Liquid effluents, which are produced from different parts of the plant, are routed to the liquid waste treatment plant to be decontaminated and chemically neutralized. Radioactivity arises mainly from impurities in the effluents that are

easy to precipitate and to trap as solid wastes. These solid wastes are then conditioned appropriately for disposal. Liquid effluents are then filtered and monitored prior to authorized release to the sea. Each release is performed after analysis of representative samples under the control of the radiation protection unit. Released volumes and amounts are recorded and sent on a monthly basis to the safety authorities.

As an example, the liquid releases at La Hague in France, measured in 2002 (Source: French Nuclear Energy Society), are given in Table 16.3.

The fundamental principles of transmutation, outlined above, have been applied in the analysis and performance assessment of different types of reactors: thermal neutron reactors, fast reactors and sub-critical, i. e. source-driven, reactors. The analysis has been mostly applied to TRU transmutation and has been described e. g. in Ref. 5. Here we shall summarize the most significant results related to TRU transmutation and to long-lived fission product transmutation. Finally the potential impact of transmutation on the fuel cycle will be discussed.

The transmutation performances of different thermal neutron reactors have been widely investigated in the last decade. Focus has been on PWRs, high temperature reactors (HTRs) and on the use of inert matrix fuels (IMF) in PWRs. (IMF are fuels without uranium, which do not, therefore, form new plutonium).

The drawbacks indicated in the previous section, i. e. the build-up of higher mass TRU elements and shortage of neutrons when the core is loaded with MA, have been confirmed by detailed studies (see, e. g., Refs 6 and 7). At present, even if some studies are still ongoing on thermal reactors (in particular there are recent studies related to CANDU reactors), most of the leading laboratories in the world have now focused their studies, both theoretical and experimental, on TRU transmutation in fast neutron spectrum reactors, FRs.

18.1.3 Radioactive waste management infrastructure

For countries that use nuclear power to generate electricity, waste storage may prove adequate for a time but what is actually needed is a permanent solution, namely a disposal facility. For LLW and LILW, a number of countries own and operate near-surface or intermediate-depth repositories, a few of which have been accepting waste since the 1950s. Over the years, public opinion has hardened against the opening of new facilities so that those that are in operation have come to be seen as valuable national assets and a crucial part of the nuclear infrastructure. It is common for governments to enshrine geological disposal of radioactive waste in national policy and to entrust a waste management organisation (WMO) to implement it. The WMO assumes the role of repository operator and is a key component of the waste management infrastructure.

Radioactive waste disposal is expensive. As yet, no facilities exist for deep disposal of SNF but for a medium-sized programme (5000 tHM) the cost of a deep repository is expected to be of the order of €5 Bn13 (2006 prices). This must ultimately be met from sales of electricity and, therefore, in advance of waste production. Normal procedure is to establish an oversight body with the power to create an accumulating fund that is dedicated to decommissioning and waste management. The fund is raised by some form of tax or surcharge on the price of electricity and the WMO may draw from it to pay for repository development.

A second important part of the infrastructure is the regulatory body, which should be independent of the repository operator. Regulations will be established to limit the hazards to both staff and public during the operational phase of a repository and after it has been closed. Often, the regulations governing repository operation will be established by the nuclear regulator while those applicable to the post-closure phase are set by the environmental regulator. Some jurisdictions place a limit on the time period over which post-closure safety must be demonstrated, others do not. Typically, peak doses tend to occur thousands, or even hundreds of thousands, of years after closure as a result of the migration of long-lived, mobile radionuclides such as technetium-99 and iodine-129 in groundwater. It follows that any ‘demonstration’ of post-closure safety will entail mathematical modelling based upon extrapolations of the past and present hydrogeology of the site.

For both operational and post-closure safety, an important consideration is that of optimisation of radiological protection and safety:

the magnitude of the individual doses, the number of people exposed and the

likelihood of incurring exposures [shall] all be kept as low as reasonably

achievable, economic and social factors being taken into account, within the

restriction that the doses to individuals delivered by the source be subject to dose

constraints14

Dose constraints for humans are usually set at 300 pSv per annum or less. This represents about one tenth of the average annual dose from natural background radiation. Because of the impossibility of knowing the habits (e. g. diet) of individuals living far into the future, dose calculations are usually performed for small groups of people who might be reasonably expected to be maximally exposed. A typical example is a group of people living in the repository discharge area and consuming produce raised on contaminated water.

