9.2.1 Belgium

There are currently 7 power reactor units operating in Belgium, Doel 1-4, and Tihange 1-3. These are all PWRs. In 2002, these plants produced 60% of domestic electricity, an increase of 2.3% compared with 2001. Although these plants continue to operate at the present time, Belgium announced a moratorium in 1999 on building new nuclear plants and a law was passed in 2002 providing for nuclear phase-out from 2015 onwards (European Commission, 2000; Foratom e-Bulletin, 2003a). This corresponds to reaching the 40-year limit on the operating lifetimes of the plants.

9.2.2 Finland

Finland operates four nuclear power reactors that generate about one-third of the country’s electricity (27% in 2003) (World Nuclear Association, 2003). The plants currently

operating are Swedish boiling water reactors, Olkiluoto 1 & 2, operated by TVO, and 2 Russian-designed VVER plants Loviisa 1 & 2 operated by Fortum.

The building of a fifth reactor was approved by the Finnish parliament in May 2002. This is a significant development within Western Europe, because it is the first decision for new build in over a decade. The intention is for the new plant, expected to be the first European pressurised water reactor (EPR), Table 9.1, to be in operation by 2009.

The application to build a new reactor has taken into account economic factors, security of energy supply, and environmental considerations. The economic criteria related to lowest electricity cost, various studies showed nuclear as the cheapest option. The significantly higher capital costs of building and initial fuel load were about three times that of a gas plant but the fuel costs are very much lower. Comparative costs of the nuclear, coal and natural gas options were estimated at 2.40, 3.18 and 3.21 EUR c per kWh on the basis of a 91% capacity factor, 5% interest rate and a 40-year plant life (Foratom e-Bulletin, 2003a), showing the nuclear option to be economically favourable. The nuclear option also showed the lowest sensitivity to possible fuel price increases.

The decision is consistent with the Finnish 1997 energy policy, which stressed availability, diversity of provider, price and security criteria for new energy generation. It also stressed the need to meet international commitments.

Advanced reactors can be designed to include a number of features to improve leak — tightness under severe accident conditions. Under such conditions, the containment structures will have to withstand much higher loads in terms of pressure, temperature, radiation and chemical attack than under normal operation. Leak-tightness must be assured under this harsher environment; leaks may occur from the containment structures themselves, pipe and electrical penetrations and isolation valves, hatches, locks, etc.

Leak rates from steel containments or containments with a steel liner are expected to be lower than those from concrete containments without a liner. However, in some designs, the concrete containment is surrounded by a secondary containment. There is generally a requirement to improve overall leak-tightness in advanced containment designs, which can be achieved by improved primary containment design or possibly by taking credit for a secondary containment without improvement to the leak-tightness of the primary containment. By way of example of improved leak-tightness, the design leak rate for AP600 is 0.12% per day against other current PWR rates which are in the range 0.25-0.5% per day (IAEA-TECDOC-752, 1994). Containment leak-tightness needs to be maintained under all plant states including shutdown.

There are containment bypass sequences such as interfacing LOCAs and SGTRs and leaks in these events also need to be covered. The general approach in advanced containment design is to reduce the number of penetrations. Other special measures include pressurisation systems that keep penetrations at pressures higher than the containment pressure. These systems have been proposed in plants where there is a direct release path to the environment. Suction systems have been proposed to collect the leak contents and treat it before release. Other special components such as bellow’s valves and seal welds on large equipment hatches are also considered. With all these systems, however there are issues concerning their likely performance under severe accident conditions.

There are issues such as how the leak-tightness of the containment under severe accident loads can be tested. This is particularly so if the severe accident postulated pressures are higher than the peak design pressure, so periodic testing is not possible.

Systems have also been proposed for establishing whether a large opening in the containment boundary is pre-existing when an accident occurs. This approach has been

considered for existing plants. One such system, developed by EDF gives a measure of leak-tightness by measuring the rate of increase on containment pressure cause by the usual leaks in the air supply system.

There are clearly significant areas of research required to realise the ADS technology. Some broad scope areas are given in Table 13.4. These relate to general requirements needed for most of the different fuel cycles and applications. There are also particular engineering-related materials issues associated with radiation damage, and the need to extend the methodologies developed for critical reactors to the more complicated ADS-coupled transport situation.

