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.

Accelerator technology has been developed over several decades and there is some confidence developed in the technology. There are several approaches. The attributes of the different systems are summarised in Table 13.3.

Linear accelerators or Linacs are thought to be achievable up to relatively high power (200 mA, 1.6 GeV). They have been demonstrated as reliable and efficient research tools, and can be made available at a reasonable cost. The most efficient operating conditions for a linear accelerator at the present time would be around 100 mA.

Cyclotron, i. e. circular proton accelerators’ technology has also advanced enabling a 10-15 mA proton beam to be achievable via a segmented cyclotron or synchrotron concept. The most efficient operating current for these is around 10 mA. They have some benefits compared with a Linac but also some disadvantages. The cyclotron

Table 13.3. Accelerator driven systems

occupies a smaller physical area and is cheaper than the Linac, but the space limitation limits the proton current, in the present day to about 10-20 mA. Linacs do not suffer this limitation.

On a larger commercial scale, one option might be to use one linear accelerator to a number of sub-critical reactors by splitting the beam. However, there may be drawbacks in the event of failure of the beam dividers, in which case the full beam might be directed against one target, or failure of the full beam would shut down all the sub-critical reactors.

This problem could be overcome by using one or more smaller cyclotrons, running several smaller reactors, but at increased cost. Regarding the status of cyclotron technology, cyclotrons of 1.1 MW beam power for a 600 MeV proton accelerator have been developed at the Paul Scherrer Institute (PSI). A number of alternative options are under consideration, e. g. a ‘multi-stage-parallel’ cyclotron arrangement in which several lower energy, low current cyclotrons input into a high-energy cyclotron. This approach would also give some cost benefits in terms of energy scaling, compared with a linear accelerator.

A small capacity transportable nuclear power and technology plant (SC TNPTP) is being considered for electricity and heat supply, production of freshwater and also hydrogen (Komkova et al., 1998). The concept has been put forward by IPPE and St Petersburg Marine Building Bureau. The plant rating is chosen in order to optimise the economics for application of the reactor in remote areas in Russia. A 1-MWe unit prototype reactor TES-M has been designed but it is necessary to increase the power in SC TNPTP by at least a factor of 2, with no significant increase in the mass and dimensions to achieve satisfactory economics.

Integral passive containment cooling tests were performed for AP600 to examine the overall containment performance at large scale. At the time, there were no other water distribution tests to provide a demonstration of water distribution over the steel containment dome outer surface and the top of the containment side walls. Wind tunnel tests were conducted to confirm the structural performance of the containment shield building air inlet and outlet.

A large-scale integral system behaviour test facility PANDA (Coddington et al., 1993), is present at the PSI in Switzerland. This was originally built to understand better, long­term decay heat removal by natural circulation in passive boiling water reactors. However, since the latter is a generic phenomenon, many of the data from many of the tests are of relevance to more general light water reactor applications.

The LINX facility (Coddington et al., 1993), is another facility at PSI that was used to investigate the thermal-hydraulics of natural convection and mixing in pools and large water volumes. In the past, aerosol transport was studied in the AIDA facility. This is a separate-effects facility for the investigation of aerosol transport and the associated deposition in plena and tubes.

A European Thematic network has been established for the Consolidation of the Integral System Test Experimental Databases for Reactor Thermal-Hydraulic Safety

Analysis (CERTA-TN) (FISA 2003, to be published). The objective is to preserve for the future, the reactor safety thermal-hydraulic databases acquired in various integral system test facilities. A database will be produced that has up-to-date data access and retrieval capabilities and uses modern web-based information technologies.

In the final part of this chapter some of the research requirements for future innovative reactors are addressed. Some of these also relate to work that will be needed to realise nearer term evolutionary and prototype reactor systems that will also be required to confirm the technologies of the longer term Generation IV reactors before they are built.

Reprocessing of nuclear fuel has been carried out for a number of years in various reprocessing facilities in many of the main nuclear power-producing countries. These facilities include the reprocessing plants at La Hague in France, the Tokai plant in Japan and Sellafield in the UK. A pilot study plant operated near Karlsruhe in Germany,

Pop. Pop.