The third component of the infrastructure is the waste producer — in most cases this is the NPP operator. These three bodies — WMO, regulator and waste producer — form the most important components of the waste management infrastructure as depicted in the so-called ‘classical triangle’15 shown in Fig. 18.2.

Once it is built and operating, an NPP will become subject to a different set of commercial risks. These include retrospective changes in policy or practice in regulation, electricity pricing, taxation, competition from other generators or even the rules governing NPP ownership. These may impact on the profitability of the plant and, naturally, will be borne by the plant sponsor or owner. During a 60 year plant life we may, for instance, expect some less profitable or even unprofitable years when electricity prices fall during an economic downturn. But the business case that was used to justify the construction of the plant at the outset should have included a market analysis that examined the probability and the consequences of, for example, an excess of supply over demand along with other possibilities such as changes in fuel prices (including those of the main competitors), carbon prices, electricity pricing, exchange rates, running costs and so on.

The sums of money that are involved in electricity generation are so large that, regardless of whether the plant sponsor is a private-sector or a state-owned utility, the making of a strong and robust business case is essential to justify the investment. Where the state is the owner, however, it does, at least, have the ability to frame the rules over such things as carbon pricing, electricity trading and taxation so that they work to its advantage. A private-sector utility, does not have this power of course. Rather, when considering the possibility of an investment it seeks assurances that, for example, the rules governing electricity trading and the outlook for carbon pricing will remain unchanged or will move in his favour for the foreseeable future and, further, that there is a consensus on this across the political divide. Governments will usually wish to see competition and diversity in the electricity market and may encourage new entrants through tax incentives and other ways of subsidising renewables, carbon capture, district heating, micro-generation or whatever. Such schemes may represent a threat to existing generators, who will always seek to ensure that the fundamental assumptions on which the original investment was made are not undermined.

This brief discussion makes clear why would-be generators of nuclear electricity spend years negotiating with governments on such matters as taxation, the rules for electricity trading, policy on reducing future CO2 emissions and the planning system. Governments on their part will aim for an electricity generation system that is least cost, secure, diverse and, increasingly, low in carbon emissions. Incentives that are currently used to encourage the growth of renewables may be extended to other low-carbon technologies such as nuclear and carbon capture.

UF6 is moved around the world in transport containers approved to international standards. Although minor variations exist in the container design, all those used for commercial transport are similar and consist of a carbon steel cylinder with a set diameter that depends on whether or not the UF6 is enriched. The most commonly used cylinder for natural or depleted UF6 is the 48Y, where the 48 signifies the diameter in inches and the Y signifies the particular design variation. The most commonly used cylinder for product is the 30B, a 30 inch diameter cylinder. The 48 inch feed cylinders hold around 12 500 kg of UF6, while the 30 inch product cylinders hold around 2200 kg of UF6.

A small heel is left in a UF6 cylinder after it has been emptied. The heel is usually less than 2 kg in weight and consists of non-volatile impurities and corrosion products, such as UF4 and iron fluoride. It may also contain elevated levels of thorium-234 (434Th), a daughter product of the 438U decay chain. As volatile UF6 is removed from the cylinder the non-volatile thorium remains and becomes concentrated by factors as high as 10 000. With uranium no longer present to act as a radiation shield the decay of this isotope and its immediate daughter, protactinium-234, can give rise to significant radiation levels, requiring that even nominally empty cylinders be treated with care; however, the 24 day half-life of 234Th means that radiation levels soon decay.

UF6 cylinders require periodic inspection and testing in order to demonstrate compliance with transport regulations and need to be washed and cleaned for testing. This is done using a water-based wash solution, which could contain additional cleaning agents, such as sodium carbonate and hydrogen peroxide. The process needs to be controlled for product cylinders so that there is no possibility of criticality as the water in the wash solution will act as a moderator. Cylinders are often not cleaned between inspections but rather have fresh feed or product material filled over the top of the heel.

Nuclear fuel is subjected to substantial forces and material changes, which affect the endurance of the fuel rods. Some phenomena are inevitable, but R&D has improved fuel performance considerably since the first commercial operation of nuclear reactors in the 1950s. In this section, the main phenomena affecting fuel rod endurance are briefly addressed and illustrated with examples from the experimental program of the OECD Halden Reactor Project where applicable. Associated modelling is treated in Chapter 14.

