India has a vibrant nuclear power programme with currently 14 units in operation, 9 under construction, and more new reactors planned (Table 9.3). There are 5 research reactors in operation (World Nuclear Association, 2003). Currently, nuclear power supplies less than 4% of the country’s electricity requirement. There is a target to reach 10% in 2005. Capacity factors are now much improved compared with a few years ago, reaching 85% in 2001-2002.

The Tarapur plants are increased capacity plants based on domestic technology and are expected to begin operation in 2004-2005. The other PHWRs will follow later; the Rawatbhata units are scheduled to be in operation by 2007. The design for future PHWRs, the first of these are likely to be Kakrapar 3 & 4, has now been raised to 680 MWe.

Two large VVER-1000s are being built by Russia. The first unit is forecast to be commissioned in 2007.

The construction of a 500 MWe fast breeder reactor is in progress at Kalpakkam. This is contributing to the government’s objective to utilise India’s abundant thorium resource as a nuclear fuel. The intention is for this reactor to be operating in 2010. This reactor will be fuelled by uranium-plutonium-carbide fuel. The plutonium resource would come from currently existing PHWRs.

The intention is to develop an advanced heavy water reactor (AHWR) thorium cycle based of the following route. Existing PHWRs will burn natural uranium to produce plutonium. Fast-breeder reactors of the type above will then burn plutonium and breed U-233 from thorium. AHWRs will then burn the U-233 with thorium. The first AHWR is due to start construction in 2004.

There are plans to build a mix of reactor types to meet India’s requirements. The forward plan is to have a 300 MWe AHWR together with a mix of 500 MWe FBRs, 680 MWe PHWRs and 1000 MWe LWRs by 2020.

India’s civil nuclear strategy is to achieve complete independence in the fuel cycle. The country has a fuel fabrication facility at Hyderabad for PHWR and BWR. There are also spent fuel and reprocessing plants at Trombay, Tarapur and Kalpakkam. There is a waste immobilisation plant and storage facility at Tarapur.

Research is in progress in setting up a deep geological repository for high-level wastes.

The Indira Gandhi Centre for Atomic Research at Kalpakkam is working on fast reactor technology development. The Bhabha Atomic Research Centre near Mumbai is working on thorium-based systems. In particular, the Centre is working on the AHWR. In addition, India is also developing accelerator-driven systems for driving sub-critical reactors.

The PBMR design has been put forward by the South African Utility, ESKOM in partnership with an international consortium. It also meets Generation IV design objectives in that it includes passive safety features to meet public acceptance criteria and offers competitive economics. The units are relatively small at 110-120 MWe with good economic and safety characteristics. The PBMR is also flexible in that it can be built virtually anywhere. It operates with a direct Brayton thermo-dynamic cycle, with target efficiency of around 45%. In principle, it can also use a thorium fuel cycle as well as a traditional uranium cycle. The design is modular in order to enable an operating utility to match the size of his station to the demand. The present capital cost is estimated at about $1000US per kWe, the construction period is estimated to be very short at around 2 years.

The PBMR offers a potential complementary service to the energy market in terms of present plant capabilities as both an electrical and non-electrical energy generator. It is of medium size, comparable with current-sized gas plants. It could offer a capability for the co-generation of heat or even dedicated nuclear heating applications, as expanded below.

The PBMR design is based on the HTR-MODULE design previously licensed in Germany for commercial operation. Present activities are aimed at the engineering design, independent safety reviews by participating countries in the ESKOM project and in making provisions for the licensing process.

The HTGRs have desirable features from various safety perspectives. The cores have a large thermal inertia, low power density and a strongly negative Doppler reactivity coefficient. As for most reactor types, the transients can be categorised into two broad categories, reactivity-initiated events and loss of flow events, either with or without depressurisation. For an un-scrammed core heat-up, the maximum core temperatures are reached within 3 days but fuel temperatures do not exceed above 1600°C ensuring that fuel particle integrity is preserved.

One concern with HTGRs is that air could ingress the core resulting in oxidisation of the graphite. This would require a multiple failure scenario of ruptures in the pressure vessel and surrounding concrete. However as noted above, even if such events occurred, there would still be several days to breach the opening of the reactor vessel.

A considerable advantage of gas systems described above is that they are free from the usual problems associated with loss of cooling in LWR systems. Thus there are no phenomena of concern such as ‘Departure from Nucleate Boiling’ loss of heat transfer or ‘Pellet Clad Interaction’ failure.

The reactor has diverse and redundant safety systems. For example, the reactor can be shutdown by three independent control systems. Each system is sufficient in itself to achieve this requirement.

