15.13.1 Critical Reactors

Many of the innovative systems under consideration have the capabilities to use advanced fuel cycles and there is a need for fuel performance research for many of these fuels (The US Generation IV Implementation Strategy, 2003). There are, however, a number of activities currently in progress (Table 15.8).

Innovative water reactor fuel cycle options are being considered whereby spent PWR fuel can be used for CANDUs, i. e. the DUPIC technology. This is attractive to avoid the separation of fissile material, particularly plutonium, during fuel cycle operations.

High-temperature reactor fuel design is also attracting research, e. g. on the stability of particulate fuel at very high temperatures and particularly under accident conditions. There is also relatively little experience on fuel fabrication. Regarding current research programmes, there is an OECD co-ordinated research programme on the physics of plutonium/innovative fuel cycles for pebble bed reactors. The IAEA is also co-ordinating analysis of experimental results for a number of high-temperature test reactors including, HTTR (Japan), HTR (China), GT-MHR (US and Russia) and ASTRA (Russia) (Newton, 2002).

Regarding innovative fuels, nitride fuels have been considered instead of oxide fuels, because they result in lower fuel temperatures, due to their improved thermal conductivity. A wide range of different fuels has been considered for minor actinide target fuels. There are several EC research programmes on advanced fuel-related issues. The CONFIRM programme is an experimental investigation of the high-temperature stability of actinide fuel in nitride form.

There are a number of initiatives in connection with fast reactor core physics, e. g. the CAPRA/CADRA programme, which includes studies of fast reactors, particularly in connection with plutonium burning and waste incineration fuel cycles. Uncertainties in fast reactor performance are the subject of a present IAEA research programme.

Thorium fuels have been considered as an alternative to uranium for some fuel cycles and some reactor types, but there is limited experience, certainly for commercial reactors. There is limited experience in general for the molten lead cooled and molten salt cooled systems.

15.13.1.1 Accelerator Driven Systems. In some respects, there are similarities between the important issues concerning subcritical ADSs and the critical reactors. However, the application of accelerators to different subcritical systems does require some new areas of research. There is also the issue of research on accelerator systems per se.

There has been research into the development of fuels for ADS. The EC FUTURE programme is investigating a number of innovative oxide compounds, in solid solution and inert matrix form. The ADS is also considered in the CAPRA/CADRA programme.

There is emphasis in designing new innovative reactors (critical and subcritical) in a way to reduce the radioactive waste burden compared with existing reactors. This is being achieved through design for high fuel burn-up; the utilisation of thorium could also achieve reduction in the higher actinides produced.

A feature of many of the new designs is that new coolants are being proposed for which little operational experience exists. These coolants are considered in Section 15.14.

15.9. ADVANCED COOLANTS

15.14.1 Present Experience

The future supercritical water concepts will operate with supercritical water on either the secondary side or also possibly on the primary side. It follows that the primary to secondary heat exchangers will require special attention. The performance of conventional PWR steam generators has not been without problems and the materials used, system chemistry control and the construction methods for the supercritical systems will need significant development research. For the supercritical heavy water concepts such as CANDU X, lower cost techniques will have to be developed for separating deuterium from hydrogen in light water.

There is some relevant experience of supercritical systems from coal-fired plants. However, there is no previous experience on the use of supercritical water in high radiation backgrounds. There is also the issue of system performance under fault conditions, e. g. LOCAs, given the very high system pressures.

For the liquid lead coolant systems, there is little experience in nuclear reactors outside of Russia. There will need to be significant effort to developing the chemistry specifications and control to ensure economic and reliable performance. Lead-induced stress corrosion cracking could also be an issue.

Molten salt systems will require developments on the control of their chemistry and the coolant composition during their extended periods of operation. The high-temperature performance of key components such as heat exchangers will need to be verified. There needs to be developed isotope separation technologies to separate out the lithium isotope 7Li from the naturally occurring 6Li.

15.14.1.1 Natural Circulation. Existing operating water reactors rely on natural circulation to remove decay heat when forced convection is lost. Many water and heavy water cooled designs include natural convection to remove decay heat after shutdown. Some of the simpler low-pressure water reactors rely on natural circulation to remove heat at all power levels. In general, there has not been a total reliance on natural circulation in the pressurised PWR and BWR systems. The innovative liquid lead and molten salt systems also allow for some level of natural circulation, namely the removal of decay heat after shutdown. Some of the ADS systems allow for natural circulation removal of heat at all power levels.

