Radioactive waste management, disposal and decommissioning are important areas of European research. Attention has focussed on evaluation of long-term disposal systems, including packaging policies and properties. Different aspects of waste retrieval are also under consideration. Underground facilities provide the best means of characterising potential disposal sites and for investigating different concepts of deep geological disposal. They are also required for data collection on the performance of the different barriers of protection. The performance of different types of rock ranging from clay, salt, marl or crystalline rock is being considered at possible sites. Within the EU, there are underground research facilities in the Asse salt mine in Germany and in the Hades facility in the Boom clay layer beneath Mol in Belgium. In France, some experimental activities are on-going in existing mines, e. g. the Amelie mine; sites for other underground laboratories are being considered.

Research tasks cover the testing of different methods of disposal, the methods of backfitting and scaling of repositories. They have also included the investigation of the long-term behaviour of components and groundwater flow and the migration of radionuclides.

An important objective of the research is to gain an improved understanding of the essential phenomena. The main requirement is to understand the release of radionuclides from the waste packages and their migration through the repository barriers to the environment. Characterisation of the different levels of waste in the waste volume is useful, to reduce the volume of highly active waste for disposal in deep underground repositories. Research is carried out into characterising the different waste forms and matrices (cement as a containment and barrier material, spent fuel itself and glass matrices of vitrified material). The quality control of nuclear waste packages and waste forms is being promoted to facilitate standardisation of checking methods, a common under­standing of R & D requirements and unification of test methods, etc.

The mechanical and chemical stability of the engineered barriers and the surrounding host rock is influenced by groundwater movements, thermal energy transfer effects at the interface and beyond and the radionuclide transport. It is important to understand any long-term degradation of these engineered barriers. The generation of gas can occur due to a number of processes affected by the nature of the waste, waste package, buffer and backfill materials and the nearby host rock. This could result in the build-up of pressure and possible structural problems in the repository.

Radionuclide migration research focuses on the thermodynamics of the solid-liquid phases’ equilibria and complexes with organic materials. These include groundwater colloid generation and transport, transport and retardation processes through porous and fractured rock systems and chemical thermodynamic and kinetic processes associated with radionuclide transport through the engineered barriers.

Studies have been performed on natural geological sites and have provided qualitative and some quantitative data on geochemical aspects (e. g. container corrosion, waste form degradation, radionuclide solubility and transport processes).

Palaehydeogeological studies also provide information on site evolution over geolo­gical time scales. Information of ancient flow patterns can provide understanding of past rates of uplift, erosion and of other, e. g. climatically induced changes of groundwater behaviour.

In addition to technical research, there are EU research programmes to enhance public understanding of the impact of waste disposal and to establish better methods of achieving public confidence and trust (European Commission, 2001). The objectives include gaining an understanding of the origins of public mistrust, evaluation of better means of communication and evaluation of decision making at different levels (e. g. local, national and international).

The decommissioning of nuclear installations is the final chapter in closing the nuclear fuel cycle. Research programmes are in place to develop innovative dismantling techniques to collect technical performance data, including data on specific wastes and doses arising from decommissioning.

The Nuclear Energy Act and a supporting Nuclear Energy Decree 1988 cover the construction and operation of nuclear facilities and all other matters in connection with the management and handling of nuclear materials and nuclear wastes in Finland (EUR 20055 EN, 2001). Additionally there is also the Radiation Act and Decree 1991 that is applied to the use of nuclear energy. Various Ministries have the responsibility for nuclear energy safety and security. The Radiation and Nuclear Safety Authority (STUK) is the primary regulatory body. It is an independent governmental organisation for the regulatory control of radiation and nuclear safety. There are also several acts (Act 1069/83 and Decree 698/97) that enforce the responsibilities of STUK.

The regulator does not specify particular design codes but there are guides that set down the requirements for the design of NPPs. The regulatory system is based on a comprehensive system of regulations and safety guides but it allows for the further development of safety culture within the industry. The current system is considered adequate for the licensing of future evolutionary LWRs.

Romania has only one operational unit at Cernavoda 1, which in 2002 supplied just 10% of the country’s electricity requirement (Foratom e-Bulletin, 2003c). This is a 655-MWe AECL designed CANDU 6 reactor, which currently has an operating licence through to April 2005 (Foratom e-Bulletin, 2003a). The operating licence is renewed every 2 years.