Pop. = Population

a multi-national plant was built at Mol in Belgium and operated from 1966 to 1974 and the West Valley facility was operated for a short time in the US. Fuel reprocessing activities have also taken place in Russia, e. g. the Chelyabinsk plant, see for example Rougeau (1997).

Over the years many thousand tons of spent fuel have been processed. The plants have been adapted to take account of different types of fuel from different types of reactors. These included early gas-cooled reactors in Europe, Magnox and AGRs in the UK, most particularly light water reactors in operation in many countries and some fast reactors.

For example from 1966 to 1987, the UP2 facility at La Hague processed gas-cooled reactor fuels. Fast reactor fuel was processed between 1979 and 1984. Since 1987, UP2 has been utilised in reprocessing LWR fuels only. MOX fuel has been reprocessed since 1992. A newer facility UP3 started in 1990, having the capacity for servicing a range of spent fuels from European and Japanese facilities.

Concerning management of waste, there are currently three industrial-scale vitrification facilities in operation in Europe. COGEMA operates the R7 and T7 facilities at La Hague and BNFL operates a plant of similar design in Sellafield. Much progress has been made in reducing the volume of high — and intermediate-level waste and also in reducing the radiological dose rate to workers, see, e. g. Table 2.14.

Despite these successful operations, public safety concerns do exist and are focussed on perhaps two main issues. These include the level of release of toxic and radioactive releases from the plants and also concerns about proliferation.

Regarding releases, there are concerns that small traces of radioactive material are released even during normal reprocessing operations. The public is not convinced that neither possible routes back to the public and the food chain are properly understood nor that the limits set by the radiological protection bodies are proven to be safe. The concerns in regard to proliferation are that reprocessing enables the recovery of plutonium, which could be utilised for nuclear weapons. It is worth noting, however, that plutonium recovered from LWRs (the most widely used reactor in operation today) is not in the most suitable form for weapons production.

One of the main drivers for reprocessing plutonium is to support a fast reactor fuel cycle, but only a few fast reactors remain in operation at the present time. However, reprocessed plutonium can be used in fresh MOX fuel. Plutonium only remains in a separated state for a relatively short time during the fabrication of MOX fuel, once in the reactor the fissile plutonium content is substantially lessened and a relatively high percentage (30%) of the plutonium content is burned.

Reprocessing activities are carefully monitored by the IAEA and other national bodies to ensure that proliferation issues are properly covered. It should also be stated that without reprocessing, the quantities of plutonium produced in-reactor will remain the same for many years (hundreds of years) after which time separation becomes easier following the decay of shorter lived isotopes.

Fuel cycle strategies should satisfy a range of criteria for optimising performance (Ion and Bonser, 1997). Clearly they should aim to utilise the available fissile material in full. This might be achieved by recycling of uranium and plutonium from already irradiated fuel or through other sources. The total fuel cycle costs must also be optimised in order to maximise the utilities’ performance from an economic perspective. There is increasing tightening of regulations from national governments and international bodies such as the EC on environmental releases. The impacts of fuel cycle operations on the environment should be minimised by optimised waste and spent fuel management planning. Since fuel cycle activities inevitably involve the handling of fissile material through reprocessing or other means; the political and proliferation issues must also be adequately managed.

The notion of a holistic fuel cycle has been put forward by BNFL; the main elements are summarised in Table 5.1. The holistic approach recognises that different systems and fuel cycle policies can exist in various countries but it is sufficiently flexible to accommodate

such differences. The approach has been adopted for current fuel cycles, e. g. for the AGR fuel cycle in the UK, for MOX fuel cycles in LWRs (light water reactors) and for LWR fuel in CANDUs. The approach is also being adopted for advanced fuel designs.

The Japanese standardisation programme was a collaborative effort between the

government and industry, led by the Ministry of International Trade and Industry

(MITI) (IAEA-TECDOC-968, 1997). It started as early as the mid-1970s with the

objective of standardising LWR designs on the operating plants of the day. A later phase

starting in 1981, aimed to establish a Japanese capability for LWR design based on in­house technology.