In summary, in addition to electricity generation, HTGRs are being proposed as candidate plants for process heat applications that require high-temperature conditions. These include hydrogen and methanol production in a steam reformer, a process that requires high-temperature heating of steam and methane. Steam could be produced and then utilised for processes such as coal densification and steam injection for the recovery of hydrocarbons. These plants would also be suitable for de-salination processes, which require low-temperature heat. There may be potential to take waste heat from the pre­cooler that would otherwise be wasted. The ways of operating HTGRs in these multi­generation modes would add significantly to the thermal efficiencies that would be achievable with the plant.

The PFBR is a pool-type sodium cooled reactor under design in India (IAEA-TEC — DOC-1083, 1999). It is a 500 MWe medium-sized reactor and extrapolates from the FBTR

13.3 MWe experimental reactor that has already been successfully commissioned.

The fuel consists of mixed plutonium and uranium oxides and depleted uranium is used as the blanket. The fuel region includes two zones of different plutonium oxide enrichment. Secondary side shielding is included in both the axial and radial directions.

There are nine primary control and safety rods for setting the power level and for shutting down the reactor. There are in addition three diverse safety rods.

The primary circuit consists of two pumps and four IHXs, with one IHX on either side of each pump. The secondary sodium system consists of two identical loops each comprising of two IHXs and three steam generator modules.

There have been fewer application for desalination than for district heating. As for the latter case, the majority of applications have been with plants operating in co-generation mode, i. e. electricity and desalination. Desalination plants have been operated in Japan (Ikata, Ohi, Genkai, Takahama, Kashiwazaki). A range of different desalination processes have been used. There has also been some experience from a plant operating in the USA at the Diablo Canyon.

Other experience has been gained in Kazakhstan (Aktau) where the liquid metal cooled fast reactor BN-350 has been operating as a multi-energy source for electricity, drinking water and heat.

A non-nuclear facility was built in Israel for testing the nuclear desalination process. The heat source was produced by an oil-fired power plant N. B. this operated for only a short period.

Desalination is the process of obtaining freshwater suitable for drinking or industrial processes through the removal of salt from saline, usually seawater. This can be achieved using either distillation processes or via membrane processes using osmosis (IAEA-TECDOC-1056, 1998). Desalination processes include:

— Multi-stage-flash (MSF) distillation;

— Multiple-effect distillation (MSD);

— Reverse osmosis (RO).

Typical energy requirements and energy consumption rates for the three processes are shown in Table 14.4. These can be compared with the theoretical minimum energy requirement of 0.73 kW h m_3 for 35,000 ppm saline water at 25°C. The discrepancies are due to significant thermal processes and irreversibility that occur during the separation process.

Reactor physics related data are available in international data banks, e. g. the International Reactor Physics Benchmark Experiments (IRPhE) databank of the NEA and various activities are co-ordinated through the Nuclear Science Committee (NSC) (NEA Annual Report, 2002). The data are used as a reference for transient analysis to address specific safety issues. For example, a safety issue for PWRs concerns thermal mixing and the impact on neutronics. In 2002, a main steam-line break (SLB) benchmark study was carried out using data from the TMI-Unit 1 PWR. Coupled 3D neutronics/thermal- hydraulics calculational methods were used. For BWRs, reactor stability is an issue; benchmarks in 2002 have been performed based on US BWR-4 reference experimental data. Russian-designed VVER-1000 reactors have also been the subject of recent benchmark studies.

The NEA Data Bank services its member countries in regard to many requests for experimental and bibliographical nuclear data. Important improvements and updates are carried out through the joint evaluated fission and fusion (JEFF) activities. Data are available for ~ 340 different isotopes or elements including thermal scattering data for five lattice structures. The NEA is co-ordinating international collaboration among the major global nuclear data evaluation projects.

Experimental reactors are available for determining neutronic characteristics of reactor lattices, although the number of facilities is much less now in the world than hitherto. The EC is supporting the RENION project (Vasa, to be published) for carrying out PWR and VVER reactor physics related experiments.

The demand for energy is closely driven by economic growth. There are, therefore, significant differences across the global sectors. Data provided in Energy Visions 2030 for Finland (2003) show the emergence of countries such as Asia with large developing economies where the regional share of worldwide energy was 25% in 1981 but rising to 37% by 2005. Growth rates in Asia have been higher than other sectors since 1993, at about 4.8% between 1993 and 2000 and forecasted to exceed 4.5% between 2000 and 2005.