The completion and commissioning of a second unit, Cernavoda 2 is scheduled. It is currently about 45% complete and commissioning is expected in 2004 (Foratom e-Bulletin, 2003c; Table 9.6).

A significant amount of experience has accumulated from liquid metal (particularly sodium) cooled fast reactor operation. Twenty LMFRs, developed over the last 50 years, have been constructed and operated, resulting in nearly 310 reactor-years of operation (IAEA-TECDOC-1289, 2002). These include major large-scale prototype and demon­stration LMFRs and experimental fast flux test reactors.

Fast reactor development is being delayed in countries with relatively slow energy consumption growth and significant fossil fuel resources. However in some countries, with more rapid growth, or with limited uranium or fossil fuel resource, there is still interest in fast reactors for power generation. There is also a more general interest in fast reactors for plutonium burning, minor actinide transmutation and also for non-power producing nuclear heat applications. The latter topics are considered in separate chapters later in the book.

The countries where there is still a significant development programme in LMFRs include France, India, Japan and the Russian Federation. Other countries including Korea and China also have an interest in LMFRs.

In this chapter, some of the proposed designs that are being considered are summarised (Table 12.5). The designs that meet the more stringent safety requirements and likely to be competitive against LWRs for energy generation include the European Fast Reactor (EFR), the Prototype Fast Breeder Reactor (PFBR) from India, the Demonstration Fast Breeder Reactor (DFBR) from Japan and the BN-800 from the Russian Federation.

District heating plants supplying hot water and steam are widely used in countries with cold winters such as Denmark, Finland, Sweden and Russia. Large cities require 600-1200 MWt, smaller communities perhaps 10-50 MWt (IAEA-TECDOC-1056, 1998). The heat is produced by extracting steam from low-pressure turbines (for base load) and/or high-pressure turbines (for peak heat demand) and then distributed in insulated pipelines. These are on the order of 10 km, the shortest being a few kilometers, the longest built is in excess of 20 km.

The majority of nuclear applications for district heating have been reactors operating in co-generation with electricity producing mode. Such plants have been operated in Bulgaria (Kozloduy), Germany (Greifswald), Hungary (Paks), Russia (Bilibino, Belojarsk, Balakovo, Kalinin, Kola, Kursk, and Sankt Petersburg), Slovakia (Bohunice), Switzerland (Beznau) and Ukraine (Rovno, South Ukraine).

With regard to dedicated heating reactors, there have been demonstration plants constructed and tested in Canada (SLOWPOKE) and also China (NHR-5). There has also been a research reactor operating in Russia (Obninsk) for more than 20 years.

The plants include barriers to prevent any release of radioactivity into the grid network. A leak tight intermediate loop is added which operates at a pressure greater than that of the steam pressure taken from the turbine cycle. The loops are also subject to continuous monitoring. In about 500 reactor-years of operating heat supplying reactors, no radioactive contamination of the network has been reported.

While much experience exists with co-generation, small — and medium-sized reactors may be more appropriate for district and process heating and also desalination applications. These have been reviewed in IAEA-TECDOC-881 (1996).

Some of the newer concepts of reactors for district heating are shown in Table 14.3 and summarised below.

There are now many nuclear power plants operating which are at least 30 years old and many are approaching the end of their original design life. Central to the continued safe operation of these plants is the structural integrity of various safety critical components.

One such is the concrete pressure vessel. These vessels have to withstand large internal pressures (~ 4 MPa at 700°C in the case of an UK AGR). During lifetime, the pressure vessels deform and age.

The MAECENA project (Crouch et al., to be published) has the objective of investigating an important area of concrete behaviour that influences the ageing process, i. e. the softening and weakening of the effects of thermal and pressure cycling, and progressive creep and relaxation. The programme involves laboratory-based experimental work together with the development of finite element code methodology.

Concrete containment behaviour under various loading conditions has been considered in the European Commission CONMOD programme (Jovall et al., to be published). This aims to create a system for the assessment of containments throughout their lifetime. An important aspect is to develop NDT techniques and integrate these with finite element modelling techniques.

More mechanistic codes have been developed to model in detail, the various phases of a severe accident. They include SCDAP/RELAP5 (Allison et al.) for the in-vessel core

melt-down phase, VICTORIA (NUREG/CR-5545, 1992) for fission product effects including transport in the primary circuit and CONTAIN (NUREG/CR-6533, 1997), a containment phenomenology code.

