin estimates enable a better quantification of actual risk to be obtained and enable a wider operating window.

However, there are developments required before the methodology is likely to be regarded as a mature engineering tool. It is necessary to be able to quantify unbiased uncertainty limits on key parameters (e. g. peak clad temperatures) and as yet the methodologies are not yet very practicable for licensing studies.

Another factor is that there is generally reluctance to change from an established methodology that is accepted by all parties.

Looking further to the future, risk informed methods (Wahlstrom, 2003) are being put forward by the USNRC but these methods require further development. Traditionally, the safety of NPPs has been justified by a deterministic approach based on the defence — in-depth principle and single failure criterion for design basis accidents, etc. Probabilistic approaches have further developed and now the probabilistic safety analysis methodology is becoming well established. These provide a means of taking a systematic approach to determining the probability and therefore risk of various failure sequences.

The idea of risk-based or risk-informed approaches is to focus on the most important issues in terms of risk. If PSAs are used to determine the risk then clearly it is important that there is confidence in the PSA methodology used. Risk-informed approaches can be applied to new reactor design or indeed to assist modifications of old plants, to target maintenance actions and inspections. PSA methodology can also be used to identify the safety categorisation of components.

The USNRC has made a commitment to move towards a risk-informed regulatory regime. Other regulators are considering the development of the approach.

The notions of design basis and defence-in-depth have been well established in the licensing of present generation reactors. For some future systems, these notions may need to be revised in the light of newer technologies with very different designs, materials and fuel cycles.

11.1. INTRODUCTION/OBJECTIVES

The purpose of this chapter is to focus on how the design basis for evolutionary water reactors is being extended. The approach continues to be based on the defence-in-depth but a major difference is to attempt to include more severe (core melt) accidents within the design basis. This is achieved in evolutionary designs by the adoption of new technical features, not only to protect against present design basis events affecting the core and primary circuit, e. g. loss of cooling accidents (LOCA), steam line break (SLB) and steam generator tube rupture (SGTR) but also ultimately to protect against early and late containment failure.

Many evolutionary plant designs incorporate passive safety systems in place of active systems but in other respects do not vary substantially from current generation designs. In this chapter, the focus is again on water reactor technology for power generation since these reactors are such an important class of interest. Reviews of advanced light water reactor designs are given in IAEA-TECDOC-968 (1996), covering evolutionary medium — and large-size reactor designs for power generation. Further review of evolutionary designs including strategic issues and economic viability is given in IAEA-TEC- DOC-1117 (1999). A common feature is that decay heat is removed from the primary circuit to large tanks or pools via natural circulation. There are some new phenomena associated with decay heat removal in advanced designs with such components that are not found in present generation reactors. These are discussed in Relevant thermal-hydraulic aspects of advanced reactor design (1996). An issue here for the plant designer is to ensure that such systems have sufficient heat capacity and also initiate as intended. In addition, reactor coolant inventory is maintained using passive injection rather than active pump injection.

Different containment designs have been proposed, utilising steel, concrete or composites. Heat removal may need to be via natural circulation cooling of the containment wall in the case of steel or enhanced in concrete based containments using passive heat exchangers. These and other passive systems are covered in this chapter.

The design basis for the containment has traditionally been that it must survive the peak pressure arising from a double-ended guillotine break of the largest primary or secondary pipes. The design basis for more advanced plants will have to cover a broader selection of accident sequences, perhaps including significant core melting. This selection will be based on a combination of probabilistic and deterministic analyses. The lowest probability high consequence sequences will still need to be covered by engineering judgement or other means. There will be a tendency for deterministic analyses to be carried out by best estimate rather than conservative methodologies.

Advanced evolutionary water containments include other measures to ensure they survive under severe accident loads. Measures (IAEA-TECDOC-752,1994) are introduced to prevent fuel coolant interactions (FCIs) to prevent direct containment heating (DCH) and to control hydrogen. They are also designed to reduce the source term by improving leak tightness. This is achieved via inherent safety features in the design, utilising passive heat removal systems in many cases. In addition to internal events, external events such as aircraft crashes and seismic events are also receiving special attention.

A number of more revolutionary designs of water reactor have been put forward as ‘inherently safe’ designs. These eliminate almost entirely active systems, e. g. relying on reactivity control via careful management of boron concentration. Some of these approaches are also summarised briefly although these are unlikely to be developed further at the present time.

12.6.1 LFR (Gen IV)

Lead and lead-bismuth systems are being considered in the GIF programme (The US Generation IV Implementation Strategy, 2003; Figure 12.5). Examples are listed in Table 12.6.

The system is based on natural convection cooling with outlet temperature 550°C. It could be somewhat higher ~800°C subject to improved materials development. It can be used within a long life closed fuel cycle of up to 30 years in some concepts. It is anticipated to be used for electricity production, hydrogen production and actinide management.