It follows that natural circulation will be an important phenomenon in innovative reactor technology. Regarding the current state of knowledge, there has already been much work on natural circulation in current and evolutionary plant. There is greater confidence in single-phase system performance, e. g. in the gas and liquid lead systems, than in the water systems, where two-phase flow can develop. Under accident conditions the presence of hydrogen can also be a problem. Additionally, there is a need for the development of correlations for transient heat transfer under all operating conditions.

There are several projects within the EC framework research programme for evolutionary systems; these were mentioned earlier, e. g. EUROFASTNET, ECORA and FLOMIX-R. Most of these are relevant to improving the understanding of natural circulation in the evolutionary and some of the innovative reactor systems. The above programmes also cover theoretical R&D, e. g. the development of appropriate numerical methods development for CFD modelling.

A review of the place of nuclear power in world energy generation compared with other energy sources has been carried out by Birol (2000). The work is in the context of the International Energy Agency’s World Energy Outlook (World Energy Outlook, 1998). This paper projects that nuclear energy generation worldwide will be broadly at the same level in 2020 as at present (Figure 2.2) and summarises differences in national policies. It is clear that there is marked difference of prospect across the various world sectors.

Nuclear electricity production is increasing in China, and in other developing countries and particularly in Asia (Figure 2.3). The most notable examples are Japan, Korea and Taiwan. Other countries planning expansion include India and Pakistan (Fisk, 1999). The main reason for the increased production is the building of new plants and indeed the share of nuclear electricity in these countries is increasing. Other Asian countries are also considering building. These include Indonesia, Thailand, the Philippines and Vietnam.

In North America, the situation is less certain. There could be a significant decline in nuclear generation since a number of the US plants are older reactors. However, there are increasing drives to extend the life of older plants. In recent years, there have been generally positive statements on the prospect of building new plants in the US in the future.

The situation is similar in Europe where no new plants have been ordered and relatively few plants are under construction. There is also a marked variation in national policies from country to country.

The reasons for the overall decline in Europe and North America also vary from country to country. In most countries, the reasons are partly economic and partly political. In the UK for example, there are no restrictions in principle on the building of new plant (subject to regulatory approval); the issues are primarily economic. A similar position exists in Finland. Elsewhere in the EU, Belgium, Germany, Netherlands, Spain and Sweden

continue to operate nuclear plant, but have a moratorium on the building of new plant. In Italy, all nuclear plants have been shutdown since 1990 and there was an immediate moratorium on building. Other EU countries, e. g. Denmark, Greece, Ireland and Norway have never built nuclear power reactors and none are foreseen in the future.

In Russia, Ukraine and the Central European Countries, there is a positive attitude to nuclear power production. A large number of reactors are in operation and dependence on nuclear power is necessary to provide these countries’ energy requirements in the short to medium term. However, some of the older designed reactors built during the Soviet era are generally recognised as having safety limitations. There are political forces that these should be closed down. Many such plants have already ceased operation. New electricity producing plants are being required to fill gaps in supply and new nuclear power plants are filling that demand.

Nuclear power plants are also in operation in South America (e. g. Argentina and Brazil) and in South Africa. There is currently a global initiative by ESKOM to design a new high — temperature reactor based on an earlier pebble bed design. This reactor type is discussed later in the book. In Australia, there are no plants currently in operation and there are restrictions in place on future building.

Seismic evaluation or re-evaluation is an important issue with many existing nuclear power plants (IAEA/NSR/2002, 2003). This may be required to take account of new information that has come forward since the original evaluation. In some cases better margins can be demonstrated with the availability of new analysis techniques. It may be that older conservative safety margins are not considered sufficient in the light of present day requirements or that original evaluations were inadequate.

There are a number of supporting facilities that also require seismic evaluation. These include laboratories, research reactors and fuel cycle facilities. In general the cases for these facilities are less advanced and they present a wide range of different situations. Seismic evaluation of existing nuclear facilities is the subject of a recent IAEA international meeting in 2003 (IAEA/NSR/2002, 2003; IAEA International Symposium, 2003).

A number of IAEA Member States have on-going seismic upgrading programmes to improve the safety of their operating plant.

The US has recently taken a decision to proceed with a spent fuel and high-level waste repository at Yucca Mountain (IAEA/NSR/2002, 2003; Figure 6.2). This is an important development, which has been met with considerable opposition and challenges to the supporting safety case. The repository will be under the control of the USDOE.

