An important issue in a deregulated electricity-generating industry concerns the trading framework for the electricity-producing utilities. The US Californian model is described by way of example, see Shiffer (1999). In this model, the generation of electricity is carried out by independent companies (generators) with customers free to choose their supplier of choice. The price that the generators receive can either be negotiated directly with the customer or is determined by a state-owned trading entity called the Power Exchange, into which all generators bid a price for their services. Other countries have similar models; indeed the US approach was borrowed from the UK model.

There are profound financial impacts on nuclear generators arising from deregulation and this modus operandum is summarised in Table 4.1 (together with other cultural and

personnel issues). The electricity price received is clearly determined from a competitive bidding process; this price may be reducing. The revenues for the plant must cover the range of costs incurred from operation, including operation and maintenance, capital, fuel and taxes, which are broadly fixed, at least on the timescale of price fluctuations. There is considerable uncertainty on the price obtained, which may vary from hour to hour.

In the UK electricity-generation sector, competition since deregulation has forced down wholesale electricity prices. The UK currently has sufficient generating capacity, but the reduction in nuclear generation and closures of some coal stations may result in smaller reserve margins. New electricity trading arrangements (NETA) have been introduced to encourage flexibility and still further competition. This will alter the market structure and remove payment for capacity; this would tend to penalise baseload generators such as nuclear plants.

Pre-deregulation, the primary objective was high-quality technical plant and operational standards. Achieving low costs was not the highest priority. Post-deregulation, low operational costs must be achieved and these can only be attained through high operational efficiency. The challenge for designers and managers is to ensure that operational efficiency and low cost operation also equate with high reliability and standards of safety. IAEA Technical Report No. 369 provides a description of good practices.

This section considers various decommissioning strategies and the options available.

Generally, the ultimate objective in decommissioning is to return the site to a state whereby it can be used without restriction (de la Ferte, 1996). The IAEA have defined three stages of operation, see Table 6.3. The timescale for carrying out these activities will depend on the decommissioning strategy. Work may proceed from one stage to the other relatively quickly or may take place over many decades, perhaps over as many as a 100 years.

Table 6.3. Stages of decommissioning

Stages Potential periods of activity

1. Removal of nuclear fuel removes 99.9% of the radioactivity

2. Dismantling of structures, e. g. other than

the reactor itself and its surrounding biological shield

3. Total dismantling, removal of all materials

with radioactivity exceeding natural background

Table 6.4. Decommissioning options

Options Features

1. Safe enclosure Following defuelling as rapidly as possible

enclose the active inventory without immediate dismantling

2. Safe enclosure together with partial dismantling Similar to Option 1 except with

partial dismantling and storage of components on site for total dismantling later

3. Immediate total dismantling Total dismantling with removal off site

of all waste materials

EBmann (1990).

On this basis, a number of options are available to the operator. Table 6.4 summarises these options. Option 1 or the ‘safe enclosure’ option leaves the plant essentially unchanged after the completion of Stage 1. Once the entire operating medium, e. g. the fuel has been removed, all the nuclear plant equipment is sealed. The objective of safe enclosure is to enclose any remaining activity as soon as possible without immediate dismantling and then when this has been achieved to wait for the inventory radioactivity to reduce by natural radioactive decay.

Option 2 or partly dismantling with safe enclosure involves placing certain active components of the plant obtained by dismantling along with other plant components in a safe store. The principle of this approach in terms of environment protection is similar to that of Option 1. Total dismantling will be completed at a later date, once the inventory has reduced sufficiently by natural decay.

Option 3 is based on the premise of total dismantling. Here all active and inactive waste materials are removed from site directly after the end of operational life.

There are various important technical, safety and economic issues that need to be addressed in all decommissioning programmes. These are summarised in Table 6.5.

Table 6.5. Important issues to be considered in decommissioning

Technical aspects — structural integrity issues, inventory management and volume of material, degree of automation, remote handling requirements, decontamination arrangements, health physics and available dose minimisation techniques, material re-usage following decommissioning

Decommissioning policy — regulator requirements, licensee decommissioning strategy and workplan, timing of operations

Safety and environment — control of hazardous releases during decommissioning operations, waste treatment, temporary or permanent storage, repository storage

Radiological issues — adherence to ALARA principle for personnel exposure, advantage in delay in plant dismantling (safe enclosure)

Public relations — management of waste disposal concerns

Economics — relative benefits/disadvantages of ‘safe enclosure’ versus ‘immediate dismantling’

The technical aspects of various activities that need to be considered in the planning of plant decommissioning are discussed in this section. The resolution of these issues will generally be site dependent and depend on the infrastructure for decommissioning that already exists, both at the national and local level.

Assurance of structural integrity fidelity and effective management of radioactive inventory are key pre-requisites towards ensuring the safe and efficient management of decommissioning operations.

Nuclear safety in Russia is governed by laws on radiation safety of the population of the Russian federation, a law on radioactive waste management and a law on the utilisation of atomic energy (IAEA-TECDOC-905, 1996).

