The Canadian designed Canadian deuterium uranium (CANDU) reactors used natural uranium as a fuel, by employing heavy water as both moderator and coolant. The first CANDU, NDP2-Rolphton of just 23 MW, entered commercial operation in 1962 and a number of 2, 3 and 4 unit plants evolved of commercial power capacity, individual units delivering power in the range 500-800 MW. A schematic of the Darlington PHWR is shown in Figure 1.5.

CANDU reactors consist of horizontal pressure tubes constructed with Zircaloy alloy. They pass through a large vessel (Calandria) filled with heavy water (deuterium oxide) at low pressure and temperature. Uranium oxide pellets are sealed in Zircaloy alloy cans, which are assembled in bundles or fuel assemblies. In a 500 MW plant, each bundle has

Steam

Steam

generator

neavy-water

(in calandria)

about 28 elements. There are about 4860 bundles in total with 12 or 13 such bundles in each pressure tube.

The heat generated is removed by heavy water at about 9 MPa, a sufficiently high pressure to prevent boiling. The water circulates around the fuel elements and passes to a steam generator, a similar principle to the PWR and BWR concepts.

The CANDU reactor is controlled by cadmium absorber rods. When fully inserted, these also provide the shutdown margin. In addition, the reactor can be shut down by voiding the cold heavy water from the reactor core. Vertical steel ‘adjusting’ rods are used to smooth out the sometimes uneven power distribution due to the use of different burn-up fuel segments, even within one fuel channel.

As for the PWR, the CANDU primary system is located in a concrete containment building, of sufficient strength to accommodate a large coolant system break within its design basis. Modern CANDUs are connected by valves to a large vacuum building. In the event of an accident these enable steam to pass from the affected containment to the vacuum building.

The CANDU reactor has a low volumetric power density, about 10 times lower than a PWR, despite fuel ratings that are comparable with PWR. It also has the lowest fuel costs because of utilisation of natural uranium. Against these cost benefits, the CANDU reactor needs considerable quantities of heavy water.

The CANDU reactor does have a number of advantages. It has on-load refuelling and hence has very high load factors. The plant has high availability and high reliability also. Since the design incorporates individual tubes, there is no requirement for a large pressure vessel. From a safety perspective, the reactivity excess is smaller than in reactors employing enriched fuel and hence power excursion transients are less likely.

In terms of disadvantages, the CANDU has a very large core (compared with a PWR or BWR) to achieve a similar power output.

The primary objective of present day fuel cycles is to optimise energy production, while at the same time minimising costs and maximising safety. This has given rise to a number of fuel cycle variants to be considered in which fuel is recycled between different reactor systems, see for example, Ion and Bonser (1997). This is discussed more fully in Chapter 5. There are particular operational safety issues relating to the management of spent fuel, etc., as mentioned in the previous section. These are concerned with safe storage practices, maintenance of sub-criticality in fuel storage ponds and flask transport safety and security.

Spent fuel represents the most highly radioactive waste arising from the fuel cycle, due to the presence of particular fission products. It, therefore, represents the significant waste radiological hazard associated with nuclear plant operations. It may also include fissile and fertile uranium or plutonium and possible breeding products, which might be re-utilised by recycling or reprocessing. The fission product radiological hazard will remain, however, whether or not these materials are recovered.

Policy issues are likely to play a continuing role in the development of advanced fuel cycles. Reprocessing and recycling, together with improved uranium resource manage­ment can lead to a reduction in waste volumes and toxicity, thus improving the sustainability of nuclear power plant operation but this may not be a preferred option due to economic reasons or proliferation concerns. Recycling of plutonium in LWR MOX cores reduces the spent fuel radiotoxicity by a factor of 3 if the MOX fuel itself is not recycled (Bertel and Wilmer, 2003). Multiple reprocessing and recycling can reduce radiotoxicity further.

Reprocessing practices also reduce the volumes of radioactive waste significantly. Each tonne of spent fuel contains about 1.5 m3 of high level waste. After reprocessing, less than 0.5 m3 of waste remains, including 0.115 m3 of vitrified high level waste and 0.35 m3 of intermediate waste. Additionally further compacting can be carried out after disposal.

The subject of fuel performance is of international importance in the nuclear industry. It has attracted significant attention in all the major countries operating nuclear plant, particularly in France, Germany, Japan, US and the UK.

Good fuel performance is a necessity for utilities and the major advances are driven by the utilities requirements. For example for LWRs, as discussed in the previous section, the use of MOX fuel and the potential to use RU from reprocessing is a particular current interest.