Severe radiation damage can occur as a consequence of high current, medium-energy protons being injected into the target (Takahashi and Gudowski, 1997). Neutrons and charged particles are generated at energies reaching those of the protons causing radiation damage to the target and surrounding structural materials. This stems from the

Table 13.4. Research requirements

Transmutation of commercial power plant waste, particularly reactor grade plutonium Deployment of weapons grade plutonium in power production Assurance of proliferation resistant fuel cycle Benefits and utilisation of the thorium fuel cycle

Impact of different ADS options on radiotoxicity of the fuel cycle reduction Materials-related research, e. g. radiation damage of the target regions ADS safety issues and their resolution

Methodologies development for ADS, e. g. necessary developments of critical reactor models

displacement of lattice atoms within the target and from the energy the atom receives following emission of a nuclear particle, e. g. g ray (Wechsler et al., 1995).

The primary concerns on the effects of damage relate to hardening and embrittlement and the changes in mechanical properties and stability. The embrittlement is characterised by radiation defect clusters, helium aggregation to form bubbles, ductile brittle transition effects, and impurities arising from transmutation products.

The areas of particular damage will be surrounding walls and the window, which therefore needs to be replaced frequently in high-energy accelerators. Thus, damage is likely to be worst for a high-power accelerator with a large sub-critical reactor. This may be mitigated by adopting a concept with a smaller current and smaller sub-criticality. Similarly the structural damage in an accelerator driven system might be expected to be higher than in a corresponding critical reactor (Takahashi et al., 1994).

The adoption of suitable materials for the beam window section and the target side walls is a subject for research.

BREST 300 is a lead-cooled, pool-type fast reactor design operating at close to atmospheric pressure (IEA/OECD (NEA)/IAEA, 2002). The reference rating is 300 MWe. It has been put forward by RDIPE, Russia. It incorporates a loop concept for primary circuit heat removal. It is based on a relatively simple and robust design with passive decay heat removal to the environment. It has similar characteristics to those of the other lead — cooled reactors described above.

It has an increased core outlet temperature compared with the PWR making it a better candidate for somewhat higher temperature process heat applications.

14.6.2 Energy Amplifier

Lead-cooled subcritical reactors driven by a proton accelerator, such as the energy amplifier, are also being considered for process heat applications (IEA/OECD (NEA)/IAEA, 2002).

15.8. FUTURE REACTOR RESEARCH

Research programmes for the innovative designs described in Chapter 12 are described in IEA/OECD (NEA)/IAEA (2002) and Background Report for the Three-Agency Study (2001). Compared with the level of R&D investment in the performance and safety optimisation of current generation reactors over the years, and in evolutionary designs, the level of investment in future generation reactors is small at the present time.

To facilitate further research, it will be advantageous to set up collaborative international R&D programmes if possible. However there are many diverse designs under consideration and collaboration will only be possible if there are common interests in a particular field or topic. There are also the issues of commercial interests and the sharing of proprietary information to be addressed.

It is suggested in IEA/OECD (NEA)/IAEA (2002) that the setting up of a comprehensive experience database may be a useful initial activity in a collaborative relationship. Reactor designers could access this database to collect information on existing experience on the advantages and disadvantages of different reactor types.

Below are sections on the areas of research that are likely to be required for future innovative reactor systems. There are programmes already in place on research of some evolutionary systems issues; these are seen as a step towards developing the later systems. The discussion in these sections focuses particularly on the designs put forward by the GIF for Generation IV systems.

In summary, there are many R&D activities that will need to be accomplished before most of the innovative systems are available. The main technical developments for the Generation IV systems are summarised in Table.15.7. Some of these R&D activities have already started, e. g. for the nearer term SCWR and HTR concepts. SCWR activities have been ongoing since 2000 in the US, Canada, Japan, South Korea and in the EU, on materials and corrosion research. For the HTR concepts there are plans for the building

Table 15.7. Generation IV technology research

R&D activities

The US Generation IV Implementation Strategy (2003), Newton (2002) and Institute of Nuclear Engineers (2004).

of a Next Generation Nuclear Plant (NGNP) at Idaho in the US for R&D as a step towards the VHTR. There are also plans for an experimental technology demonstration reactor (ETDR) looking forward to the advent of GCR technology.

One of the limitations of fission power is that it depends on uranium (and possibly thorium) reserves, which are a finite resource. The utilisation of fast reactors and accelerator-driven reactors, especially if used in a thorium fuel cycle (since the reserves of thorium are greater those of uranium) would substantially increase the energy available from this resource, but nonetheless the statement remains true at least in principle. The goal of generating almost limitless energy from the fusion of appropriate light isotopes of hydrogen or lithium has been a dream of scientists for many years. This dream is not yet realised but it is deemed that sufficient progress has been made towards achieving controlled fusion, that fusion reactors deserve a mention in this introductory chapter on present generation reactors.