The regulatory standards that are applied to nuclear plants in Russia are described in a basic document ‘Atomic Power Plants General Safety Regulations’, OPB-88 (IAEA — TECDOC-905, 1996). This is supplemented by a number of documents on basic regulations for the assurance of safety in nuclear plants (OPB-88), PNAEG-1-011-89; radiation safety standards, NRB-76/87; nuclear safety rules for NPPs, PBYa-Ryu-AS-89, PNAE G-1-024-90; and rules of design and safe operation of equipment and piping of NPPs, PNAE G-7-008-89. In 1993, a regulation was issued by the Federal Nuclear and Radiation Safety Authority of Russia on NPP Siting to deal with limiting the consequences of severe accidents and requiring that they be considered in the design of future reactors (Federal Nuclear and Radiation Safety Authority of Russia, 1993).

There is a large work programme in improving the safety of currently operating plant to ensure their safety standards are consistent with latest regulations.

South Africa has 2 PWR units operable in its present nuclear programme.

The South African utility ESKOM is ready to go forward on the development, construction and commissioning of a demonstration unit for the pebble bed modular reactor (PBMR), subject to the required statutory approvals (Foratom e-Bulletin, 2003d).

Considerable Russian experience has been built up from operation of a number of experimental and prototype fast reactors including BR-10, BOR-60, BN-350 and BN-600 (IAEA-TECDOC-1289, 2002). BN-600 has operated reliably at Belojarskl in Russia (IAEA-TECDOC-1083, 1999). BN-600 has a nominal power output of 600 MWe and has been in operation in 1980 with an average load factor of 70% (IAEA-TECDOC — 1289, 2002).

The major emphasis in the Russian Federation is in continued design improvement and improved economics. The main applications are seen to be for energy production and the conversion of plutonium and minor actinides. The design of BN-800 has been completed and a site licence issued for the construction of a BN-800 at Yuzno-Uralskya and Beloyarskaya in Russia (IAEA-TECDOC-1289, 2002).

BN-800 is based on a three circuit flow system incorporating three primary loops, three secondary loops and three steam generators. The reactor core and radial blanket are built up with assemblies in a hexagonal lattice. The fuel is MOX sintered pellets. Compared with BN-600, there are a number of design improvements.

These include improved economic and operational performance, enhanced reliability of components and simplification of systems (e. g. single steam turbine compared with three in BN-600). Improved safety is included with the introduction of a passive emergency protection system and improved safety system redundancy and diversity.

For many years Russia has developed marine nuclear reactors to power their ice-breaker transport fleet. Taking advantage of this experience, there are recommendations for KLT-40 type nuclear energy floating complexes to supply electricity and heat to remote regions in the far north and east of Russia.

This reactor type is being developed by the Experimental Machine Building Bureau (OKBM) of the Russian Federation. It incorporates the nuclear reactor in a barge and can, therefore, be regarded as a ship reactor (IEA/OECD (NEA)/IAEA, 2002). The floating power unit (FPU) is assembled in a factory. Factory fabrication can be optimised by the delivery of two units, including the steam side plant.

The plant is a conventional pressure-vessel loop type reactor including hydraulic loops, pumps, steam generator and pressure vessel. It contains a number of safety features including diverse and redundant shut-down and passive decay heat removal systems. There is self-regulation of the power levels at all power levels because of negative temperature and void coefficients.

There are reduced volumes of low and medium level waste. All the waste obtained over the 12-year operating cycle is stored on the FPU. The outlet temperatures are similar to other current PWRs.

15.3. FUEL BEHAVIOUR

This section examines the experimental programmes devoted to fuel behaviour under normal operation but focussing particularly on transient/accident conditions. Examples of important current programmes are given in Table 15.3. Research is progressing on the behaviour of advanced fuels and clads in current generation plant to optimise performance, without challenging the margins of existing safety cases. Much of the present work is also relevant to evolutionary water designs. Many of the issues were discussed earlier in Chapter 5. Research programmes for advanced fuel cycles play an important part in the progress towards the future innovative designs that are being considered under the Generation IV initiative. These are discussed later in this chapter.

Finite element techniques have offered a substantial modelling improvement capability over the more classical mechanical equilibrium codes. They can be used for evaluating stresses, strains and displacement of components for different accident situations. They can model both static and dynamic effects. They can be used to evaluate the failure mode of structures, e. g. containments under increasing loadings (IAEA-TECDOC-752, 1994).

It is considered desirable (Sammarato et al., 1992) that the methodologies for future advanced containment should be based on ‘best estimate’ approaches. The more advanced codes all offer this capability. The traditional modelling approaches were generally much more conservative.

The development of improved codes can only be realised if there are corresponding improvements in input data. There is a need therefore for ensuring that adequate materials data are available.