In the UK, within its gas reactor programme, much experience has been gained from many years of successful Magnox and AGR operation. The fuel performance in AGRs has been very good in matching the performance of most other reactor types. Similarly experience on fast reactor fuel performance has been gained in the UK and France; this has helped to support the fuel design for the EFR.

New designs of plant are being proposed to cover a wide size range of power outputs. For example, large evolutionary plants with outputs of the order of 1500 MWs are being developed utilising proven active engineered systems. Medium to large-scale plants are being considered which take account of more inherently safe features such as passive safety systems. These passive systems are being scaled up for higher power output plants, in which previously the safety functions could only be accomplished by active systems. Small-scale plants are being designed which encompass novel fuel technologies, etc.

All these designs have common objectives, e. g. high availability, user friendly man — machine interface, competitive economics, and compliance with internationally recognised safety targets (Juhn, 1999).

6.3.1 Canada

Canada has 14 operating nuclear power plant units, which in 2002 produced 14% of the country’s electricity, compared to 13% of the previous year (Foratom e-Bulletin, 2003b). There are a total of 22 nuclear power units but 8 of these have been shutdown for several years. In 2003, permission has been given to load fuel into Units 3 & 4 of Bruce A nuclear power plant. The Canadian Nuclear Safety Commission (CNSC) has granted permission for restart subject to certain specific requirements (Foratom e-Bulletin, 2003c).

Looking to the future, Canada is participating in the Generation IV International Forum (GIF) to facilitate the R & D for these reactor systems.

The long-term management of nuclear waste is under study. In 2002, the Nuclear Fuel Waste Management Organisation was set up to investigate various concepts (World Nuclear Association, 2003). The main proposal under consideration is to bury the waste in the rock of the Canadian Shield, at depths of 500-1000 m, below the water table.

The very high temperature reactor (VHTR) has been put forward by the GIF members as part of their Generation IV programme (The US Generation IV Implementation Strategy, 2003; Figure 12.2). This could be used for high efficiency electricity production but is also

seen as a good candidate for hydrogen production by thermochemical water splitting or through high-temperature steam electrolysis. The reference reactor is a 600 MWt helium — cooled reactor with a coolant outlet temperature of 1000°C or more with a design efficiency of 50%. At this efficiency, it could produce 200 metric tonnes of hydrogen per day. This technology requires advances in high-temperature materials, alloys, ceramics and composite materials.

The VHTR is seen as a natural development of the gas turbine modular helium reactor (GT-MHR) and PBMR reactor designs out lined below. These designs are the latest in evolutionary high-temperature reactor technology that are being proposed for short — to medium-term deployment.

CEA are conducting a research programme (Viala and Salvatores, 1994) on the potential of thermal or fast reactors for transmutation of waste in partnership with EdF, FRAMATOME and COGEMA (Salvatores et al., 1997a). Different laboratories within CEA are working within the ISAAC programme on the physics of ADS including accelerator technology, the physics of source driven multiplying systems and spallation physics.

Feasibility work on accelerator structures with coupled cells and on beam dynamics has been carried out. Theoretical studies on high intensity accelerators have been carried out in support of experiments (FODO, on the beam dynamics) and (SATURNE, on the design of a 100 mA proton source) (IAEA-TECDOC-985, 1997a).

The physics of multiple sub-critical systems plays a central role and are being studied in several experimental programmes. The MUSE experiments (Salvatores et al., 1997b) in the MASURCA facility in CADARACHE are providing understanding on the neutron source and the impact of the source spectrum and environment at different levels of sub­criticality. Supporting experiments to determine actinide and fission product cross­sections are being carried out in the Geel LINAC, other experiments were also carried out in Superphenix.

On the spallation physics, thin and thick targets, spallation residual nuclei measurements, differential cross-sections’ measurements and neutron production rates

are being studied in the SATURNE experimental programme. Codes to model cascades include the code system SPARTE, supported by the Monte Carlo code TRIPOLI and the nuclide time evolution code DAEWIN. Future work envisages the coupling with the standard neutronics code ERANOS (Doriath et al., 1994).

System studies have been performed based on various scenarios, in which an ADS is used to develop a relatively clean source of nuclear energy within a fuel cycle, where LLFP are eliminated and radioactive wastes are concentrated in a small number of facilities in a nuclear reactor park.

Basic nuclear and particle physics is performed by the Institute National de Physique, et de Physique des Particles of CNRS, and also in the Direction Des Sciences de la Materie of CEA. The PRACEN research programme was set up in these laboratories to perform radiochemical studies within nuclear storage facilities. A recent joint research programme ‘GEDEON’, involving a collaboration between CNRS, EDF and CEA, has been set up to encompass the common areas of interest of the ISAAC and PRACEN programmes and to explore innovative options for waste management.

