The design of molten salt reactors is similar to those for other thermal reactors, including similar neutron fluxes and reactivity temperature coefficients (Rosenthal et al., 1972). In particular, the graphite-moderated MSR has much in common with the HTR: a graphite moderator at an average temperature of 600-700 °C, Th-233U fuel cycle, and similar fuel-moderator ratios. However, there are some important differences. Table 13.9 shows some relevant parameters for MSR, using 233U or 235U as the fissile material. The main parameter shown in this table is the critical fuel concentration; to obtain the same reactivity only half as much 233U is needed as 235U, because of the superior neutronic characteristics of 233 U (Rosenthal et al., 1972). That is an important feature to take into account for the design of reactors that use Th as fertile material.

7LiF-BeF2 (66/34 in mol%) salt is used as the fuel carrier for the moderated (thermal) molten salt thorium breeder, producing 7LiF-BeF2-ThF4-UF4 as fuel salt. All of the alternatives to this salt reduce the breeding capacity of the reactor

(Renault et al., 2009). SFR systems use a molten salt actinide burner. A carrier salt with good solubility for actinide trifluorides is needed; something which can be achieved using 7LiF-NaF-(KF) as the solvent or 7LiF-(NaF)-BeF2 melt. An interesting alternative is the use of Pu and MAs as start-up fuel for the Th cycle in the MSR, leading to 7LiF-NaF-ThF4 carrier salt (Renault et al., 2009). A single stream Li, Na, Be/F molten salt actinide recycler and transmuter (MOSART) fast spectrum system fuelled by combinations of Pu with MA trifluorides (AnF3) from UOX and MOX has been developed in Russia (Ignatiev et al., 2005). A modified two-fluid Th-U molten salt system (based on MOSART) has also been developed more recently (Feynberg and Ignatiev, 2010). A comparison between MOSART, MSBR and MSFR is provided in Table 13.10 . In order to summarize different potential combinations, some reference salts and fuel compositions are reported in Table 13.11.

Salt processing relies on both on-line and batch processes to maintain smooth reactor operation while minimizing losses to waste streams. The removal of lanthanides is necessary because of their low solubility in the molten salt and their adverse effect on reactivity through neutron capture (Renault et al., 2009). A potential processing scheme is shown in Fig. 13.14 . The main innovation is stages 2 and 3, which combine chemical and electrochemical extraction methods with the back extraction of actinides and lanthanides. This allows fuel processing to occur with no variation in effluent volume, while reducing the fuel processing balance to just one reaction: 2LnF. + 3H. O(g) = Ln. O3 + 6HFfe). An effective method for actinide/lanthanide separation is still needed (Delpech et al., 2008). A practicable fuel clean-up rate is 40 l per day, corresponding to the

processing of 100 kg heavy nuclei per day. This value is almost two orders of magnitude lower than that needed by the reference MSBR scheme (Renault et al., 2009).

As the isotopic composition of spent fuel is dependent on the burnup, higher burnups will result in somewhat different characteristics of spent fuel. Higher burnup will result in increased levels of fission products, degraded composition of uranium isotopes and increased levels of transuranics, Mainly Pu.3 By degraded composition of uranium it is meant that concentrations of the U-232 and U-236 will increase and, due to higher initial enrichment, the concentration of residual U-235 will also be higher until a burnup above 70 GWd/tU when it starts to decrease. This effect (shown in Fig. 15.4) occurs because of the current limits on initial enrichment of fuel to 5%. The isotopic composition will

affect later spent fuel management, which may be either reprocessing and recycling or disposal. Higher concentrations of transuranics are important for the safety analysis of further spent fuel management steps as well as the usability of spent fuel for recycling (i. e. U-232 contributes to radioactivity since it is a precursor to the high-energy gamma emitter Tl-208 and the non-fission long-lived neutron absorber U-236).

The higher enrichment of fresh fuel will also raise concerns about criticality safety in spent fuel management where, as is sometimes the case, the (simplifying and conservative) assumption is made that this is unchanged by irradiation. As enrichment increases, therefore, the benefit to be gained from ‘burnup credit’ also increases. Increased enrichment also typically results in higher contents of fissile material in the spent fuel with burnup in the vicinity of 70 GWd/t. Applying burnup credit for such nuclear fuel would show more realistic safety margins for storage and disposal and may also have an impact on the cost of storage, transportation and disposal system designs.