Dynamic load modelling under severe accident loads is now within the capabilities of the computer codes but the problem may be in specifying the appropriate boundary conditions, e. g. in assessing the load resulting from a hydrogen detonation. The thermo-mechanical assessment of core catchers is also a modelling requirement for the assessment of advanced containments. This has been investigated in France (Millard et al., 1992) and Italy (De Rosa et al., 1992).

Reactor Accidents (or the potential for accidents) are undoubtedly a public concern for the continued operation of the civil nuclear power programme. There have been comparatively few serious accidents but those that have occurred, e. g. Three Mile Island-Unit 2 (TMI-2) and Chernobyl have had a pronounced affect on the expansion (or lack of it) in the nuclear developed countries.

A good review of the accident record of the nuclear industry is given by Mounfield (1991). Some incidents have occurred during the handling of industrial isotopes or other exposures to ionising radiation; these have resulted in a small number of fatalities. Some accidents have happened at experimental facilities; e. g. a criticality accident occurred in 1961 in a small prototype BWR reactor (SL-1) located at Idaho in the US resulting in the death of three technicians. Other accidents have been recorded in experimental and power reactors involving criticality and also fuel melting. A partial fuel meltdown occurred at St Laurent 1, a 480 MWe plant in France in 1969. Incidents have taken place at San Onofre 1 in California in 1973, Brown’s Ferry, Alabama in 1975 and other more minor (but still serious) events have occurred on some other plants.

In the UK, the only serious accident that caused public concern was the Windscale fire. This fire resulted in significant releases of radioactivity; estimates of 20,000 Ci of I-311 are given by Mortin (2000). As a consequence of this accident, 14 workers at the plant received serious doses of radiation, and there was a suspension of milk production in the surrounding area.

The first most damaging event in terms of limiting nuclear power plant expansion was the TMI-2 accident in 1979. The accident resulted, partly through mismanagement, in a severe core melt down that threatened the integrity of pressure vessel boundary. TMI-2 had various important consequences. It effectively terminated the construction of new power plants in the US (Chung, 1998). It also impacted on the philosophy of approach to severe accident safety. Previously, attention had focussed almost entirely on prevention, after TMI-2 there was a much increased focus on accident mitigation. Nevertheless, despite the substantial core melting, the only significant releases to the public resulted from Xenon-133 and the health consequences were not judged to be significant.

A study was carried out (Blee, 2001) to look at lessons learned over the last 22 years since the TMI-2 accident (Table 2.13).

Table 2.13. Lessons learned from the post-TMI-2 and Chernobyl era

Along with safe operations and good economics, effective communication is vital, particularly in the aftermath of abnormal events

Industry fortunes are global as further demonstrated by Chernobyl — crisis management is vital Environmental linkages have yet to embraced — the beneficial role of nuclear energy in protecting the environment should be proposed The need to manage media publicity Recognition of the benefits of long-term vision

Blee (2001).

The World’s worst nuclear power plant accident occurred at Chernobyl in 1986. The consequences of this accident have been much discussed and publicised. This accident resulted in a massive explosion, dispersing radioactivity over much of Northern Europe. The cause was essentially operator error but subsequent investigation indicated major weaknesses in both technical specifications and management. The Chernobyl accident resulted in moratoria for the construction of nuclear plants in some European Countries, e. g. Italy.

A relatively recent incident involving fatalities occurred in 1999 at the Tokai-mura uranium processing plant in Japan (Suzuki, 2000). This accident resulted from a fission reaction in a precipitation tank of uranyl nitrate solution. Several workers suffered severe radiation sickness ultimately resulting in their deaths, several months after the incident. This accident was investigated in depth and various deficiencies in operating processes (operational and technical, management and control), in the licensing process and in the safety regulations were identified.

As with most industries, experience from accidents has resulted in the implementation of better operational practices and technical improvements. Many of the accidents to date have resulted from human error and the need for improved training and understanding of ‘human factors’ issues is one of the most significant lessons learned.

In current generation and new plants, digital instrumentation, control technologies and also self-diagnostic systems are under development. There are also new control room and man-machine interface improvements that include human factors engineering consider­ations. Examples of reactors with improved control room design include the latest PWR designs under consideration in Japan and also the control room design in the Korean next generation reactor.

Many of the improved practices in regard to design and technology of new plants and in the back-fitting of older plants comply with the traditional design basis objectives to include increased redundancy and diversity (IAEA-TECDOC-1175, 2000). Thus more emphasis is placed on reducing vulnerability to single component failure and in ensuring that the design accommodates sufficient scope for maintenance during plant operation. Increased diversity reduces the frequency of common mode failure. The VVER-440/230 improvements referred to in Section 4.9 are a good example of back-fitting improvements resulting in improved redundancy and diversity. Other practices include the careful planning of the plant geography to ensure access for inspection, maintenance, replacement and repair; these are being studied in Japan as are the increased use on-line testing and maintenance practices, referred to in Sections 4.9 and 4.10.

4.2. NUCLEAR FUEL CYCLE

Improvements in nuclear fuel technology and the fuel cycle are leading to better performance and economics of power plant operation. These are considered in the next chapter.