The development of severe accident LWR codes (and supporting experimental programmes) has attracted significant research and development investment, much greater than that invested in other reactor types. The work have resulted in the development of improved accident management guidelines for the existing plants and improved robustness against severe accident challenges in the design of evolutionary plant.

Finally in this section, two examples of life extension proposals are cited for illustrative purposes.

2.8.6.1 Magnox Stations. In the UK, safety cases were assembled to extend the operating life of the Magnox reactors up to 40 years (Twidale, 1999). The original design life of these plants was 20-25 years. These plants were commissioned in the 1960s and 70s

Table 2.11. Issues for Magnox plant for extended operation

Degradation due to ageing — ageing of civil structures

Hot gas release — evaluation of gas ducts’ failures on post-trip cooling

Seismic events — evaluation of risk, plant integrity and ALARP principle

Extreme wind — engineering assessments of RPV and cooling pond foundations

Extreme flood — assessment of groundwater level extreme changes

Design codes and standards changes — structural modifications where required

Twidale (1999).

and many are still in operation today. The last station, Wylfa, is not due to shutdown until 2010.

Issues of concern in the cases covered both hardware and software, together with plant and system reliabilities and key performance parameters such as load and temperature histories (Bolton, 1996; Table 2.11). The hardware concerns covered the ageing of any materials that might result in loss of structural integrity and how these could be inspected and also electrical hardware. Software issues concerned the demonstration of safety margins and system reliabilities in terms of economic performance were also evaluated.

As a consequence, a programme of long term safety reviews (LTSRs) was instigated in the early 1990s to be followed up with PSRs at least every 10 years up until the end of each station’s life. These PSRs in some cases require more frequent inspection of some key components than every 10 years.

The scope of the Magnox PSRs covered re-assessment of the remaining safety margins in the plant, using probabilistic methods, where possible. Issues considered included: ageing degradation of the structures, fault loadings, external hazards (e. g. seismic, wind, flood, etc.) and changes to design codes and standards. Where structural ageing had occurred, modi­fications or repairs were carried out to ensure structural integrity was maintained. Studies were carried out to confirm the plant’s safety level for fault loadings and external hazards.

In particular, for Magnox plants, ageing degradation could occur due to one of a number of processes including carbonation, chloride attack (for coastal plants), and thermal and movement effects. In general though, the Magnox structures have remained in good condition. Loadings on the structures under low probability pipework failures outside the vessel have been assessed; in addition post-trip cooling of the reactor has also been demonstrated. Seismic qualification was performed taking advantage of more modern analytical ‘finite element’ methods than were available in the original design phase. Similarly, engineering assessments of the reactor vessel and cooling pond foundations were carried out using more up-to-date mathematical modelling and taking advantage of better understanding of the soil mechanics.

There have been new equipment and techniques developed for component inspection, maintenance, repair and replacement. These have been developed and tested in the laboratory and in full-scale experiments before being applied to plants (IAEA-TECDOC — 1175, 2000).

In Japan, techniques are being developed for the chemical decontamination of LWR reactor systems in preparation for the replacement of core internals. For the latter operations, full-scale mock-ups have been used. Techniques for the replacement of a PWR core barrel and bottom mounted instrumentation systems are being examined. Methods for the replacement of BWR core housing, core shroud, control rod housing and jet pump riser braces are also under consideration.

Holographic methods of inspection for the recognition and sizing of cracks are being developed. In Japan, intergranular stress corrosion cracking has affected reactor internals and core shroud and ways are being evaluated for mitigating this phenomenon. One solution is to replace components fabricated with SS type 304 with corresponding components made of SS type 316.

Manipulators are being developed for the purposes of in-pipe inspection, grinding and for repairing cracks in welds, e. g. between the vessel nozzles and pipes.

The main arguments for the continuation of nuclear power have already been discussed in earlier chapters, i. e. it offers a carbon-free energy source of energy via an established and proven technology. New plants are now needed for electricity generation to replace the plants that came on stream in the 1970s and which are now reaching the end of life. If new build programmes go forward, the issue is what plants to build? The drive for most new advanced evolutionary reactor designs is to achieve higher performance and safety by virtue of the design, rather than by simply improved operation which is often the only viable option available in the case for present generation plant. Satisfying these requirements is crucial to meet the increasing competition from natural gas electricity

generators, coupled with an increasing trend of de-regulation. Broad design objectives are considered in Section 7.3.