Figure 15.5 shows the change in the concentration of the various plutonium isotopes with burnup (the higher burnups above 70 GWd/tU were test burnups). Figure 15.6 shows the transuranics inventory of spent fuel in relation to burnup. Pu-238 inventory is an interesting case: in-reactor it is generated from Am-238 and has a relatively short half-life (87.7 years) and high heat generation because almost all its decay is through alpha emission. High concentrations of Pu-238 in spent fuel may render the extracted plutonium unattractive for use as a nuclear weapon because this high heat generation makes it unmanageable.

15.5 Concentration of plutonium isotopes in spent fuel as a function of burnup.3

The specific activity of the fission products in spent fuel is almost directly proportional to the discharge burnup. So, for example, the activities of Sr-90 and Cs-137 would double for double the burnup.

Other physical-chemical characteristics of spent fuel are mostly the result of the isotopic composition in addition to the radiation exposure of, for example, the cladding and other non-fuel materials. The key characteristics are:

• decay heat

• radioactivity of spent fuel

• gas and volatile radionuclide build-up in the fuel pellet and properties of the fuel pellets (formation of pellet rim with high porosity)

• cladding properties (rod growth, clad hardening hydrogen build-up)

All these characteristics may lead under adverse circumstances to increased incidences of rod failure at higher burnup either in the reactor or subsequent storage.

15.3.2 Decay heat

Decay heat also increases linearly with fuel bumup. Most decay heat during fuel storage (first 20 years) is from fission product beta particles with less from actinides. Later, this contribution is shifted towards predominantly alpha-emitting actinides. Figure 15.7 shows decay heat dependence on spent fuel burnup showing the contributions of fission product to the total decay heat. It shows also decay heat for different cooling periods (5-200 years).

Solvent treatment

Used solvent contains degradation products and residual radioactive impurities. The main degradation product from TBP is di-butyl phosphoric acid, whose sodium salt is soluble in the aqueous phase. So, the solvent is re-generated using alkaline solutions: before recycling, the solvent is washed with sodium carbonate and sodium hydroxide (Fig. 16.17).

To reduce the losses of TBP in the aqueous phases but also to prevent potential reactions of the degradation products with nitrates during the concentration

16.16 Overview of the whole PUREX process including the routes for wastes (Source: AREVA).

16.17 Solvent management (La Hague reprocessing plant UP3) (Source: AREVA, International Seminar on Nuclear Fuel Cycle, 19 October 2010, INSTN).

through evaporation of the refined products, the aqueous phases coming the extraction are stripped by a thinner (hydrocarbide C12H26), which removes residual TBP and TBP degradation products.

The development of P&T strategies aims to bring benefits to waste management. Furthermore, by deploying FRs as part of this strategy, it will be possible to realize a significant increase in the amount of energy that is extracted from uranium. Inevitably, the extraction and increased use of plutonium within the nuclear fuel cycle raises the issue of proliferation. Of the P&T impact studies quoted above, however, none attempts a comprehensive assessment. Evaluation methodologies for proliferation resistance and physical protection have been developed44-50 but to date only very preliminary attempts have been made to apply them to specific advanced fuel cycle strategies. Early indications42 are that the introduction of FRs would appear to be best for proliferation resistance, based on an extension of the ‘Technological Opportunities to increase the Proliferation resistance of global civilian nuclear power Systems (TOPS)’ methodology.4 1 In this TOPS report,5 1 transmutation technology is picked as one of several specific technical options for reactor and fuel cycle systems that have been proposed to improve proliferation resistance. However, it should be noted that ‘no single diplomatic, military, economic, institutional, or technical initiative alone will be able to fully deal with this proliferation challenge. The best prospect for achieving non-proliferation goals while expanding nuclear power is to engage all appropriate means’.52

The transport of dangerous goods has been subject to regulation for many years. National regulations governing these materials are known to have existed over 225 years ago. However, not long after the end of World War II, inter-modal problems were increasingly being encountered where dangerous goods were trans-shipped. It was recognized that in the interests of safety and commercial economics, dangerous goods transport safety regulations should be harmonized both among the various modes of transport and internationally.

