Governmental and regulatory infrastructures have to a large extent developed in parallel with national nuclear power programmes, certainly in Western Europe, US and Japan. The independence of regulatory bodies from the organs of government or private industry promoting nuclear power is an important requisite that is now generally internationally accepted. Countries in former Eastern Europe have made significant progress in establish their own independent regulatory bodies, during the 1990s, having previously relied on the centralised systems of the former Soviet Union.

IAEA Basic Safety Standards were established in the mid-1990s to ensure the safety of all applications of nuclear technology, particularly industrial and medical applications. In some countries, these had developed without adequate infrastructures to ensure the safety of these applications (IAEA/NSR/2002, 2003).

As stated in the IAEA principles earlier, one of the tenets for a strong independent regulator is the availability of an adequate pool of qualified staff. As noted in Chapter 2 with declining nuclear programmes in some countries, there are fewer qualified engineers available to regulatory bodies who frequently seek engineers who have acquired on-site experience in industry.

To meet these requirements, the IAEA has instigated various education and training programmes. These aim to promote self-sustaining capabilities in the member states, at all levels, national and regional. These include programmes to train trainers, disseminate materials and harmonise on-the-job training programmes. They are also establishing centres for education and training, centre networks and exploiting modern technology for distance learning and e-learning.

There is a large amount of information available on the safety and operation of nuclear power plants (NPPs), which has not been fully disseminated worldwide. Networks are being developed to share this information and provide a means of mutual sharing of information. International bodies including (e. g. IAEA, EC, CSNI) act as facilitators in various ways with regard to sharing this information.

6.3.2.1 Bangladesh. Bangladesh currently has one operating research reactor. There are plans to reconsider building a 600 MWe reactor (World Nuclear Association, 2003).

6.3.2.2 Indonesia. There are currently three research reactors in operation in Indonesia (World Nuclear Association, 2003). The potential for nuclear power generation is under review. An original feasibility study recommended that first units totalling 1800 MWe should be commissioned about 2004 but according to World Nuclear Association (2003), nuclear power has been deferred indefinitely.

6.3.2.3 North Korea. In North Korea, there are two partially built units and a research reactor (World Nuclear Association, 2003). A South Korean Standard Nuclear Plant type is also under construction.

6.3.2.4 Philippines. The Philippines have one research reactor but it is not currently operating (World Nuclear Association, 2003).

6.3.2.5 Taiwan. Taiwan currently has six units in operation meeting 22% of its electrical energy requirement. Two further advanced reactor units are being built (World Nuclear Association, 2003).

6.3.2.6 Thailand. Thailand has one research reactor and one reactor under construction (World Nuclear Association, 2003). There are some tentative plans to have a power reactor in operation in the next 10 years. It would be followed by five further units.

6.3.2.7 Vietnam. Vietnam has one research reactor (World Nuclear Association, 2003). The country is studying the viability of nuclear power and possibly installing some nuclear power plant by 2010.

The ETGBR grew out of the AGR technology developed in the UK during the 1970s. The core design took advantage of lessons learned from both AGR technology and the fuel design took advantage of experience of the sodium-cooled fast reactor (SFR) experience.

An important objective for the fuel and core design was to obtain a good breeding gain. The fuel consisted of MOX or UOX in a steel clad. Reactivity was controlled by three diverse and separate control rod systems for both control and shutdown. The burn-up target was 10%.

The reactor coolant system consisted of an integrated AGR design, with the boiler and circulators contained in a pre-stressed concrete pressure vessel. The main components were based on AGR technology in terms of materials and design. Since the core temperatures were cooler than those for AGRs (limited by maximum clad temperature), the cooling system had to withstand a less demanding environment than for AGRs.

The containment was designed to be less embracing compared with the primary/se — condary containment adopted for LWRs. It was, however, vented to mitigate the release under severe accident conditions.