International Energy Agency (IEA) data forecast an average global annual growth rate of 3% over the next 20 years. This equates to about a 57% growth of primary energy requirement over this period. The main increase in demand will come from the developing

countries. This demand is likely to be met from their indigenous resources of fossil fuels together with additional imported energy resource to meet demand. The fossil fuel share could be as high as 90% by 2020 unless this additional resource can be supplied by other means, e. g. nuclear, hydropower or possibly renewables.

Another forecast for the EU is little different (EC Green Paper, 2000; European Energy Strategy, 2001). The distribution of total energy consumption across the EU for the various sources is shown in Table 17.1. To meet this demand, Europe currently imports about 50% of its requirement, and this would rise to 70% in 2030 if current trends continue. Without new build of nuclear plants, the nuclear component would drop from 15 to about 6% in 2030, the European energy sector would become much less autonomous and without a significant increase in renewable energy, carbon dioxide emissions and global warming would increase.

At the present time, there is a general decline in many areas of support for the nuclear industry. Research and Development programmes have been particularly affected, certainly in the Western world. The reasons for this are that the knowledge base to support currently operating plant is at a relatively mature state and the lack of new building programmes means that little new work is needed. In addition, nuclear energy is having to compete with other forms of energy producers in the market. The nuclear business is not seen as a popular business in which to work. The net result is a significant reduction in resource due to a failure to attract new graduates into the industry, a failure to keep new people in the industry and the loss of people in retirement.

It has been recognised for some time therefore, (Storey, 2001) that there is a requirement to maintain technical competence, not only to ensure safe operation and decommissioning of existing plant, but also to be available in the future, if new reactors are required. The continued operation of existing plant does provide a means to ensure some level of competent resource is maintained for both operation and regulation.

Regulators are focussing on a number of areas through the NEA described by Storey (2001). Specific problems have been considered by the Committee on the Safety of Nuclear Installations (CSNI)’s Senior Expert Group and more recently by the European Commission through its research and training programme (RTP) in the field of nuclear energy. The Senior Expert Group has made recommendations for research in a number of important technical areas in the OECD Community. These include the maintenance of a major thermal-hydraulic rig for each reactor type, for fuel and reactor physics facilities, for research on the integrity of equipment structures and for the continued availability of hot cell and research reactor facilities. EC initiatives include the creation of centres of excellence (COEs) for severe accidents and for fission products expertise; the setting-up of databases on seismic activity and support of other areas in respect of human factors and plant monitoring and control.

Other initiatives are moving forward under the auspices of the NEA Committee of the Nuclear Regulatory Authorities (CNRAs), which are more general (including non­technical topics). Some of these are aimed at maintaining safety competence in the industry and the regulator. The NEA and the EC in its RTP programme, referred to above, are also addressing nuclear training and education.

Thus, maintaining a sufficient degree of overall competence is a particular issue in the nuclear industry at the present time (BNIF/BNES Conference — Energy Choices, 2002). As nuclear power is declining internationally and particularly the lack of ‘new build’, there are problems with the retirement of suitably experienced and qualified (SQEP) staff and difficulties in recruiting high-quality personnel into the nuclear industry. In countries where there are continuing nuclear programmes, there is at least a steady stream of work to support plant operation so some capability is maintained.

The advent of deregulation in many countries brings with it some form of commercialisation and competition in the way that its nuclear plants are now being operated. Part of this culture is to recognise that change becomes a normal way of life. If a nuclear plant utility is to succeed, then the management must recognise and be capable of managing change. Change also brings with it an increased risk, as performance targets are set to stimulate production and new and innovative courses of action are encouraged in the management team to achieve these targets.

A number of important characteristics applying particularly to a change in management system have been defined in IAEA-TECDOC-1123. These include the following, some of which were mentioned earlier.

Organisational changes must be communicated by managers in a way that all levels of staff can understand and accept. Limits of authority must be clearly set. Critical performance variables must be monitored and systems should be in place to do this. Advantage should be taken of latest information and technology (IT) systems to facilitate good communication and feedback to management. There must also be sufficient internal controls and audits to confirm that procedures are being performed satisfactorily.

The condition of inventory will need to be managed to meet radiological, environmental and possibly other safety concerns.

For example, there may be chemical corrosion processes that affect the handling of fuel downstream in the disposal route. The timescale of these processes could impact the timing of certain operations depending on whether a corroded or an uncorroded state of the inventory is easier to manage. The gaseous chemical products of reaction may also be a concern, e. g. in Magnox plants, the Magnox/water reaction produces hydrogen (Twidale, 1999).