The extent to which phenomena affecting fuel rod endurance are allowed to develop is limited by safety criteria and rules for normal operation. When formulating safety criteria, the general goal is to ensure that nuclear power is used safely, but details and specific limits differ between countries. An overview is given by NEA (2003).

High neutron economy and good fuel utilization make the CANDU reactor complementary to LWRs, since the used fuel from the latter has more than enough fissile material to fuel a CANDU reactor. One such fuel cycle is RU from reprocessed LWR fuel. The use of RU in a CANDU reactor is simpler, more economical and derives more energy than would be obtained by re-enrichment followed by irradiation in an LWR. RU could be used as-is in a CANDU reactor, where a U-235 level of 0.9% would result in a burnup of about 14 MWd/kg HE. Alternatively, a simpler option would be to down-blend the RU with depleted uranium (DU) from enrichment plant tails to form ‘natural uranium equivalent’ (NUE) fuel. This would make use of DU, which otherwise is of little value, while the reactor licensing and operation would be essentially unchanged from the use of natural uranium. A demonstration irradiation of 24 CANDU NUE bundles successfully took place in 2010 in two separate channels at the Qinshan Unit 1 CANDU reactor in Haiyan, China (Jioa et al, 2009).

DUPIC (Direct Use of used PWR fuel In CANDU) is another example of a CANDU fuel cycle that is synergistic with the LWR. Used LWR fuel has about 0.9% U-235 and 0.6% fissile plutonium, giving a total fissile content of around 1.5%. The high neutron economy of the CANDU reactor results in the ability to use that material without removal of fission products. DUPIC involves thermal/ mechanical processing of used LWR fuel to convert the LWR pellets into new CANDU pellets, without selective removal of isotopes. In the DUPIC process, the cladding is first removed. The LWR pellets are then subjected to a series of oxidation and reduction cycles, which convert the pellets to powder, which can then be milled, if necessary, before the powder is pressed and sintered into new CANDU pellets. The pellets are loaded into new CANDU fuel sheaths. Of course, the energy derived from the used LWR fuel is not as great as it would be if some or all of the fission products were to be removed. However, the process has a high degree of proliferation resistance, is simpler and is expected to be more economical than conventional reprocessing. AECL and the Korean Atomic Energy Research Institute (KAERI) collaborated on the DUPIC cycle, along with the US Department of State. In AECL, three full-length CANDU DUPIC fuel elements were fabricated from used LWR fuel, and successfully irradiated in the NRU (National Research Universal) reactor (Floyd et al., 2003).

Whether or not DUPIC becomes commercialized, it is illustrative of the unique recycling opportunities between LWR and CANDU reactors.

Until recently, applications of MSRs have been limited to thermal-neutron — spectrum graphite-moderated technology. From 2005 onwards, studies have concentrated on creating fast-spectrum MSRs, i. e. MSFRs, which combine the efficiency of fast neutron reactors with the specific benefits of molten salt fluorides discussed in the previous section. MSFR systems have been recognized as a long­term alternative to solid-fuelled fast neutron systems. They offer several advantages, including negative feedback coefficients, smaller fissile inventories, easy in-service inspection and a simplified fuel cycle. The main characteristics of MSFRs are summarized in Table 13.8. The first system developed was the ORNL molten salt breeder reactor (MSBR) project (Nuttin et al., 2005; Mathieu et al, 2006). Since then a variety of core arrangements, reprocessing methods and salt compositions have been proposed, most noticeably a graphite-free core (i. e. with no graphite moderator) (Fig. 13.13).

Two types of fuel cycle have been suggested (Renault et al., 2009): [23] [24]

Table 13.8 MSFR reference design characteristics (Renault et al., 2009)

Reference value(s)

Thermal power (MWt) 3000

Fuel molten salt composition (mol%) LiF-ThF4_233UF4 or LiF-ThF4_(Pu-MA)F3 with

LiF = 77.5 mol%

Fertile blanket molten salt composition (mol%) Melting point (°C) Operating temperature (°C)

Initial inventory (kg)

Density (g/cm[25])

Dilatation coefficient (/°C) Core dimensions (m)

Fuel salt volume (m3)

Blanket salt volume (m3)

Thorium consumption (ton/year) 233U production (kg/year)

Breeding ratio (233U-started MSFR)

13.13 Molten salt fast reactor (MSFR).

©

produces an improved doubling time of 35 years. The existence of other fissile elements within the spent fuel would reduce 233U consumption, making MSFR systems more feasible (Renault et al., 2009).