It was felt that the design and safety philosophy of the ETGBR could be potentially licensable in the UK. There has also been recent interest in the design because it can be flexible in its fuel cycle. This could provide the option of achieving modest breeding or alternatively to enable the burning of plutonium and minor actinides.

The cost was reviewed in the 1970s, being found to be 10% greater than a PWR of the day. These costs were favourable in comparison with the AGR figure (increase of 25%) and the LMFBR figure of 60%. More recent studies have shown that the ETGBR could be economically competitive in comparison with other advanced reactors.

There have also been interest in GCFR technology in the US and a number of designs developed.

It is clear that ADS offer some interesting additional features that complement the conventional critical reactor technologies that currently exist or that may be considered in the future. However, there would need to be significant investment, research and

Table 13.6. Future activities Technical research in selected fields

Investigation of different fuel cycles and energy systems for different applications Further experimental programmes in demonstrating ADS feasibility Extension of nuclear data into the ADS applications Increased international collaboration and information exchange

development in ADS technology if these systems are to become available commercially. Future activities that could be foreseen are outlined in Table 13.6. Clearly ADS would be subject to the same economic competitive pressures that face existing nuclear plant. Similarly ADS would have to meet the increasingly more stringent safety and environmental standards that are being imposed for licensing.

The ageing materials European strategies (AMES) network was set up by the EC to bring together expertise on nuclear reactor materials (Gerard et al., 2001; Sevini et al., 1999). The most important area for research for effective plant life extension and management is the reactor pressure vessel (RPV), but metallic components in general (e. g. internals, pressuriser and piping) were targeted in AMES. The other principal areas are irradiation embrittlement and thermal ageing. In recent years, the network has been enlarged to include representatives from the Central and East European countries. AMES members collaborate in the TACIS and PHARE programmes to integrate findings for PWR and VVER LWRs.

The phosphorus influence on steel ageing (PISA) programme (English et al., 2001) is an experimental study to investigate the influence of phosphorus on RPV steel irradiation embrittlement. The objective is to improve understanding by segregating the phosphorus to grain boundaries and determining the effect of brittle inter-granular failure mechanisms on the RPV properties. The experiments focus attention on investigating various irradiated steels and metal alloys. The lack of phosphorus segregation data on certain steels under irradiation conditions relevant to end-of-life was recognised in a recent review (English et al., 2002). Both PWR and VVER reactor designs are covered in the project. The understanding of phosphorus segregation in irradiated and thermally aged fuels is now advancing significantly.

Another EC programme, fracture mechanics based embrittlement (FRAME) (Valo et al., 2001) aims to irradiate a relatively large number of different materials, chosen to determine the effects of chemistry on embrittlement. The objective is to develop fracture mechanics based trend curves. Irradiation shifts are measured and these are compared with existing Charpy-V (CH-V) based regulatory and other trend curves. Since the cleavage initiation fracture toughness material property KJC, is required for pressurised thermal shock (PTS) safety analyses, the availability of directly measured data will help to remove uncertainties (Sokolov and Nanstad, 2000) from the utilisation of CH-V test data.

The need for accurate data on neutron fluence to be used in conjunction with materials data is important for determining the life of nuclear power plant components, particularly the RPV. The RETROSPEC Dosimetry programme of the EC (Voorbraak et al., 2001) aims to provide retrospective fluence data by focussing on the niobium reaction 93Nb(n, n0)93Nbm. The methodology is being developed by examining specimens from material test programmes in research reactors, the Petten High Flux Reactor (HFR) and from specimens in surveillance capsules from the Dukovany NPP and the Loviisa NPP. Four steels have been selected, which are representative of the RPV in East European VVERs. The methodology is validated by comparing results from the retrospective analysis with the measured fluence at the locations of the specimens. It is concluded that retrospective dosimetry is useful in determining the neutron fluence at various locations inside a nuclear reactor, e. g. at RPV welds. Retrospective dosimetry has been reported previously by a number of researchers, see also van Aerle et al. (2000).

