Finally, there are some possible very high-temperature applications that could be conceivable with the VHTR designs that are the end objective of Generation IV. These
include iron manufacture, cement and even glass process applications (Institute of Nuclear Engineers, 2004).

Advanced clads are being developed to exhibit better corrosion, mechanical properties and reduced growth under normal operating conditions. The models are under review for application to different cladding materials. As noted above, the clads may also experience different loads from newer fuels, e. g. MOX fuels compared with more traditional uranium oxide fuel.

16.3.1 Transient Fuel Rod Codes

Transient codes have been developed that include not only the physical models of the steady-state codes but also include additional modelling for transient thermal behaviour, heat capacity and heat transfer, transient mechanical properties such as long-term creep, cladding plastic stress-strain phenomena and ballooning at high temperatures. Other effects such as the effects of annealing, oxidation and hydriding, and changes of phase will also be modelled. Examples of transient fuel rod codes are the EPRI codes, FALCON/FREY, FRAPTRAN and the French IRSN code SCANAIR (IRSN: Scientific and Technical Report, 2002).

The main purpose of the transient fuel codes is for analysing the fuel rod response for RIAs and LOCAs. The main issues in modelling are related to the time-scales of different transient phenomena in relation to the time-scales of these transients. For example, fission product release may occur on both short-term and long-term time-scales. The time-scales of non-transient swelling and axial growth are much longer than the above accident transient time-scales. Different codes include different modelling assumptions, e. g. in addition to modelling pre-failure fuel behaviour, some of the codes include rod failure models.

At higher burn-ups, for example greater than 40-50 MWd kgU1, the Rim zone in the fuel requires special modelling attention. This is to make sure that the degradation in fuel thermal conductivity caused by structural changes in the fuel in this region is correctly modelled. Further differences in modelling requirements may also exist for MOX fuel and for advanced clads compared with more traditional UO2 fuel and clads.

Regarding other reactor types, HTR technology was under development in the 1980s but is now believed to be a realistic alternative to LWR. Ceramic fuel technology has been

established but further research is required to ensure that fuel performance is sufficiently reliable at high temperature (Hesketh, 2001). Current research is focussing on the fuel manufacturing process but methods will need to be developed to demonstrate that the fuel will be reliable to its design discharge burn-up.

Modelling codes for liquid metal fast reactors have been developed in various national programmes (IAEA-TECDOC-1083, 1999). The principal codes are TRAFIC (UK), GERMINAL (France), SATURN-TRANSIENT (Germany), LIFE (US), CEDAR (Japan) and KONDOR (Russia). These codes have a reasonable validation for moderate levels of burn-up (less than 12-15 at.%). The codes predict fuel pin thermal and mechanical behaviour for oxide fuels in steady-state and transient conditions. Some of these codes, e. g. TRAFIC also describe the behaviour of fuel pins after failure.

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.

10.1. INTRODUCTION/OBJECTIVES

The purpose of this chapter is to describe briefly the evolutionary water reactor designs that have evolved from current generation commercial reactors. These evolutionary designs have been developed during the 1990s, taking advantage of lessons learned from existing plant. The chapter focuses on water reactor systems because these occupy the dominant position among the evolutionary reactor designs that are currently under consideration for building in the short term. Other types of advanced reactors are considered later in the book. There is no attempt to describe all possible designs in detail. Rather the approach is to categorise the various designs into different types and then describe the representative features of the reactors within a given type. This enables the reader to understand the general design features that are currently being put forward. References are given for the comprehensive range of reactor types.

Various evolutionary improvements have been proposed for all the major water reactor types currently in operation, i. e. PWRs, BWRs, and HWRs. Common general features are simplification in design to reduce cost, coupled with increased safety features. Many of the designs are available at different power capacity ratings, from medium size, e. g. ~ 500-600 MWe range, through to 1000-1300 MWe range. These have been put forward to provide more flexibility to meet the current market demand but also have evolved to meet perceived changes in demand. There was a trend in the mid-1990s to produce medium-range designs to take advantage of increased passivity in design. However, the economics of larger plants are now thought to be more favourable, and present trends are more towards the larger plant scale. Further it has been shown that the medium-sized passive designs can be scaled up.

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.