The advanced pressurised water reactor (APWR) and advanced boiling water reactor (ABWR) have been developed against these utility requirements policy. Future LWRs based on evolutionary developments of these designs are being investigated by MITI and other industry groups. Mitsubishi and Westinghouse initiated in 1994, a successor programme to the APWR. Japanese BWR utilities together with Hitachi, Toshiba and General Electric (GE) initiated the ABWR evolutionary programme in 1990.

9.1. INTRODUCTION/OBJECTIVES

This chapter discusses the status of nuclear programmes that are proceeding in the various countries that currently operate nuclear plant. It covers the European countries, North America (US and Canada), the countries of Asia (Japan, Korea, China, and India), the Russian Federation and other areas (e. g. South Africa and Latin America). The majority of countries with nuclear programmes are focussing on water-cooled systems to provide their requirement. Reference is also made to progress with other reactor systems in the countries where they occur. The emphasis though in this chapter is on current and near-term activities; longer term initiatives are reviewed later in the book.

At present, there are significant differences among the countries with current nuclear programmes in regard to their position on nuclear power for the future. New build is continuing in Asia, including some evolutionary plants. One European country, Finland seems likely to place an order for a large water reactor in the near future. In contrast, some countries with large current programmes have moratoria on the building of new plant; others remain uncommitted or neutral. There are also some countries in Central and Eastern Europe that have plants at different (in some cases advanced) stages of completion. Progress has been halted in some cases due to economic or other reasons.

Advanced plants have a number of design features, active systems and attributes relying on natural processes to reduce the source term in the event of vessel failure.

Clearly corium emanating from the vessel should be appropriately quenched (in such a way to avoid a steam explosion). Ways in which this can be achieved have been discussed above.

MCCIs result in the emission of a large quantity of aerosols that carry fission products into the containment atmosphere. Possible measures to reduce MCCIs, e. g. using ‘core catchers’ have also been discussed above.

Large surface areas are useful for the plate out of aerosols. There are many natural processes, agglomeration, sedimentation, diffusiophoresis, thermophoresis and hygro — scopicity that promote deposition onto surfaces.

Internal containment sprays provide a means of entraining or dissolving air-borne fission products in water which can then be retained in the containment sump. There are chemicals such as sodium hydroxide, sodium thiosulphate or hydroxine that can be put into the water in the spray systems to enhance the removal of some fission products, especially iodine and caesium.

Fission products can be scrubbed in large pools of water. Similarly, water flooding of debris also provides a potential for scrubbing.

Elemental iodine resuspension can be reduced by the maintenance of a pH > 7 in water pools.

In the SBWR design (Naitoh et al., 1992), steam released to the drywell is channelled through a condenser. It is then condensed and then returned to the gravity-driven cooling system pools. This provides a mechanism for aerosol deposition and fission product removal.

The source term can be mitigated in some designs by introducing ventilation systems for cleaning exhaust air. The SPWR, which is a variant of the AP600, developed by Westinghouse and Mitsubishi, includes in its design an emergency passive air filtration system to mitigate releases into the lower containment penetration area. The air is filtered before being mixed with the cooling air of a PCCS system (similar to that in AP1000/600).

Ventilation systems may also be useful for designs with a secondary confinement if it became contaminated as a result of leakage from the primary containment. In some cases, primary containments are surrounded by additional containment buildings maintained at a slightly sub-atmospheric pressure. This is to ensure that residual fission products released from the primary containment do not escape. Controlled release from filters or stacks may then be considered.

13.9.3.1 Sodium-Cooled Fast ADS. ADS have more ‘inertia’ than the corresponding critical reactor in that they are less sensitive to both positive and negative feedbacks (Bell, 1994). For example, with the source still on, they have lower but wider power peaks than in the critical reactor. In the ADS, the power rises earlier due to the lesser influence of negative feedbacks such as Doppler, axial expansion and structural effects. It falls later due to the lesser influence of fuel dispersion. In the sodium voiding phase, pin failures could occur.