The EC GRETE programme (Delnondedieu et al., 2001) is concerned with the development of innovative non-destructive techniques for the inspection of critical components that may affect decisions on the lifetime of the plant. The objective is to assess techniques that aim to detect changes in materials before macro-structural defects occur, thus allowing remedial action to be taken. The techniques are evaluated in relation to neutron irradiation damage of the reactor vessel and the thermal fatigue of piping of the primary loops. Aged samples are being examined metallurgically and mechanically and then tested using various non-destructive techniques. All the known NDT techniques and their limits and limitations have been listed within the frame of the AMES project, see Delnondedieu et al. (2001) and Series of AMES reports (1975).

Large thermal-hydraulic system codes have been developed for the analysis of various fault conditions and initiating events. Examples of such codes include TRAC (Guffee et al.), RELAP5 (RELAP5/MOD3 Code Manual, 1995), CATHARE, ATHLET and RETRAN together with other industry system codes. Recently the TRAC-M or TRACE code is being developed which constitutes an amalgamation of the TRAC and RELAP5 codes (Spore et al., 2001). For the PWR, these codes calculate the flow, temperature and pressure in the primary circuit and secondary side. They include modelling of the reactor vessel, hot and cold legs, pressuriser and steam generators and safety systems using fundamental components of pipes, vessels, valves, etc. Most of the system codes can be adapted to other water reactor systems, e. g. BWR, VVER and RBMK.

In addition to thermal-hydraulics models, these codes typically contain point kinetics models to model the reactor power, and also 1D (radial) fuel rod models. Many have now been coupled to 3D neutronics codes of the type described above. In the UK, for example RELAP5 has been coupled with the PANTHER code, e. g. using the TALINK code (Page et al., 1998). RELAP5 has also been coupled with other neutronics codes. Generally, a few individual fuel rod models are coupled to a single thermal-hydraulic channel, e. g. an average rod and a hot peak rated leading rod. The fuel rod/coolant heat transfer exchange includes cladding to coolant heat transfer correlations, a gap conductance model between the fuel and clad, and thermophysical properties for the fuel. Ballooning, oxidation and rupture models are also required for the clad for LOCA analysis.

In order to extend the operating life of a reactor, it is necessary to review all aspects of the safety case. Plants are usually designed for a certain life, which is based on knowledge at the time and forecasts extending over a period of 20 years or more. In practice, the actual working life may be different and depend on a number of factors. These might include (Twidale, 1999):

— changes in operating conditions compared with the assumption in the design (these could affect margins);

— findings from maintenance inspections;

— results of test programmes and;

— outcome of safety assessments.

In addition, the operating experience of the plant (and possibly sister plants) and the accumulation of materials and other plant data will also impact the life. If these are favourable, a licensee may seek permission from his regulator for life extension.

As discussed earlier, PSRs have been introduced as a means of reviewing the safety of a plant on a regular basis. The results of PSRs strongly influence decisions for future plant operation and become increasingly important for older reactors or plants where life extension is under consideration. PSRs are conducted, usually at least every 10 years; for some plants they are conducted more frequently, particularly during later life.

The case for plant life extension would have to confirm the plant’s safety for the proposed additional operation. It would include identifying any features that might restrict the plant-operating envelope during this period. A secondary objective may be to assess the plant’s safety standards against current safety standards. This objective is usually realised somewhat partially since it is realised that it is not reasonable to expect older designed plant to wholly meet the safety standards of the day. In this circumstance, consideration would be given to the age of the plant and the intended life extension.

The top priority components for safety justification review are likely to include the reactor pressure vessel, control rod drive mechanisms, internals, supports and the biological shield, the primary circuit including the pressuriser and steam generators, coolant pumps, containment structures, and control and instrumentation. Other more minor components clearly have a safety justification, e. g. valves, smaller pumps, sensors, etc. but these components would have been routinely replaced under regular operation within the original design life.