While the varieties of all nine classes of dangerous goods that are transported represent a wide spectrum of potential hazards during transport, there are similarities in the controls that need to be exercised to ensure their safe transport and to facilitate domestic and international movement. These materials must be suitably classified based on their potential hazard during transport, packaged commensurate with their hazard; and information must be communicated about their potential hazard (including emergency measures) to carriers, handlers at facilities and potential emergency responders.

Although radioactive materials present unique hazards during transport, they are included in the overall system of dangerous goods transport safety. They are included as one of the nine classes of dangerous goods that warrant regulation. Radioactive material is denoted as Class 7 in the international regime of regulating the packaging and transport of dangerous goods. This allows the radioactive materials to be shipped commercially, and also facilitates the application of these materials to beneficial uses.

The harmonized system of regulatory control that has evolved over the ensuing years is based on a combination of national and international instruments. The need for national laws and regulations that are compatible with the international regulations and standards has given rise to a highly interactive global system in which member states of international organizations, and the international organizations themselves, perform key roles. Together, these complementary regulatory systems provide an integrated network of requirements to ensure safety during the transport of dangerous goods.

Commercial competition should ensure a continued drive for cost reductions and it is clear that, for nuclear power, the area to focus on is capital costs and the reduction of commercial risk (and therefore discount rates) through replication. But competition cannot exist without a market so let us assume that, in the next 20 years:

• governments of European and North American countries continue to exert pressure to cut back on carbon dioxide emissions

• renewables, for whatever reason, amount to no more than 20% of nominal generating capacity

• oil becomes scarcer and more expensive

• no new intensive energy source for carbon-free electricity generation (e. g. laser fusion) becomes available.

With these assumptions it seems inevitable that nuclear power will provide an increasingly large slice of electricity generation, not only for heating, cooling, lighting, machines and the many other uses with which we are familiar but also as a substitute for fossil fuels by, for example, supplying the energy for battery — or hydrogen-powered vehicles. This would be supported by CCGT and, depending on the success of development efforts, coal-fired generation combined with carbon capture and sequestration. At the present time, however, the latter looks too expensive and will therefore need some form of government support. Obvious candidate countries for the introduction of coal+CC are those with access to low — price supplies of coal not least because successful implementation of the technology could help to support continued coal exports.

Such an expansion of nuclear power will reduce capital costs through the force of competition and the economies of scale; it will also considerably reduce uncertainties in cost estimates, many of which arise because of the long layoff from reactor construction. Other costs that are likely to be reduced through improved design are operation, maintenance, decommissioning and waste management. Fuel costs will probably rise as demand for uranium increases but, as we have seen, nuclear fuel costs constitute only around 12% of the LCOE and the cost of uranium itself constitutes about one third of this. Interestingly, a recent report19 has suggested a $210 per lb maximum price for uranium based on the cost of extraction from seawater. If this became reality, fuel costs would then constitute about 25% of the LCOE for nuclear. Even before this happened, however, wider use of MOX fuel would probably be an important factor in limiting fuel prices. More widespread use of nuclear will also increase the need for plant that has an improved ability to load follow.

Partly driven by the dearth of orders for large-scale plant, much thought has been given in recent years to the production of small modular reactors that would be capable of being constructed off-site, a practice that is normal for nuclear submarine reactors. Economies of scale would not come from the size of the plant but from the numbers produced. Many such reactor designs have been proposed: the World Nuclear Association website20 lists 16, most with an output in the range 100-300 MW(e). It is claimed that savings can be made on capital cost because these designs would allow simpler safety systems. If overnight cost could be reduced that would, indeed, be an advantage. From the sensitivity study in Section 5.2.6, however, we can see that equally powerful economic arguments might be made based on a shortening of the construction time and a reduction in discount rate because of the scaling down of the overall size and cost of the project.

D. GRENECHE, NuclearConsulting, and M. CHHOR, AREVANP, France

Abstract: This chapter compares the use of thorium as a nuclear material with conventional fissile materials. Rather than being a real alternative, thorium is a complement to the current uranium/plutonium fuel. Despite the benefits of thorium, its use presents technical challenges, which are described in this chapter. This review shows that significant experience has been gained on thorium-based fuel in both test reactors and power reactors, but not on an industrial scale.

