There are some features of evolutionary plants that require new models and extension of the codes that have been developed for present day plants.

Some integral codes have special models that have been developed for particular plants. They therefore cannot be applied or it is difficult to apply them to new plants. It is also difficult to apply them to new experiments for the purposes of code validation. The MAAP code developed by EPRI is an integrated severe accident code, which has specific models for specific plants and phenomena. A special version AP600-MAAP has been developed for evaluation of AP600 safety (IAEA-TECDOC-752, 1994).

In general, in advanced LWR designs, there is a requirement for much stronger thermal-hydraulic coupling between the primary circuit and containment. This has led to the coupling of some system thermal-hydraulic codes, e. g. RELAP5 with the containment code CONTAIN.

Many evolutionary passive designs have large pools as heat sinks and condensers. To be effective, these need to be well mixed and the effectiveness of these pools needs to be established. The system thermal-hydraulic codes do not have the required mechanistic mixing models and therefore need to be benchmarked against CFD codes.

The system codes have limitations in their modelling of condensation, particularly in the presence of non-condensables or 3D effects.

Finally, it has been established that the performance of the system codes in buoyancy-driven situations is less robust, than in their application to the modelling of high-pressure forced convection flows, the regimes for which they were originally developed. Much effort has been expended in improving the performance of these codes in low-pressure applications in current generation reactors, e. g. in the modelling of shutdown accidents. Generally, later versions of the system codes, e. g. RELAP5 are much more robust (compared with earlier versions in this respect).

In addition to technical and economic factors, there are undoubtedly political issues surrounding the future operation of nuclear plant There is the issue of public or stakeholder confidence. There is not simply the question of the safety of nuclear reactors where only a few significant accidents have occurred, but no longer can the nuclear industry claim that severe accidents are incredible. In regard to normal operation, there are also public concerns on waste management issues, about the fuel cycle and on the issue of proliferation. These issues are reviewed in this section.

Reduction in outage time has been achieved in a number of plants that have been operating for some years through both technical and administrative improvements. Another factor has been the introduction of more computerised systems to aid in the planning and managing of outages. Technologies for improving light water operation and maintenance,

including outages, have been published in IAEA-TECDOC-1175 (2000). Figure 4.4 shows the global reduction in average planned outage time that has been achieved over the last decade.

The fuelling scheme clearly has an important influence on outage planning. For example, there is a general tendency towards longer fuel cycles. There are obvious requirements on the load design, the most important of which is safe operation, e. g. limits on shutdown margin, and ensuring negative moderator temperature reactivity coefficient. Once safety constraints have been satisfied, outages may need to be synchronised in order that there is a period of time between outages of different units at the same plant. Clearly outages should be avoided at times of peak demand, e. g. during winter periods, etc. Finally, outages should be planned in order to optimise fuel economy and cost.

Outages may be of different duration depending on the work planned. In-service rules may require complete in-service of the reactor vessel including complete unloading of the fuel, the removal of reactor internals, inspection of the reactor vessel at regular intervals, e. g. every 4 years. Shorter outages would be used to carry out a less ambitious programme. By way of example, in the Paks plant in Hungary, the new outage strategy includes outages of short (25-30 days), normal (30-35) and long (55-60 days) duration.

Another example (IAEA-TECDOC-1175, 2000) where a different technical approach has resulted in improvements of outage time concerns the Indian Point 2 reactor in the US. Here two safety systems modifications have been implemented, replacement of conventional hydrogen ignition systems with passive auto-catalytic hydrogen recombiners and the replacement of the conventional containment spray additive tank with baskets of tri-sodium phosphate. The original systems consisted of several hundred components which required significant maintenance and testing, the new passive systems require much less.

For the future, ways to reduce outage are being considered at the design stage, e. g. in the advanced European pressurised reactor (EPR) design. These include features such as improved accessibility of the reactor building during operation, and better logistics support including the need for special tools and availability of spare parts. Some of these approaches are also being considered for current plant.

Utility requirements’ documents have been produced which aim to provide direction to designers by taking advantage of experience from current plants. The aim is to reduce costs and uncertainties of licensing by demonstration of a sound technical basis for advanced designs.