The more sub-critical the ADS, the more the above features are seen. To avoid core meltdown, the source must be switched off, before much sodium voiding occurs. As stated earlier, fuel slumping can lead to re-criticality and power excursion because in a fast system the core is not in its most critical configuration prior to the event (Theofanous and Bell, 1985).

The ADS has some advantages over the critical reactor in that the time constants for the power excursion are longer and rapid power excursions are not possible at least when the ADS is in its original configuration. There is similarly a longer time period to detect sodium boiling or pin failures and hence initiate a beam switch-off.

13.9.3.2 Gas-Cooled Fast ADS. Gas-cooled fast ADS share the similar advantages and disadvantages that gas-cooled fast critical reactors have compared with fast sodium systems. Advantages include the utilisation of a chemically inert gas, and it may be possible to use water for post-accident cooling, e. g. if an in — or ex-vessel core-catcher can be designed and concerns of re-criticalities can be addressed.

Gas-cooled systems can also clearly suffer cooling failure events, but since system pressures are comparatively much higher than the sodium coolant system, LOCA accidents are an additional issue. In all cases, shut-off of the beam is crucial for preventing a core melt. A disadvantage of the gas system is that decay heat removal cannot be achieved solely by natural convection, thus back-up diesel generators are needed to be on stand-by in the event of loss of power to the active circulation pumps.

13.9.3.3 Lead-Cooled Fast ADS. Lead coolant has a number of advantages as a coolant compared with liquid sodium. A lead system would not suffer from positive feedback effects on reactivity in the event of boiling. It is only a weak moderator and, therefore changes in reactivity do not result due to changes in density effects.

It is also relatively chemically inert to air and water. The one disadvantage is the relatively high melting point 327°C, which means electrical heating would be required during start-up and there may be the possibility of freezing and blockage in the event of electrical system failure.

The Rubbia design includes lead as a coolant and it has been analysed against cooling failure transients such as LOF due to pump failure and LOHS due to loss of feedwater. The system has good natural circulation cooling characteristics so LOF is not an issue. For LOHS, meltdown could occur if the beam is not shut-off. This could also occur for slow reactivity insertions under similar conditions. However, the Rubbia system incorporates a specific provision to shut off the beam, based on shielding of the target by a rising liquid lead level under accident conditions. Further provision is also included in the Rubbia design to ensure long-term removal of decay heat by air natural circulation of the guard vessel.

13.9.3.4 Thermal ADS with a Circulating Salt-Fuel Mixture. Thermal systems have generally larger cores than fast systems because power densities are lower. This has the advantage of greater thermal inertia under coolant failure accident conditions allowing more time for accident detection, prevention or mitigation. Without switch-off of the beam though, pressurisation, heat-up and boiling would occur. However, this could be mitigated by spallation target melting and material movement leading to neutronic shut-down.

The time-scale for decay heat-up of the larger core systems is of the order of tens of hours since the fuel is distributed around the core and primary circuit and the system is in natural circulation mode. Some long-term cooling system/procedure though would need to be established, i. e. the salt-fuel mixture may need to be drained into a cooled tank. For smaller systems it may be possible to remove all the heat via natural circulation.

An advantage of a liquid fuel system is that short-lived fission products can be removed to reduce the fission product inventory.

The precipitation of fuel or MA may be a concern in salt-fuel ADS. This phenomenon could lead to flow impedance and loss of cooling in selected areas but also the density variation around the circuit could lead to criticality concerns.

There is also some concern that loss of cooling could lead to power increases due to a positive temperature coefficient in pure salt/Pu/minor actinide mixtures, since no 238U or 232Th with their absorption resonances would be present.

There is also the issue of possible explosive contact between molten salt and water; there may be potential for this event in some designs (Hohmann et al., 1982).

Inspection of components is also difficult in molten salt-fuel systems because pumps and heat exchangers become contaminated with radioactive material. Furthermore leaks would result in contamination of the whole containment.