A significant problem in the development of a fusion reactor has been the confinement of the nuclei in order that the fusion reactor can proceed in a controlled manner. Fusion with a positive energy balance is only possible at very high temperatures. These must be so high that the thermal agitation of the atoms is sufficiently energetic that the electrostatic repulsion of the positively charged nuclei can be overcome, enabling collisions to occur.

A number of different fusion reactions have been postulated between the isotopes of hydrogen, helium and lithium. However, the majority of research efforts have concentrated on the deuterium-tritium reaction. This is the easiest reaction to achieve. Nevertheless, temperatures must be of the order of 100 million degrees.

There is also a confinement criterion, which requires that the period of the confinement time and the neutron density must exceed a stringent limit (Lawson Criterion).

Focus has concentrated on essentially two types of confinement, magnetic and inertial. Of these, magnetic confinement has received the most attention.

In magnetic confinement, a strong external magnetic field consisting of a high density of field lines is imposed. In a toroidal system, the field is circular such that the nuclei in the deuterium-tritium mixture travel in helical paths around the magnetic lines of force. This gives rise to the shape of a torus. In an ‘open’ system, the field lines are not closed but a series of magnetic coils are arranged to reflect particles back into the centre of the field. These are referred to as ‘magnetic mirrors’. The challenge with either of these methods is that the plasma should not contact the confining vessel, otherwise the temperature will fall.

In inertial confinement, pellets are made from a mixture of deuterium and tritium in a mixture frozen at about 15 K. These are then irradiated either by very powerful laser beams or by electron (or ion) beams. These compress and heat the material to fusion level temperatures; inertia results in very high densities for very short periods of time (order of a nanosecond). However, there are practical difficulties with this approach associated with the laser efficiency and engineering problems in achieving a continuous power output.

With regard to on-going research, the Tokamak system has probably attracted the most attention. The ideas were originally conceived in the former Soviet Union. The system is based on the closed magnetic field configuration in the shape of a torus. The Joint European Torus (JET) project in the UK has made progress in generating significant amounts of power, in 1991, 2 MW were achieved. However, the break-even point, i. e. the generation of as much fusion power as is required in heating up the plasma has not yet been achieved. The trend is generally for larger Tokamaks in the quest to achieve higher and higher temperatures and conditions that will satisfy the Lawson criterion.

In the US, the Lawrence Livermore Laboratory and the Los Alamos Laboratory are carrying out work on inertial confinement. Lower fractions of energy produced against input have been produced in comparison with the Tokamak approach.

Significant progress in fusion technology has been achieved to date and these have been described in this chapter. For the next generation of Tokamaks, the resources of the interested nations are likely to be pooled in the International Tokamak Experimental Reactor (ITER) Project.

In many countries periodic safety reviews are required to be carried out by the plant operator as a condition for his site licence. The primary objective of most PSRs is to undertake a detailed and comprehensive review of the safety of the plant, taking into account operational safety, the possible deleterious effects of ageing, and also advances in safety standards since the original construction or time of the last review.

Periodic safety reviews are usually complementary to the normal regulatory reviews that are carried out, e. g. between fuel cycles and do not affect them. PSRs have developed

for a number of reasons. Public confidence has diminished and regulatory requirements have become more stringent over the past decade or so driving a demand for higher standards of safety not only in new plants, but also in currently operating plants. Perhaps rather more importantly though, experience has shown that there are positive benefits from PSRs to both safety and performance and they are supported by both operators and regulators.

A list of safety issues identified for PSRs is given in Table 3.8 together with a frame­work for review that was endorsed by the 1991 IAEA Safety Conference (Goodison, 1997).

A review of experience of PSRs has been published in CEC Working Group (1990) and Goodison (1997). It concludes that PSR practices show considerable commonalties. This is particularly so within the EC due to similarities in regulatory regimes within the EC countries. At the top level, the procedure is broadly similar. There is agreement of the scope between the licensee and the regulator. The licensee undertakes the review, implements the modifications and reports to the regulator. The procedure is then followed by review by the regulator and the identification of any further modifications. Finally agreement is reached between the licensee and the regulator on how to fulfil the agreed programme.

The differences in PSR practices depend mainly on the methodology, the standards and scope that are adopted. These differences might relate to the standards for radiological protection or on the level of redundancy and diversity of the safety systems. The criteria for PSAs are also not universally agreed. There are also differences in the periodicity requirements for PSR reviews.