EC activities have been performed over the last few decades to meet the requirements of several resolutions (EUR 20055 EN, 2001). These were the Council Resolutions of 1975 (European Commission, 1975) and 1992 (European Commission, 1992).

The 1975 Resolution stated that European Community actions in the area of nuclear safety were necessary because of the importance of nuclear power as an energy source in the Community, the need for the Community to address the technological problems of nuclear safety in view of possible environmental and health implications, the need to keep the public informed, to realise the safety and economic benefits of a harmonised approach for nuclear safety authorities, constructors and producers, and the desire for the Commission to influence global nuclear safety.

The 1992 Resolution not only acknowledged the continuing importance and relevance of the earlier resolution but also recognised some additional requirements. For example, the Council reaffirmed the importance of progress, nuclear safety research and innovation including future generations of reactors, but recommended that experience gained should be extended to third countries, particularly those of Central and Eastern Europe and the republics of the former Soviet Union.

Since 1995, there have been further developments, including the 1995 Consensus Document on European LWR safety, the publication of recommended licensing procedures, a document on the implementation of the 25 Principles of Nuclear Safety in different EU countries and the Convention on Nuclear Safety by EU member states and the EC. The EC is encouraging the spirit of common approaches to nuclear safety via dialogue and the synthesis of information. Nevertheless, at the present time, there are no Euratom

Treaty obligations on the EU member states to harmonise their nuclear safety criteria and regulations.

The EC Euratom framework programme for future reactors is linked with the national and international programmes above (Ion et al., 2003). Additionally, there is a major investment in fusion. With regard to fission reactors, the EC MICANET objective is to provide an R & D strategy to enable the nuclear option to be kept open via the development of innovative systems. Euratom also participates within the GIF project.

There are various measures that are included in advanced containment design for the control of pressure and temperature. The objective is to limit the pressure to below an acceptable limit and to reduce the pressure down to atmospheric pressure as quickly as possible to limit fission product release to the environment. The systems should be passive so that they can still function reliably under severe accident conditions. The heat loads can arise from decay heat and in the event of a severe core meltdown from the Zircaloy/steam exothermic reaction and also possibly from MCCIs.

11.5.1.1 Passive Containment Cooling. This method is used with a steel containment with good heat transfer characteristics. Pressure reduction using this system will be relatively slow and will also depend on the partial pressure of non-condensable gases in the containment. External cooling is enhanced by external sprays.

The AP1000/600 containments comprise an inner steel containment shell surrounded by an exterior concrete shield building, Figure 11.1. The inner steel containment not only acts as a barrier to radioactive release but also serves as an integral part of the heat release system. It is prevented from over heating by a PCCS that provides a natural circulation draught of air cooling between the steel containment shell and the shield building (Scobel and Conway, 1990). This serves to enhance the heat removal from the PCCS.

A similar approach is adopted by the simplified PWR (SPWR) of Westinghouse — Mitsubishi design. This design has been scaled up to 900 and 1200 MW units (Lillington and Kimber, 1997; Naitoh et al., 1992).

Other conceptual designs (Lillington and Kimber, 1997; Kuczera, 1992) for example have been put forward by KfK, which consist of an inner steel containment surrounded by a strong re-enforced concrete wall. Both the inner steel and the concrete walls share the loads. There is an annulus between the two shells through which air flows by natural circulation.

11.5.1.2 Condenser Systems. In this system, heat is transferred to the external atmosphere via an intermediate circuit, which carries single — or two-phase water under natural or active system circulation. As for the PCCS described above, the effectiveness of heat transfer will depend on the partial pressure of non-condensables inside the containment. Pressure reduction is also relatively slow.

In the EP1000, a finned condenser is located at the top of the concrete primary containment. This transfers heat via a thermosyphon loop through the concrete containment walls to an external heat exchanger located in a tank. This is initially immersed in water but later in the accident is air-cooled (Yadigaroglu et al., 1998).

The cooling of the containment atmosphere by a condenser is also proposed for the SWR 1000 design. This transfers the heat to a secondary system connected to an external pool.

In the CANDU 6 system, a containment condenser transfers heat to a secondary side connected to the Passive Emergency Water System Tank (Hopwood, 1999).