It may be possible to dilute liquid inventories as a means of reducing the specific radioactivity of the liquid. This could provide significant benefits in dose management of the work force. Further, by appropriate chemical treatment, it may be possible to reduce the impact on the environment.

Repackaging of the inventory into a safe form to meet all the necessary safety requirements is likely to be necessary. Interim storage is likely to be adopted in most countries where the approach for long-term storage, e. g. in a repository, has yet to be agreed.

Decommissioning safety risk is primarily associated with risks associated with public health and safety and the risks associated with waste management (de la Ferte, 1996). All OECD countries with nuclear programmes have in place decommissioning regulations, either as part of their general legal infrastructures for nuclear plant licensing or specifically for decommissioning. The IAEA have also set down the general principles to be followed, and defined the respective responsibilities for regulator and operator.

National licensing procedures define whether the operator or public authorities are empowered to decide on shut down and decommissioning of facilities. There are some differences between countries in terms of responsibilities. In the UK and Germany, the responsibilities for the shut down and decommissioning of facilities lies solely with the operator under normal circumstances (Willby, 1996). In other countries such as France, the operator has less independence. There have also been instances where governments have taken a political decision to shut down plant as in the moratoria imposed by Italy and Sweden. The body that has the responsibility for decommissioning operations is also different in different countries, in Canada for example, it is the operator; in Belgium and Spain there is a specialised public agency responsible for radioactive waste management.

In the UK, the HSE has set down policy issues and broad requirements on the licensee (Bacon, 1997). These cover requirements on the licensee in regard to defining strategic plans, work plans, and schedules and priorities for the progressive reduction of hazards (Walkden and Taylor, 2002). These requirements are summarised in Table 6.6.

EC legislation exists to control the environmental impact of decommissioning activities (Nash and Woollam, 2002; Statutory Instrument No. 2892, 1999). This requires the provision by the operator of an environmental statement to support decommissioning activities. UK regulations have a similar requirement.

In general, operations must be managed to ensure that any hazardous releases are prevented as far as is reasonably practicable. Any plant conditions that have a potential for uncontrolled releases must be dealt with as soon as possible. Any discharges that do occur must be within agreed limits. If a potential to exceed these limits is recognised, e. g. prior to defuelling, then early decommissioning may be necessary (Twidale, 1999), or authorisations may have to be renegotiated.

In regard to different national practices, authority may be administered at either a national or local (federal) level depending on the laws of the country concerned. There are differences in approach reflecting differences in local legislation (EUR 16801 EN, ISSN 1018-5, 1996; Table 8.2). In the UK, the licensing process for power plant lies with a single licensing authority, the NII. However, in Germany for example, regulatory responsibilities lie with the Ministries of the Federal States. In addition, the licensing of different sectors of the nuclear industry may be administered across several organisations.

There are significant differences across countries on how regulatory technical support is acquired. Few authorities have sufficient in-house technical capability to meet all their requirements within their own organisation. Certain authorities require work to be contracted out to other organisations. In some countries, technical support organisations

(TSOs) are supported within the countries’ national safety framework. This is the case, e. g. in Germany and France. In other countries, the regulator may call on various contractors; (usually privately owned or commercial companies) to supply his technical need for the particular work required. This has implications on how an adequate level of technical resource is maintained and indeed on the number of personnel available. There are also substantial differences in the levels of resource (both licensing and technical) available within regulatory authorities.

In the past, the approach to nuclear regulation has varied between national governments and their political persuasions. For example, in the countries operating Russian-designed plant under the former Soviet regime, the safety culture was very different from that of the West. The differences of approach are now much less marked as all these countries move towards common licensing approaches.

There has been a major global influence of US practices in many countries. This is because US-designed plants are in operation in many countries around the world. It has also been a common principle from many regulators that NPPs should be licensable in their country of origin.

In addition, there has been considerable influence from IAEA principles, (Govaerts, 1996; IAEA Safety Standard Series, 2000), which have promoted the safety culture and co-operation between regulator and operator. These cover commonly accepted principles such as operator responsibility for operation and quality assurance principles agreed between regulator and operator, etc.

As a general rule, the investment in nuclear safety has far exceeded that which has been devoted to other industrial operations. As a consequence, the safety standards in the nuclear industry are as high or higher than those existing in many other industries.

Clearly, the licensing of future plants may require some adjustments of regulatory approach. For example, an application for the licensing of a Generation IV system would introduce new safety issues associated with new technology. The further the new system had advanced from existing technology, the greater the regulatory adjustment that would be required.