The potential benefits of PSAs include improved safety via the implementation of modifications to an improved safety level (closer to that of a modern plant), including the

Table 3.8. Safety issues to be addressed in PSRs

Safety issues Recommended procedures for assessing each issue

1-5

Assess each issue with current methods to determine the safety status Compare the safety status with current standards Identify shortfalls

Assess the safety significance of any

shortfalls and carry out remedial measures Implement practicable modifications and assess safety significance of remaining shortfalls Repeat for each issue

Gain indication of safety level compared with modern plant and identify shortfalls from current safety standards and best practices Improve plant routine operations including optimisation of maintenance, test and inspection techniques and improve plant availability Identify strengths and weaknesses of personnel

Gain improved understanding of plant safe working life and life-limiting causes Improve regulator confidence in the continued safe running of the plant, improve the licensee’s confidence for future planning and investment and improve national public and international confidence

Goodison (1997).

identification of short-falls in present practices and improved confidence (in the regulator, operator, public, etc.) (Table 3.9).

The first comprehensive review of the plant is usually the most demanding. Subsequent reviews would be expected to be quicker (and cheaper). Initial PSRs, of very old plants may require extensive modification or possibly result in closure. For future plants, initial PSRs might be expected to be less onerous.

A good summary of future fuel cycle issues and reactor strategies over the next few decades is given in Meneley (1998). This report considers short-, medium — and long-term time frames extending out for the next few decades. Clearly the choice of reactor and fuel cycle are inextricably linked. For example, the most widely operating reactor type is likely to be thermal reactors burning mixed uranium and plutonium fuel. As discussed earlier, the fast reactor could be operated as a stand-alone technology or in combination with thermal reactors. There is then the possibility of the thorium fuel cycle.

The largest change is the introduction of MOX fuel in LWRs and HWRs. PWRs are already being loaded with up to 30% MOX fuel. Higher percentage MOX fuel loadings are being considered but further technical work is required to establish whether fission gas release at high burn-up is a concern. There is also the question of high burn-up fuel under accident conditions. The capability of multiple recycle is also not assured; it may be that MOX fuel is limited to two or three cycles. MOX fuels are feasible for up to 100% loading in HWRs.

Further development and proof testing of fuel elements, either of MOX or uranium fuel, will be necessary for fuels capable of utilisation to higher burn-up. This will mean higher fresh fuel enrichment. It is expected that there will be a continuous drive towards higher burn-up because of the improved economics, certainly for batch rods.

For HWRs, the life of fuel can be greatly increased by a small amount of enrichment. Natural uranium imposes an inherent limit on fuel life. This enrichment leads to more flexibility in design and fuel management. RU can also be used in HWR since the U-235 content of uranium remaining after plutonium extraction is about 0.9%. A sequential once — through cycle in two different reactor types is under construction called ‘ double-burning’. The idea is to use discharged fuel from the first cycle for the second cycle without re­enrichment. Another cycle is the ‘DUPIC’ cycle, which aims to reform LWR pellets into HWR pellets.

In the short term over the next 15-20 years there will be an opportunity to conduct small-scale fuel development experiments, before prototyping in large-scale experiments in the medium term. It is likely that uranium-based fuel will take precedence over thorium — based technologies but there is the possibility for more consideration to be given to the latter. In the longer term, it is possible that recycling will be a more routine practice. Either the FBR or accelerator breeding could be used to convert fertile material to fissile material in large quantities. Thorium would have the advantage over uranium of a very high conversion ratio.

Future work programmes could, therefore, focus on increasing the reactor conversion ratio resulting in higher burn-up for a given enrichment, and reducing the need for burnable poisons. This could be achieved either through a thorium cycle in thermal reactors or FBRs utilising metal uranium-plutonium fuel. Other research will target increased fuel burn-up, and reduction of reprocessing costs. Finally, on-line fuelling carries with it none of the disadvantages of periodic shut-down of batch fuelling. Flexibility is much increased and parasitic neutron absorption is reduced for fuelling at full power.

Laws and statutes exist in most countries to ensure the safe operation of nuclear plant, see, e. g. EUR 20055 EN (2001) and EUR 16801 EN, ISSN 1018-5 (1996). Health and Safety laws are defined by government ministries, taking advice from various other supporting organisations. Safety standards are enforced by Safety Authorities and Regulators who grant licences for operation in accordance with national laws. These are reinforced by various international bodies, e. g. IAEA.