In the ESBWR (Orsini and Pino, 1992), a PCCS is incorporated into the design of the containment to remove decay heat from the drywell (Figure 11.2). In this system, containment steam is condensed in an external pool. Non-condensables are discharged to the suppression pool.

11.5.1.3 Internal Containment Sprays. Internal spray systems may have both significant advantages but also disadvantages. The present designs tend to have active components and are, therefore, susceptible to not functioning in a hostile environment. Passive systems have been considered but have reduced capacity and may not function
correctly in the presence of aerosols. Sprays will also not reduce the pressure if there is a significant partial pressure of non-condensables, e. g. hydrogen from metal water reactions or other gases from core concrete interactions.

11.5.1.4 Sump Water Cooling. Systems to reduce the containment pressure by cooling the sump water are another possible method. However, there needs to be good natural circulation cooling of the sump water which is necessary for effective heat removal.

During the spallation process, the collision between the energetic particle and the target nucleus leads to direct reactions referred to as intra-nuclear cascade. In this cascade, small groups or individual nucleons (protons and neutrons) are expelled from the nucleus. At energies above a few GeV per nucleon, the nucleus can fragment. After the intra-nuclear cascade, the nucleus is in an excited state and subsequently releases ‘evaporates’ nucleons, mainly neutrons.

The spallation process is complicated and depends on the target thickness and the target materials. For thick targets high energy (> 20 MeV) secondary particles may take part in further spallation reactions. For some target materials, low energy (< 20 MeV) neutrons produced from cascade evaporation, can enhance neutron production. For heavier nuclei, high-energy fission may compete with evaporation. Examples of materials that undergo spallation/high-energy fission include lead, tantalum and tungsten. Some spallation target materials, e. g. thorium and depleted uranium may be fissioned by both high — and low- energy neutrons. Regarding target particles, deuterium and tritium produce more neutrons than protons in the below 1-2 GeV energy range but the low-energy part of the accelerator tends to get contaminated, resulting in higher maintenance costs.

The requirement for an ADS target is to convert a high-energy particle beam to neutrons at low energy. It is desirable for it to be of compact size, to couple to a surrounding blanket, operate in the 10-100 MW power range, and have high neutron production efficiency. Other requirements, in common with other nuclear devices, are that it should be reliable and of low cost, be safe and generate only a small amount of waste. Molten lead is a good choice for meeting these requirements. Lead-bismuth eutectic has also been considered because of its lower melting point than lead, but this eutectic produces polonium, the release of which may be a problem at high temperature.

The blanket (sub-critical assembly) surrounding the target multiplies the spallation neutrons for the transmutation of the minor actinides (MA) and LLFP. Taking account of many aspects, safety, operations, material cost and incinerator costs, keff values for the target in the range 0.9-0.98 are typically considered.

The different neutron spectrum modes have different advantages and disadvantages. The thermal cross-section for transmuting MA and fission products is larger than the fast neutron cross-section enabling core inventories to be reduced substantially, but the thermal neutron cross-section of the transmuted products is also large; so neutron capture is a problem. From the point of view of neutron economy, the fast reactor is better than the thermal reactor.

The Th-U fuel cycle is an attractive option for future ADS because it produces a relatively small amount of higher actinides compared with the U-Pu cycle. The Th-U cycle is safer from a weapons proliferation standpoint because of the existence of the hard gamma emitter in the U decay chain and because U can be diluted by depleted or

natural uranium in the start-up or feed fuel. Against these advantages, the Th-U fuel cycle has a less favourable neutron balance.

The MRX is a small PWR suitable for low-temperature process and cogeneration applications, developed by JAERI. It is based on a similar approach to CAREM-25 except that forced coolant circulation and all steam generators are required during full power operation.

Decay heat is dissipated by a passive heat removal system. There is a comprehensive containment system with a large volume of water available in the containment.

The remaining two systems KLT-40C and SMART in Table 14.6 have already been described in the previous two sections on district heating and desalination applications.

The temperatures available in MRX, KLT-40C and SMART for process heat applications are similar to that of current PWRs.

It is also worth mentioning that there are some additional low-temperature process heat applications for which nuclear heating could also be considered, e. g. urea synthesis and wood pulp processing (Institute of Nuclear Engineers, 2004).