The potential for spent fuel to lead to proliferation of nuclear weapons is clear. The primary means of preventing this is the safeguards system established under the Treaty on Non-proliferation of Nuclear Weapons (NPT) under which signatories agree to safeguards inspections of fuel and other nuclear material by the International Atomic Energy Agency (IAEA). The NPT does not, however, cover the safety of spent fuel management nor is safeguards information publicly available. As a means, therefore, of promoting safety and transparency in this area, the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management was introduced. This came into force in 2001 and aims to incentivize the achievement and maintenance of a high level of safety in the worldwide management of radioactive waste and spent fuel. Contracting parties agree to participate in review meetings, which occur approximately every three years and to prepare a national report for each meeting. These national reports describe national policy on spent fuel and radioactive waste management and are a useful source of information on the disposition and actual quantities of these materials held by the contracting countries; they can usually be found on the IAEA web site.

Under the NPT, non-nuclear weapons states agree not to develop nuclear weapons. In return for this undertaking, they are entitled to technical advice and support from IAEA when establishing their own (peaceful) nuclear power programmes. A number of other avenues for support now exist; however, these include GNEP (Global Nuclear Energy Partnership), which evolved to IFNEC (International Framework for Nuclear Energy Corporation), and a 2006 Russian initiative to establish (amongst other things) an International Uranium Enrichment Centre.

These aim to ensure that nuclear fuel and energy are accessible to all countries without a need for developing their own fuel enrichment and manufacturing industries. To relieve the burden (and the proliferation risk) of spent fuel management, concepts such as leasing of fuel or fuel take-back contracts are being considered; these are seen as particularly helpful for countries with small nuclear programmes. Information on these initiatives can be found at various internet sites.

We mentioned several technologies above. Three kinds of devices for liquid — liquid contact are generally used:

• Mixer-settlers

• Pulsed columns

• Centrifugal extractor

Mixer-settlers (Fig. 16.10) have been used for a long time in the nuclear industry and have some benefits (efficiency and flexibility) but have some drawbacks:

• Long contact times between the solutions (high residence time of the liquid phases) promote radiolysis of the solvent in highly active environments.

16.10 Mixer-settler principles (Source: AREVA, International Seminar on Nuclear Fuel Cycle, 19 October 2010, INSTN).

• Difficulty in achieving favourable geometry for high flow rates.

• Presence of impurities at the interfaces, which that tend to accumulate and are difficult to remove

Mixer-settlers are a series of mixing chambers, each with an impellor, combined with separation chambers for decanting.

A pulsed column (Fig. 16.11) is a several metres high vertical cylindrical or annular tube where liquid/liquid extraction is performed; two other volumes, again cylindrical or annular, located at both ends of the column are used for the separation of the two phases.

Pulsed columns are easy to use and efficient. The ability to continuously remove impurities from the interfaces is an essential advantage compared to a mixer-settler, as well as a lower residence time, which that reduces solvent degradation by radiolysis. Also, pulsed columns are prefabricated pieces of equipment and do not need maintenance. For these reasons, they are usually preferred over mixer-settler devices.

A centrifugal extractor (Fig. 16.12) employs a spinning rotor that first mixes and then separates the two phases inside the rotor where the centrifugal forces can be as high as 300 g, resulting in efficient and fast phase separation. The separated phases exit the contactor by overflow and underflow weirs, similar to a mixer-settler. The devices are compact and residence times are very low, which minimizes solvent degradation. They offer a wide range of capacities, and are geometrically safe up to significant capacity. On the other hand they are very sensitive to process disturbances such as the presence of crud or particles, which may settle and cause clogging. They are also mechanically complex, needing regular maintenance and significant heat is generated during mechanical mixing, which must be removed. Overall, the process advantages must be balanced against the greater complexity (Fig. 16.13).

16.11

Pulsed column photo and schematics (Source: AREVA, International Seminar on Nuclear Fuel Cycle, 19 October 2010, INSTN).

16.12 Centrifugal extractor schematics (Source: AREVA, International Seminar on Nuclear Fuel Cycle, 19 October 2010, INSTN).