There has also been significant progress in Finland and Sweden. A good review of the important issues is given in Ryhanen (1996) together with a status commentary of the position in Finland, a leading country in developing long-term waste disposal strategies.

Some typical examples of waste disposal principles, identified in Ryhanen (1996) are the following. International recommendations exist for the various stages of waste

management, and these are reflected in national legislation, albeit with some specific national modifications. Waste management facilities are licensed by national regulatory bodies. Relevant research, including experimental and theoretical R & D programmes, is conducted to confirm the technologies. An important issue concerns the financing over future years, e. g. one principle is that these costs should be recovered from ongoing electricity revenues, and set aside to guarantee future funds, but finding an appropriate model is an issue in many countries.

It is widely accepted that improved communication is needed to educate the public on the safety of the proposed technologies. It is not always clear what the public’s concerns actually are. It is not always the safety issues that dominate the argument on long-term disposal. Could the presence of a waste disposal site impact on the commercial success of a region, e. g. by militating against the sale of its produce or through an adverse effect on tourism? Is the idea of final disposal less attractive than long-term temporary storage, the latter implying easier monitoring?

Many of the technical issues are complicated, covering wide ranging topics including organisational frameworks and responsibilities, technical details of the disposal, site selection criteria, licensing proceedings, etc. Therefore, an objective in communicating the technical issues to the general public is to make these as simple and straightforward as possible. In Ryhanen (1996), a number of simple observations are suggested. Is the risk of environmental pollution from spent fuel realistically perceived? For example, much of spent fuel is relatively insoluble, its radioactivity decreases with time, it does not radiate to the surface from an underground repository, etc.

Different groups of people need to be targeted in different ways. The most important groups to be targeted are the political decision makers both at national and local levels and the public. In Finland, the Posura Company’s information programme ranges from press conferences, meetings with the municipalities, open houses to the public, exhibitions and lectures. Presentations are tailored to the particular audience and supported by relevant documented material. (Posura is a company that has been established by the Finnish Utilities TVO and IVO to specifically address the issue of the final disposal of spent nuclear fuel.)

It is clear that much understanding and progress has been made towards meeting the concerns of the problem of waste management and the management of spent fuel. However, it is also clear that much work is still required; this will need time and patience if the goal of finding an acceptable long-term solution is to be achieved.

Within Europe, the EC is likely to set a timetable in the near future for Member States to identify sites and to set up repositories for spent fuel and high-level waste disposal. The objective is to accelerate the various delays in decision making on the waste disposal issue that exist in some countries.

OECD/NEA is also encouraging countries to find long-term sustainable solutions to the waste problem (NEA Annual Report, 2002). It is aiming to facilitate improved technical and societal confidence in geological disposal in repositories. Activities in 2002 have included peer reviews of the Belgian and Swiss proposals, workshops on stakeholder involvement and technical reviews on the status of engineered barrier systems and ways in which geological science can be used to support repository safety cases.

The IAEA has recently reviewed the major issues and trends in radioactive waste management (IAEA/NSR/2002, 2003). This review reinforces the importance of the social and political aspects of radioactive waste policy. Issues relate to control of discharges and the availability of a retrieval option from repositories, particularly for spent fuel and high — level waste. A long-established principle is that waste should not impose an ‘undue burden on future generations’. This is now being broadened towards the idea of an ‘obligation’; that the current generation should avoid taking irreversible actions that may mean that certain necessary or desirable future options are not available to future generations.

Additional to its current nuclear generation commitments, the US has in place programmes for the design and potential licensing of advanced light water reactors (ALWRs). Various ALWR designs have been certified in readiness for construction in the event of a new build programme.

The important codes, guidelines and URs are set down in a prescriptive set of documents that encompass existing reactor regulations, 10 Code of Federal Regulations (CFR) 50 (USNRC, 10 CFR Part 50, 1988). A new regulation, entitled ‘Early Site Permits; Standard Design Certifications; and Combined Licences for Nuclear Power Reactors’ was published in 1989 (USNRC, 10 CFR Part 52, 1989). This has been used in the process of issuing design certifications for the ALWR designs.

In addition to these codes, various USNRC policy statements have been issued on standardisation, regulation of advanced NPPs, goals for safety and severe accidents. The USNRC have also issued safety evaluation review reports, NUREG-1242 (USNRC Review of Electric Power Research Institutes, 1992-1994) on the EPRI ALWR URDs (EPRI NP-6780, 1990), discussed in Chapter 7.