Future developments in process heat applications are focussing on higher temperature reactor systems. These are considered in Section 14.8.

The EC research programme has included research on the understanding of radiation mechanisms to provide greater understanding of the physical, chemical, molecular and cellular biological processes as a consequence of radiation. It has also included epidemiological studies of people exposed to radiation. Collectively the mechanistic and epidemiological studies provide a good basis for quantifying the risks from radiation at low doses.

Research tasks of the former kind included the modelling of radiation ontogenesis and related biological effects and the repair of and recovery from DNA damage. Also carried out have been radiation sensitivity and molecular studies of radiation ontogenesis and predisposition to cancer and the effects of in utero radiation. The epidemiology of populations exposed to radiation has covered further analysis of populations exposed to large doses, e. g. atom bomb survivors through to less extreme exposures, e. g. uranium miners and radiation workers, together with the follow-up of cancers incidence. These have been complemented by further research of the treatment of exposed individuals. Other studies have addressed hereditary and genetic factors and the epidemiology of medically treated patients.

In order to evaluate radiation risks, it is necessary to have available high-quality methods for the assessment of levels of exposure to external and internal radiation. European studies have addressed the parameters that determine the fluxes of radionuclides in various ecosystems, in particular the fluxes of radionuclides in surface and groundwaters and the consequences of accidental contamination of environments. Also studied have been the intake of radionuclides and their dosimetry and the monitoring of external radiation. Code developments have been directed towards quantifying the predictions of probabilistic accident sequence codes and the develop­ment of decision supporting systems for emergency site management. Finally risk perception and communication has been studied, including comparative risk assessments of different systems.

The reduction of exposures in accordance with the ALARA principle is the primary goal of radiological protection. The main focus of research is to optimise radiological protection in many complex situations giving rise to radiation exposures, from nuclear reactor operations to participation in various other activities. Research has considered management strategies and techniques for the restoration of contaminated sites and optimisation of radiation protection of patients undergoing diagnostic radiology.

Research has also been conducted in understanding events from the past. The aims of this work are to improve the management of land (territories) that have been contaminated with radioactive material and to contribute to the future health and well being of populations that have been exposed. European research has considered in particular the consequences of the Chernobyl accident and other radiation incidents. An objective of this work is to develop more effective means for managing the radiological consequences of any future accident.

The proliferation of plutonium and the threat from terrorism in modern society is a major driver towards a closed fuel cycle. Another driver is to develop a process for effective management of spent fuel and waste. Advanced reactor concepts provide a solution to these requirements.

For many years fast reactors have offered the attraction of a sustainable fuel supply based on a uranium-plutonium fuel cycle. Uranium resources will last for at least 60 years; so from this perspective there is no immediate need for fast breeder reactors, which (in addition) are about 50 times more efficient than current thermal reactors. There is now a current interest in exploring particular advantages of the fast reactor to consume plutonium, and reduce the stockpile of weapons fuel. Also the fast reactor can be used to irradiate minor actinides (MA) and fission products to reduce the toxicity of long-term wastes.

There are a number of international programmes at the present time that aim to develop the above technology. There are EC initiatives in this area; e. g. a review of gas cooled fast reactor concepts (Mitchell et al., 2001) was carried out within the Fifth Framework programme. The review partners concluded that the gas-cooled fast reactor (GCFR) has a number of potential advantages to offer.

The EC CAPRA (Consummation Accrue de Plutonium dans les reacteurs Rapides) project originally focused on technologies to consume existing plutonium stocks arising from the operation of commercial reactors (IAEA-TECDOC-1083, 1999). Work is currently underway in the EC CAPRA/CADRA project to evaluate the potential for the transmutation of plutonium and MA from waste. A wide variety of reactor concepts of metal cooled fast reactors (Smith et al., 2003; Hesketh, 2003; Vasile et al., 2001) are being considered. The aim is to transmute these actinide species to species with much shorter half-lives.

There are also various international activities on the application of proton particle accelerators in connection with subcritical reactor systems as a means of separating and eliminating actinides via transmutation.

Reactor systems for plutonium burning and the partitioning and transmutation of nuclear waste are among those selected for development within the Generation IV initiative.