Maintenance activities fall broadly into four headings. These include the policy implemented by the plant manager including the balance of maintenance activities and the clearing of backlog activities, the planning and scheduling, the procedures, and the conduct of the maintenance (IAEA-TECDOC-1098, 1999; Table 4.4).

Good availability and reliability are the key objectives. Meeting these objectives requires adequate resources to be available to predict the need for the necessary maintenance, to prevent unnecessary activities, and to ensure maintenance is carried out correctly. The majority of work has to be performed during outages and plant availability depends on it being carried out efficiently and successfully. Unplanned outages should clearly be avoided. In nuclear plants today, computer scheduling systems are used to co-ordinate activities and ensure that adequate materials (spares) are available.

Clear maintenance back-logs

Apply maintenance performance indicators

Readiness for eventual unplanned outages

Employ reliability and condition-based decision analysis

Advanced planning of routine maintenance and outages

Use plant-approved procedures in conducting maintenance

Post-maintenance testing to verify satisfactory completion and restart readiness

Use PSA for planning on-line maintenance

IAEA-TECDOC-1098 (1999).

Factors that enhance the effectiveness of maintenance management identified in IAEA-TECDOC-1098 (1999) include, ensuring that maintenance backlog actions do not accumulate and that indicators of various different maintenance activities are recorded. Examples of such indicators include the number of requests, the distinction between preventive and repair maintenance activities and the number of repetitions of work on the same plant components or systems.

On-line monitoring of equipment together with reliability and condition based decision analysis help to reduce preventative maintenance. Preparedness for routine maintenance and outages may be facilitated by advanced planning using mock-ups and other practical demonstrations. These should allow for unplanned outages and should include lists of the different maintenance activities that can be carried out. Regarding technical matters of plant operation, there must be careful monitoring of foreign material in the plant in order to protect equipment. Maintenance should be performed in line with approved procedures, including PSA to facilitate on-line maintenance. Finally, post-maintenance testing must be carried out to verify satisfactory completion of work and to confirm the readiness for the plant restart.

For many of the facilities that are currently being decommissioned, little attention was given to decommissioning in their design (Review of Radioactive Waste Management Policy, 1995). This has resulted in an increase in costs in some cases. In the UK for example, the regulator now requires that consideration be given to decommissioning in the design of a plant. This is in regard to a number of factors, construction techniques, choice of materials, the provision of suitable access and the availability of adequate waste storage facilities.

The costs of decommissioning for different reactor types and different countries were considered in an IAEA review of selected cost drivers for decisions on the continued operation of the older nuclear reactors (IAEA-TECDOC-1084, 1999). This review covered pressurised water reactors (PWR and VVER), BWRs, HWRs, light water cooled, graphite moderated reactors (LWGW or RBMK type) and gas reactors (GCR and AGR).

Two categories of decommissioning costs are considered. The first category (Stage 1 and/or Stage 2 decommissioning followed eventually by Stage 3) is decommissioning with long-term storage. This takes advantage of the natural decay of the radioactive isotopes, which makes dismantling operations at a later time much easier. The second category is the decommissioning approach with immediate dismantling of the plant up to the ‘green­field’ (non-restricted use) or ‘grey-field’ (somewhat restricted use) condition (Stage 3 decommissioning). This structuring of the decommissioning stages is based on the well — established IAEA terminology.

It is noted that decommissioning practices differ substantially from country to country and this affects any comparable cost estimates. For example in some countries, the cost of fuel unloading is included as a standard part of decommissioning costs. In most countries, it is not. There is not necessarily a consistent practice within a particular country. The study in IAEA-TECDOC-1084 (1999) aimed to focus on total costs and made no attempt to consider the relative importance of various cost components.

Section 6.11.1 summarises the estimated costs of decommissioning after storage, for the principal types of reactor in operation at the present time. Data are taken from IAEA-TECDOC-1084 (1999). It should be noted that the costs considered were total costs excluding discounting. It should further be recognised that not all costs in the data were normalised to exactly the same time period.