The USNRC have established a number of important principles regarding the safety of future reactors. In general, future reactors should achieve a higher degree of safety than is deemed acceptable for currently operating plant. Severe accidents need to be considered in the design process. The US requires a complete Probability Safety Assessment of the design (as do the URs Documents that are more conservative than the USNRC goals for current generation plants by at least a factor of 10). Further accident management measures need to be identified during the design process and should be an important mitigation in severe accidents.

The Czech Republic has six nuclear power plant units in operation, four at Dukovany (VVER-440/213) and two at Temelin (VVER-1000) (Foratom e-Bulletin, 2003b). In 2002, the nuclear share of electricity generation was 25%, (International Atomic Energy Agency, 2002). Of these, the Temelin units are the newest plant — these underwent commissioning and active tests in 2002.

6.3.5 Hungary

In Hungary, there are four operating VVER-440/213 units operating at Paks. In 2002, they produced 36% of the country’s electricity requirement (Foratom e-Bulletin, 2003c). Upgrading projects have been implemented for the Paks plants following recommen­dations from the G7 countries to improve the safety of all Soviet-designed reactors in Central and Eastern Europe.

6.3.6 Lithuania

A very large share of the country’s energy requirement comes from operation of the Ignalina RBMK units 1 & 2. In 2002, these plants generated 80% of domestic electricity

(Foratom e-Bulletin, 2003c). Despite improvements, these plants will be shutdown before the end of their 30-year design lifetime (World Nuclear Association, 2003). Ignalina 1 is due to close by 2005 and the closure date for Ignalina 2 will be set in 2004.

The Lithuanian government has approved the design and eventual construction of an interim spent storage facility to be built at the Ignalina site.

Four different design concepts were also investigated in the early 1970s by the European GBR Association, taking the LMFBR fuel and core technology with a gas thermal reactor system. These designs were not developed commercially at the time, due to their cost, and the preference for LMFR and LWR systems, which were further developed. They are now being reconsidered in modular designs with more favourable economics and with more natural safety characteristics.

Three 1000 MWe systems were considered, GBR1 (helium (He) and fuel pins), GBR2 (He and coated particles), GBR3 (CO2 and coated particles) and one 1200 MWe system GBR4 (He and fuel pins). The latter was the favoured option at the time.

GBR4 had vented pins containing coated particle fuel. The coolant pressure was 90 bars, necessary to achieve the efficiency using fuel pins. Two independent shutdown systems were employed. There is an advantageous safety feature associated with a negative reactivity expansion coefficient. C&I systems were based on 1970s technology and therefore a new design concept today would require a more up-to-date approach.

The reactor pressure vessel enclosed an integrated system, incorporating the boilers in individual pods in the vessel in a design similar to that employed by the AGRs. An independent decay heat removal system was also included. The containment included an inner steel liner and outer concrete shell.

Additional safety features were incorporated to accommodate cooling in a depressurisation accident, un-tripped loss of flow, etc. At the time, it was concluded that further development was required in the plant safety concept, particularly in the field of severe accidents, core melt and containment.

The temperature requirements for different heat applications vary considerably. Temperatures of the order of 100°C are required for district heating and seawater desalination whereas for some process heat applications and hydrogen production temperatures of the order of 1000°C and above are required. Different reactor types supply different temperature ranges of output, typical ranges are shown below in Table 14.1.

There is a wide range of applications, and different applications have different requirements, particularly temperature requirements (Table 14.2). The lower temperature end with water reactors and the higher end with high-temperature gas reactors have received the most attention to date (IAEA-TECDOC-923, 1997). A standard requirement for most users is reliability and availability. This is particularly so for the process industry, where production depends on energy supply to continue. In industry, energy must usually be available as a base-load commodity. This contrasts the load requirements for district heating where the demand is dependent on climatic conditions. Consequently load factors for energy producers for district heating applications may be much smaller than those required for industrial applications.

For reactors operating in co-generation mode for electricity and heat, there are issues of balance that need to be considered. For large power reactors, the main output may be electricity and these reactors will be optimised for base-load electricity generation.

For small reactors, a higher proportion of their output may be heat; therefore, significant fluctuation of the heat demand could result in fluctuation of electricity output. Thus the technology needs to ensure that the electricity production and the grid load are compatible.

In this chapter, some of the various reactor designs that are being considered for heat and other applications are discussed. To date, most of the operational experience has been on water-reactor systems. Some operational experience has been gained with liquid sodium. Lead and particularly lead-bismuth systems are being examined in Russia, both for district heating and for seawater desalination. As discussed in Chapters 12 and 13, various innovative reactor systems are being considered for high-temperature applications.