In these processes, low-temperature steam is taken from the power plant turbine of the supplying plant to heat the saline solution. In commercial distillation, there are a number of heat recovery stages in series, because of the high heat of evaporation of water. These stages are at progressively lower pressures, resulting in flashing and mechanical vapour compression to occur.

In general, the more stages in place, the more efficient is the process. The number of stages is limited by both economic and technical reasons, e. g. the overall temperature

difference between the heat source and the cooling water sink. The typical temperature reduction per stage for a commercial plant is 2-5°C. In terms of thermodynamic efficiency, expressed as kg of water produced against kg of steam used, the figure is 6-10 for MSF applications and up to 20 for MSD. These processes are described below.

14.4.1.1 MSF Distillation. In this process, seawater is passed through a number of stages where it is progressively heated (see below) until it reaches the main heating section supplied by the process heat source, see for example (IAEA-TECDOC-1056, 1998). The brine is then returned through these stages and freshwater is eventually obtained through a series of flashing and condensation processes. In particular, as the heated brine returning from the heat source passes into the first stage heat recovery section, flashing occurs due to pressure reduction. Vapour is produced which condenses on the entry pipe-work to the heating section within the first stage (providing the progressive heating referred to above). The condensate is collected in trays. This condensate together with the remaining brine (that has not flashed) is passed on the second stage. The process is then repeated for a number of stages and the separation process completed. Non-condensable gases are removed by a steam-jet ejector system. The seawater is also chemically treated to remove scale.

14.4.1.2 MED. This process also consists of a number of heat-exchange sections. At the first stage steam from the heating boiler passes through a tube bundle which is cooled by evaporating the entry seawater on the other side of the tube bundle. The resulting steam is then passed to a second stage heat exchanger. Any seawater not evaporated at the first stage is passed on to the second stage. The process is then repeated to complete the separation process. MED plants require similar scale removing processes as do MSF plants.

Several designs have been used. The main difference is in the design of the heat exchangers. The low-temperature horizontal tube multi-effect process (LT-HTME) has horizontal tubes and the brine is sprayed over the outside of the tubes. In the vertical-tube evaporation process (VTE), the evaporation is inside vertical tubes. The LT-HTME is the more dominant process used.

In general, MED plants are more efficient than MSF plants because their heat transfer processes are more efficient for given heat transfer area and similar temperature difference between the heat source and cooling water.

14.4.1.3 RO. RO is also used as a separation process (IAEA-TECDOC-1056, 1998). This process has been applied commercially and can produce freshwater down to between 100 and 200 ppm of total dissolved solids. The electricity consumption is in the range 4-7 kWe h m_3.

In this process, seawater (brine) and water are held in a vessel in two-solution compartments separated by a semi-permeable membrane. Pressure is applied to the compartment containing the brine, sufficient to overcome the natural osmotic pressure of the solution and the permeate pressure (NB this is negligible compared with the natural osmotic pressure). In these circumstances, water flows from the brine compartment to the water compartment, the brine become more concentrated and purified water is obtained in the water compartment.

As the seawater is fed into the brine compartment, it is compressed up to 70-80 bars, sufficient to overcome a natural osmosis pressure of the saline solution of about 60 bars. In practice, only a portion of this water flows through the membrane, the remainder is discharged. The flow through the membrane is proportional to the pressure gradient of the applied pressure less the solution osmotic pressure. The proportionality factor depends on a range of factors including the geometry (shape, area, thickness) and the chemical properties of the membrane, the pressure, concentration, pH and temperature. Membranes have been used of varying design, spiral-wound, hollow fibre, also tubular, plate and frame type, the former two designs being the most commonly used.

14.4.1.4 Hybrid Desalination. Hybrid desalination systems can be used to combine power generation, with MSF or MED, and RO processes. This combined capability can be utilised to advantage in different ways, depending on the size and type of energy source available and the water quality product requirements. There are economic and technical advantages of hybrid as compared to single process technology.

These include the utilisation of a common seawater intake, optimised feedwater temperature for the RO plant, taking cooling water from MSF or MED plant, blending of product waters, common water treatments and various other optimisations that can be made through common process requirements. Some of the different hybrid desalination systems are reviewed in (Awerbuch, 1997).

Some of the reactor concepts that are under consideration for desalination applications are shown in Table 14.5 and discussed below.

Severe accident research has been the main focus for reactor safety research over the past two decades. This largely started with the TMI-2 accident with phenomenological research, and the need to reduce severe accident risk further was re-enforced by the Chernobyl accident in 1986. For existing plants (Krugmann, 2001), measures have been introduced to reduce severe accident vulnerabilities, such as primary and secondary feed and bleed, filtered containment venting, hydrogen control by recombiners, igniters or by inerting, and filtration of control room air intake. For new designs, the IAEA has set more restrictive technical safety objectives (IAEA, 1999) such as severe core damage frequency less than 10 5 per plant operating year, elimination of sequences that could give rise to large early releases, and prevention of containment failure, thus limiting the need for off­site protection measures. These objectives have led to greater emphasis in reducing severe accident risk in the newer evolutionary designs.