There may be differences in the regulations and standards for future reactor licensing. In countries favourable to nuclear power, the licensing frameworks are likely to evolve or be extensions of existing frameworks for currently operating plant. However, there are a number of countries that are not favourable towards nuclear power where the approach will be dependent on the perception of relative risks to benefits. There are also a number of countries where no new plants are planned or where moratoria are already in place in which case there is no issue. Within Europe for example, five out of the eight EU member

Table 8.6. Standards for future reactors (in countries where there are not moratoria)

Current regulations and standards are likely to be appropriate for evolutionary type plant Current regulatory regimes may need to be extended for more revolutionary type plant Future reactors are expected to include greater protection against severe accidents. Increased level of PSA

Increased use of BE methods to demonstrate more realistic safety margins Increased use of Risk Informed methods following USNRC lead

EUR 20055 EN (2001).

states with nuclear power are in this category. These include Belgium, Germany, Netherlands, Spain and Sweden.

The standards for future reactors will depend on internationally accepted standards, e. g. as shown in Table 8.6. Some countries believe that their current regulations and standards are already appropriate for future evolutionary LWRs. There are, however, certain requirements that are likely to be imposed which assume greater significance for future reactors.

It seems likely that future reactors will have to include greater provision against severe accidents. This may be required to be demonstrated by both PSA and deterministic means.

PSA methodology is used currently on existing plant to identify weaknesses and therefore enable modifications to be implemented. PSA practices have improved significantly over recent years; these methods provide an accepted means of assessing the safety of a plant. PSAs provide a means of verifying the design basis of a plant supported by deterministic analysis making conservative assumptions. They also can be used to assess whether there are any ‘cliff-edge’ concerns about safety with the design, e. g. just beyond the design basis. This has led to modern LWR designs, which restrict the source term for radioactive release for beyond design basis, including severe accidents.

There is certainly a trend in some countries to extend the design basis for new plants to cover severe accident challenges. However, this could clearly have a major impact on competitiveness of the plant, through the cost of including specific or additional components.

From a severe accident perspective in LWRs, the strength of the containment is crucial in limiting radioactive release. However, at present, the containment is only built to withstand DBAs where safety systems are assured, and assumed to respond subject to a single failure criterion. Present analyses demonstrate a margin between the design pressure and the actual failure pressure and this is useful in evaluating the implication of certain severe accidents. However, if all severe accidents were included within the design basis envelope, then the containment would need to be strengthened to withstand higher loads.

There are differences worldwide in the approach to containment and its function in severe accidents. For example, within the IAEA member states, some countries have already made significant improvements, others have plans for improvement that have not yet been implemented, others prefer to adopt a different approach to severe accident management.

The extension of the design to include severe accidents has been proposed in Germany. However, at the time of writing, there is a consensus to terminate the use of nuclear energy over the next 30 years and therefore standards for future plants are no longer under consideration. Other countries, e. g. France are taking a similar position. If such proposals are adopted, there could clearly be major differences in approach to licensing across the nuclear operating countries.

In many cases, the evolutionary designs contain more advanced safety features, some of which already mitigate against severe accident vulnerabilities. The EPR design, for example, has a debris retention component. Some VVER-1000 reactors’ future designs will adopt a similar approach. Thus, the addition of core catchers to prevent melt attack of the containment base-mat is one feature that has already been introduced into the design to cover severe accidents.

It would, however, be very difficult to extend the design basis to cover all potential severe accident scenarios. Steam explosion vulnerabilities are still uncertain and it would be difficult to demonstrate by deterministic means that a containment is sufficiently strong to withstand all possible loadings, taking account of the uncertainties.

Other approaches on design have been to take advantage of more inherent mechanisms, passive injection, gravity driven flow, e. g. as in AP-600.

A chapter is devoted to passive plants later in the book.