In 1993, a network for the evaluation of structural components (NESC) (Rintamaa and Taylor, 2001) was formed, based on a multi-partner collaboration agreement and managed by the Joint Research Centre at Petten. It was composed of utilities, regulators and research organisations and the main objective was to develop and validate structural integrity techniques for assessment. There was a broad representation of countries across Europe, covering countries operating a wide range of nuclear power plants.

There have been four NESC projects.

NESC-1 was the first large-scale project to evaluate the whole process of structural integrity assessment. In particular, the spinning cylinder PTS test was designed to simulate the conditions of an ageing RPV subjected to a severe PTS loading. It demonstrated the beneficial effect of cladding in inhibiting cleavage initiation in the cylinder surface. It was used to validate structural mechanics assessment techniques and to validate no-destructive inspection techniques.

The NESC-2 programme included two large-scale PTS tests on thick wall (200 cm) cylinders with shallow defects. The objective of the tests was to consider brittle crack initiation, the propagation and arrest of shallow cracks in a cladded vessel under PTS loading. The first test, which included a circumferential under-clad notch of depth 8 mm, exhibited a crack growth that was arrested. In the second test, there were two shallow semi-elliptical through clad effects but no growth occurred.

In NESC-3 there is a large-scale test on a dissimilar weld pipe assembly of aged PWR Class 1 piping. It is a benchmark test to demonstrate the load to cause failure at a large defect. The purpose is to quantify the accuracy of assessment procedures for a defect containing dissimilar metal welds, to address issues regarding inspection performance, and to promote best practice.

The NESC-4 test series is to test defect-containing beams, designed to clarify the role of bi-axial stress effects on shallow flaws in RPV weld material.

There are several other EC R&D programmes that have links with NESC. These include exploratory or pilot projects that investigate certain aspects that could lead to further major tests (Tice et al., 1999; Faidy et al., 2000; Leggart et al., 1999) and other collaborations that utilise NESC results (Lidbury et al.).

The European SMILE project (Bezdikian et al., to be published) considers whether the structural margins of aged embrittled RPVs can be improved if a particular potential beneficial effect of load history is taken into account (warm pre-stress). The programme will provide data from representative steels.

In recent observations, different cracks have been discovered in different US and European nuclear power plants (VC SUMMER, RINGHALS, BIBLIS). The issue is the integrity of aged cracked metal welds involving different materials, e. g. ferritic to stainless steel. The extent of crack growth and paths followed by a crack through the weld will be followed under various loads in the ADIMEW project (Faidy et al., to be published).

16.6.1 Integral Codes

Integral computer codes are being developed to provide a LWR accident analysis capability for modelling the course of a severe accident through its various stages.

They provide a phenomena coupling capability from degradation of the fuel rods through to formation of a molten pool and if the accident progresses unchecked, to the containment loading and release to the environment. They include modelling for the release of fission products and aerosols (e. g. from control rod materials and core-concrete interactions). They include models for fission product transport through the reactor coolant circuit to the containment, including deposition, re-suspension of aerosols and also the fission product source to the environment, should the containment fail or be vented by operator action.

Examples of such codes include ASTEC (Jacq and Allelein, 2000; Allelein et al., 2000; Allelein et al., 2001), ECART, MELCOR (NUREG/CR-6119, 1998) and MAAP (IAEA-TECDOC-752, 1994; Table 16.5). These have been validated against various severe fuel damage and fission product release experiments during the course of their development. Further data are now becoming available from the integral PHEBUS FP experiments. The first objective of PHEBUS is specifically to provide high-quality data on the strongly coupled processes that occur in severe accidents, as described above. The second objective is to validate the codes against these data and to define the envelope of validation of the codes. The PHEBUS programme is ongoing currently. In addition to integral analysis interpretation, it is supported by additional analysis from detailed CFD codes.

The Accident Source Term Evaluation Code (ASTEC) aims to model all stages of a severe accident sequence from the initiating event through to fission product release from the containment. It is a European code developed by GRS (Germany) and IRSN (France). The code adopts a best estimate approach and aims to include all the major phenomena and their interactions and also the main plant systems. The other requirements are that it should be fast running, flexible for performing sensitivity analyses and with appropriate validation. The code has been made available to the EC European Validation of the Integral Code ASTEC (EVITA) 5th Framework Project for further validation activities. The applications of the code are for determination of source terms, support to level 2 PSA and to promote better understanding of the physical phenomena.