There have been numerous research programmes over the past few decades to develop understanding of severe accident-related phenomena and also to develop guidelines for the prevention and mitigation of severe accidents. Much knowledge has been gained and at the present time, there is a reduction of effort on severe accidents R&D worldwide. Some research workers (Krugmann, 2001) believe that sufficient knowledge of severe accident phenomenology now exists. Confirmatory research, however, is still in progress in some areas. Also, clearly further work may be required to support a particular design in the event of new building. Recent research programmes are summarised in Table 15.5.

There have been relatively slowly changing trends in energy infrastructures over the past few decades but the mix of future energy providers is likely to change in the future. A number of countries are reviewing their energy policies for some time in the future. For example, the UK government published an Energy White Paper (Energy White Paper, 2003) in 2003, which proposed an energy policy looking forward to the year 2050. The paper covered all forms of energy requirement, from electricity generation, heating and lighting to transport, industry and communications. It was based on in-depth analysis following a report published by a UK-appointed strategy unit in 2002 (The Energy Review, 2002). Other countries are performing similar reviews, see e. g. the forward vision to 2030 published by VTT, Finland (Energy Visions 2030 for Finland, 2003). The strategy for the UK is outlined below, by way of example.

The major global challenges that need to be faced are:

— environmental and climatic change from carbon dioxide levels increase;

— decline of the world’s indigenous energy supplies, from oil, gas, and coal and how these may be replaced (e. g. by nuclear, renewables);

— the need to update national energy infrastructures over the next few decades to meet new energy mixes.

The goals of most of the industrialised countries are to:

— reduce carbon dioxide emissions with specific targets. In the UK the goal is to cut carbon dioxide emissions by some 60% by about 2050, with significant progress by 2020. Many, but not all, countries support the Kyoto Protocol;

— maintain reliability of energy supplies;

— promote competitive markets, raising the rate of sustainable economic growth and improving productivity. There is an increasing trend toward deregulation;

— meet other energy (non-electrical) requirements for industrial and domestic supply (e. g. to ensure every home is adequately and affordably heated).

To meet these goals, it is likely that an energy system will be required that is quite different from that of today. Much more diverse systems are envisaged. These will include a balance between imported energy and fuel, a mix of large power stations, that could include offshore marine plants, including wave, tidal and wind farms and also onshore wind farms. There would be an increase in local generation, including biomass, local wind and tidal generators and micro-generation from combined heat and power (CHP) plant, fuel cells or photovoltaics. Energy efficiency improvements would be expected from improved home design. Gas might be expected to form a large part of the energy mix whereas coal fired generation would either play a reduced part or be linked to carbon dioxide capture and storage.

There have been debates in many sectors (industry, learned societies), etc. on how goals for security of energy supply can be achieved and there are many different opinions. In the UK for example, the future of energy was the focus of the 2002 Parliamentary Links Day, organised by the Royal Society of Chemists (http://www. rsc. org/lap/parliament/linksday. htm). This included an audience of distinguished scientists and politicians and covered energy-related activities taking place in government and industry. The scope was broad across the energy spectrum, covering nuclear and non-nuclear, electrical and non­electrical applications.

Many of the presently operating nuclear plants will be shut down over the next two decades. In the UK, by 2020, the existing AGR nuclear power stations will almost all have reached the end of their lives and all the Magnox stations will have shut down. However, new build continues in Asia and some new plants are likely in Europe in the next few years. Nuclear power remains an option for the future for the UK. However, the Government White Paper did not propose it and stated that before any decision to proceed with the building of a new power station, there would need to be the fullest consultation and publication of a White Paper setting down the Government’s proposals. The arguments for a delay were both on economic grounds and concerned with the issue of waste disposal (sustainability).

It is increasingly recognised internationally (within the EC, US and Japan, as described in the section on hydrogen generation) that the ‘hydrogen economy’ has significant benefits as a clean and flexible energy system. In a report to the Parliamentary Links Day, the UK Government’s Chief Scientific Adviser also anticipates a significant move towards a hydrogen economy by 2020 (http://www. rsc. org/lap/parliament/linksday. htm). This view is also supported by the UK nuclear industry (Clegg, 2002) and others. The issue is how to produce hydrogen without releasing carbon dioxide.

In the Section 17.4, world events of recent years are examined. After that, a discussion is given on how these and other developments may shape developments in the near future. The remaining sections continue to look further into the future, covering likely developments over the next half-century.