Internationally, IAEA principles have been established that govern the relationship between the regulator and operator. These are summarised in Table 8.1. These principles are embodied in the regulatory requirements of most countries.

In particular, these principles have played considerable influence in furthering the progress in the EU Enlargement countries from a closed safety culture to one of greater openness. Progress towards generally accepted international standards has also been

Table 8.1. IAEA safety principles (abbreviated form)

1. National governments shall establish a legislative and statutory framework for regulation

2. Prime responsibility for safety is assigned to the operator

3. Independence of the regulatory body from the operator

4. In all activities, safety matters have the highest priority

5. Establishment and implementation of appropriate Quality Assurance (QA) programmes

6. There are sufficient available adequately trained and authorised staff

7. The capabilities and limitations of human performance must be recognised

8. Emergency plans for accident situations must be in place and appropriately exercised

9. Site selection must take account of all relevant features affecting safety

10. The design must be suited to reliable, stable and manageable operation

11. Design shall include appropriate application of the defence-in-depth principle

12. Design technologies shall be proved by experience or testing or both

13. Man-machine interface and human factors shall be considered in design and operation

14. Radiation exposures to site personnel and to the environment shall be ALARA

15. The design shall be confirmed via comprehensive safety assessment and independent verification

16. Specific approval of the regulator is required prior to the start of operation

17. Operational limits must be defined from safety analysis, tests and subsequent operational experience

18. Operation, inspection, testing and maintenance must be conducted by adequately trained and authorised

personnel

19. Competent engineering and technical support to be available throughout installation life

20. Documented procedures must be established for anticipated operational occurrences and accidents

21. All plant operational incidents significant to safety must be reported to the regulator

22. All radioactive waste must be kept to a minimum (both in terms of activity and volume)

23. The design and decommissioning programme shall aim to limit exposures during decommissioning to

ALARA

24. The operator shall verify by analysis, testing and inspection that the physical state of the instillation remains

in accordance with operational limits

25. Systematic safety assessments shall be performed throughout life

Govaerts (1996).

influenced by other international bodies, e. g. OECD and the EC (within the EU and EU Enlargement countries).

South Korea currently has 18 reactors operating supplying 39% of the country’s electricity World Nuclear Association (2003). The reactors operating are PWRs and PHWRs. The first three units were purchased as turnkey projects; later plants involved local manufacturers. There were various vendors, Combustion Engineering (US), Framatome (France) and AECL (Canada).

In the mid-1980s, Korea embarked on a 10-year plan to standardise the design of its nuclear power plants via a collaboration with Combustion Engineering (now Westing — house). The exception to this plan was the building of three more AECL CANDU 6 units to add to the earlier Wolsong power plant.

The CE System 80 design was chosen as the standardised design and this evolved into the Korean Standard Nuclear Plant (KSNP) design. In addition to CE System 80 features, it also included many US advanced light-water design requirements. All further 1000 MWe units were of this type. In the late 1990s, an improved KSNP + programme was started. There are 6 such KSNP or KSNP + units under construction or on order, Ulchin 5 & 6, Shin Kori 1 & 2, and Shin Wolsong 5 & 6. These are scheduled to start up at various times between 2004 and 2010.

The advanced pressurised reactor (APR)-1400 is a further extension drawing on CE System 80 + design features. The System 80 + was chosen because it has USNRC design certification. The design for APR-1400 was completed in 1999 with enhanced safety and a design life of 60 years. The units scheduled are Shin Kori 3 & 4 and 2 units near Ulchin; these are not scheduled for start-up until 2010-2015. By 2015, nuclear power is expected to supply 45% of requirement (Table 9.5).

Fuel cycle facilities exist within the Korea Atomic Research Institute (KAERI) and the Korea Nuclear Fuel Company (KNFC) to supply PWR and PHWR fuel from uranium imported from Canada, Australia and elsewhere.

A revised waste-management programme came into being in 1998. Spent fuel is stored on the reactor site. The intention is to build a centralised storage facility by 2016. The long-term solution for high-level waste is deep geological disposal. Low — and intermediate-wastes are also stored on the reactor site. For this waste, a central repository is envisaged from 2008. This will allow shallow geological disposal of such waste.

Vitrification is also planned from 2006. In 2003, four sites were selected for detailed examination.

In the longer term, there are various plans for extending nuclear-related opportunities. Plans include the development of liquid metal reactors, the direct use of spent PWR in CANDU reactors (the DUPIC process) and utilisation of research reactors. The HANARO 30 MW research reactor started up in 1995. South Korea is participating in the US Generation IV programme.