Of all the high-temperature applications, there is probably maximum interest in hydrogen at the present time. Hydrogen production is an important long-term objective of the Generation IV Programme.

The challenge is to develop a hydrogen generation process that does not release greenhouse gas such as carbon dioxide (Institute of Nuclear Engineers, 2004). The classical fossil fired steam reformation of methane has this problem and methods of reducing the CO2 release are under development. Other techniques being investigated include high-temperature electrolysis and also thermo-chemical water splitting. Neither of these methods produce CO2.

The generation of hydrogen using nuclear heating is under consideration within the JAERI HTTR programme described below. Hydrogen production technology is also being considered within the US Next Generation Nuclear Plant programme at Idaho.

16.1. INTRODUCTION/OBJECTIVES

In parallel with current experimental research programmes, there are many activities devoted to improved model and associated computer code developments. Prior to plant application, these are validated against experimental data usually enveloping their application. With the increasing cost of experimental research, the improvement of models and the advent of faster and faster computers, the proportion of theoretical work in comparison to experimental work is increasing.

Computer codes are required for both design substantiation and safety analysis. The emphasis in this chapter will be mainly on the methods that have been developed and are available for safety analysis. A wide range of methods and codes have been developed and validated for current generation plant. This chapter examines the present status of research for current generation plant together with the implications for advanced applications. Current research is not only targeted towards more advanced modes of operation of existing plant but also for application to advanced, particularly evolutionary, plant applications. In many instances, code validation for current plant remains valid for evolutionary plant. This chapter does, however, describe how new phenomena relevant to evolutionary plant can be modelled, e. g. associated with passive system performance.

This chapter considers the role of different types of codes, integral, system, lumped parameter, computational fluid dynamics (CFD) and other specialist codes in the context of reactor design and safety research. More stringent safety standards imply more exacting quality assurance standards for all levels of code development, verification, validation and applications. Advanced software techniques offer more automated tools. Modern computer platforms enable detailed safety analyses to be performed that were not feasible at the times of licensing of many of today’s plants. These theoretical topics are covered.

The primary focus of this chapter will be on water reactor technologies. Water reactors occupy the overwhelmingly largest fraction of existing reactors in operation today and the nearest term evolutionary reactors are also likely to be of this type.

Analytical methods have been developed for other reactor types and some of these will be developed further as the innovative gas and liquid metal reactors’ concepts move forward. They will receive a brief mention in the last section of this chapter on innovative reactors. Analytical methods developments for the innovative reactors will proceed in parallel or slightly behind the corresponding experimental programmes that will be needed to develop innovative reactor technology. The latter were described in the previous section.

As far as possible, there is a need that all plants take into account developments in safety standards and technology. It is unlikely that older plants will meet the same standards as modern plants but they must have adequate operating safety margins. These are assessed, for e. g. by following plant modifications, new fuel cycles and also during periodic safety reviews (PSRs), discussed in more detail later in the book. Most utilities are required by their regulators to carry out PSRs at regular intervals, typically at least every 10 years. The purpose of these reviews is to consider all facets of the long-term operation of the plants (rather than the particular every day running of the plant).

Considerable experience has been gained in the UK on the continued operation of nuclear power plants over the past 50 years. Some particular activities that are being carried out in support of the Magnox reactor programme (Mortin, 2000) are described below and are typical of the practices that need to be adopted for older generation plants. These are summarised in Table 2.7.

Plant maintenance and monitoring practices must be reviewed to take advantage of improved techniques. Particular checks must be made on the functional testing of

Table 2.7. Important issues for continued operation

Increased safety demands — impact of new standards and technology on performance and operation Plant maintenance and monitoring — availability of improved techniques Ageing — status of plant and how undesirable effects can be mitigated

Long-term technical support — availability of Suitably Qualified and Experienced (SQEP) personnel

Mortin (2000).

components and on equipment settings. If necessary, components under wear must be replaced. Another purpose of the reviews is to establish that the frequency and scope of inspections is optimal from the point of view of both safety and cost effectiveness.

The issues of structural plant ageing have to be addressed. Plant ageing can result from many wide-ranging and different phenomena depending on the plant in question. Pressure vessels may become embrittled as a consequence of high fluence particularly at welds. Chemistry effects such as oxidation of graphite cores is a particular issue in gas reactors that needs to be considered.

Another issue identified concerns the technical support of plant. There are issues arising from the potential loss of staff, perhaps recruited in the early days of plant operation but who may be nearing retirement several decades later. The problems of recruiting into a nuclear industry that may be scaling down are well recognised.