The proliferation of plutonium for nuclear weapons purposes is a public concern. Significant quantities of plutonium were present in nuclear arsenals of countries with a nuclear weapons’ capability, particularly, e. g. in the US and the former Soviet Union. Plutonium is managed in nuclear fuel cycles and large amounts of plutonium are present in spent fuel from civil nuclear power plants. The subject has been studied and reported on by the American Nuclear Society and a number of other studies (American Nuclear Society, 1996).

In the short-term weapons grade plutonium from the weapons production progra­mmes is a significant proliferation risk. Weapons grade plutonium has 90% or more plutonium-239 which is the more suitable isotope for explosive applications. An important objective to ensure non-proliferation is to convert such plutonium to a different form, e. g. to the spent fuel standard (American Nuclear Society, 1996). However, the timescales for such action are relatively long, anticipated being as much as 15 years.

The longer-term issue is concerned with the increasing quantities of the plutonium being produced by the civil nuclear power programme. If it is sufficient to leave plutonium in spent fuel, how difficult is it for plutonium to be recovered, etc? There are issues associated with the choice of future nuclear power plants, e. g. whether a fast reactor will be built with a requirement for plutonium fuel. It will also depend on the utilisation or otherwise of advanced fuel cycles, e. g. MOX fuel. Either scenario would require the

Table 2.15. Reasons for deferment of plutonium in the civil nuclear fuel cycle

Minimise global nuclear catastrophe risk from irresponsible factions, e. g. terrorists Limit military applications from civil programmes

Reduce the risk from adverse political changes in nations with existing arsenals Remove international barriers to weapons stock destruction

It is not possible to guarantee protection despite IAEA safeguards and international monitoring Separation of plutonium is not justified by current or anticipated market conditions for the next few decades

Cochran (1996).

separation of plutonium from spent fuel with the increased risks of such action on proliferation.

The threats of proliferation have been categorised in American Nuclear Society (1996) in terms of ‘national’ and ‘sub-national’ threats. National proliferation is defined as the use of weapons grade material or material separated in the fuel cycle for weapons devices, authorised by national government approval. Sub-national proliferation would relate to the threat of seizure of nuclear material by smaller groups of people, acting without the support of government.

Retaining plutonium in spent fuel is likely to be an effective deterrent against sub­national threats, but reprocessing, albeit at small scale, is within the capability of many industrialised countries and so the spent fuel standard barrier does not provide protection at a national level. It was mentioned in the previous chapter that IAEA have defined controls to provide a high degree of protection across many eventualities. Nevertheless, it is clearly the effectiveness (or lack of it) of the implementation of these controls that is the issue. It is an issue of increasing importance as the stocks of spent fuel continue to increase and as the radioactivity of older stocks of spent fuel diminishes, the recovery of plutonium becomes easier.

It is argued by Cochran (1996) that due to the general increase in terrorism in the world and for other reasons, there are strong reasons for deferment of the further chemical separation of plutonium at the present time (Table 2.15).

There are sufficient high-grade uranium reserves to service the present fleet of reactors for at least 60 years based on the present fuel strategies and a demand of 65,000 tonnes per year (Energy Visions 2030 for Finland, 2003). Since the price of uranium has been dropping from the early 1980s, the emphasis of the mining industry has been to concentrate on the high-grade resources. There are considerable additional reserves anticipated from undiscovered conventional deposits and even more from less conventional sources such as seawater (Table 5.2).

Low-cost uranium resources are distributed worldwide as shown in Figure 5.1. In terms of production, the largest producer is Canada, generating about one-third of the total world supply. The next largest is Australia, about one-sixth, and other significant contributions come from Nigeria, Namibia and Russia.

Current production is around half of demand with the remainder coming from uranium stockpiles for the civil nuclear programmes in the US and Russia. There are considerably

more supplies available from highly enriched uranium and plutonium from dismantled weapons.

Regarding the efficient utilisation of fuel, a fast breeder reactor fuel cycle would use the uranium 50 times more efficiently, compared with other fuel cycles.

A thorium cycle would result in further fuel resource; thorium fuel is four times more abundant than uranium.

The Korean Standard Requirements Document (KSRD) was completed in 1990 and defines the requirements for the Korean Standard Power Plant Design, the generation of PWRs that were built in the mid-late 1990s. It has some similarities with parts of the EPRI URD but aims to reflect the wishes of the Koreans to develop their own design and construction capability.

User requirements for future plant designs began in 1993 with the objective of developing particular features and characteristics of future reactors suitable for Korea. The development of the requirements is being carried out in such a way that the requirements are being made available ahead of the design work of the Korean Next Generation Reactor.

Other requirements’ documents have been produced to support specific tendering specifications, e. g. the Taiwan power company requirements document was produced to support the Lungmen project in the mid-1990s.