Key words: thorium, uranium, nuclear power.

8.1 Reasons for considering the thorium cycle

8.1.1 Thorium fuel as an alternative or complement to uranium and plutonium fuels

Almost all of the world’s nuclear reactors in operation today use U-235 to sustain the neutron chain reaction because this isotope is the only naturally occurring isotope, which is fissile by slow neutrons (the fuel is then either natural uranium or, in most cases, uranium enriched in U-235). In this fuel, neutron captures by the fertile material U-238 produce plutonium and, in particular its two thermally fissile isotopes Pu-239 and Pu-241, which are burned partly in situ (typically, about half is consumed in a standard light water reactor). The plutonium that remains in the discharged fuel may be separated by reprocessing and then recycled in reactors in the so-called MOX fuel cycle, which leads to a saving in the use of natural uranium of about 12% for a single plutonium recycle.

As with uranium, thorium is also a naturally occurring material but it contains only one isotope, Th-232, and this is not thermally fissile, although it is a fertile isotope. Therefore, thorium is only useful as a resource for breeding new fissile materials, in this case U-233, which, for reasons explained later, is the best fissile isotope in the thermal neutron spectrum. Furthermore, a neutron chain reaction can only be sustained with thorium if sufficient quantities of fissile materials are available (U-235, U-233, plutonium) and mixed with the thorium. Then, it becomes possible to operate a nuclear reactor with thorium in which U-233 is produced, and, by doing so, the so-called ‘thorium cycle’ would be initiated. As with plutonium, U-233 is partly burnt in reactors and what remains in the discharged fuel can be recycled. However, the potential role and attractiveness of
thorium-based fuels within the nuclear enterprise will depend on the implementation and deployment scenarios, which must take into account economic factors and strategic choices.

Rather than being a real alternative to natural uranium and plutonium fuel, thorium is more accurately considered as a complement to the current uranium/ plutonium fuel. Indeed, it may, for example, be used to increase the available quantity of fissile material by the production of U-233 during irradiation, provided that enough fissile material is initially available to supply the neutrons needed for this breeding process. It can also provide possible avenues towards longer fuel cycles and higher burn-ups as well as multiple recycling of uranium/plutonium fuels in thermal spectrum reactors. Introducing thorium and thus, via irradiation, U-233, in such fuels may also allow the amount of natural uranium needed to be reduced.

The main phenomena limiting cladding endurance are waterside corrosion and hydrogen pick-up. The latter is closely linked to corrosion since the hydrogen absorbed by the cladding stems mainly from the corrosion process itself. Cladding corrosion is primarily a concern for PWRs because of their higher coolant temperature.

Waterside corrosion will make the cladding wall thinner and thus reduce its load-bearing capability. A thin oxide layer is in fact protective, but if it grows too much, it will crack and eventually spall. This happens when the oxide thickness

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9.13 Fuel centre temperature response to rod overpressure and clad lift-off.

approaches 100 pm, and regulation usually limits cladding corrosion to this amount (NEA, 2003).

Some fraction (10-20% for Ziy-4) of the hydrogen generated in the oxidation process is picked up and diffuses into the metal matrix. At the operating temperature of the cladding (PWR: about 350 °C), the solubility limit of hydrogen in Zircaloy is about 100 ppm. For concentrations above this amount, hydrogen precipitates as hydride plates (ZrH^ 66). In cladding tubes, hydrides will normally have a circumferential orientation. A high concentration of hydrides will cause embrittlement and weaken the load-bearing capability of the cladding. Regulation usually limits the average hydrogen concentration to 500-600 ppm (the corresponding average oxide thickness is approximately 50-60 pm).

Since the oxide of zirconium has a much lower thermal conductivity than the metal, a spot with somewhat lower temperature will arise where the oxide layer has spalled. Hydrogen that was picked up tends to migrate to these cold spots and may reach concentrations of several thousand ppm, causing severe hydriding and weakening of the cladding. It has been found that spalling with consequential hydride concentration and material embrittlement considerably lowers the failure limit in reactivity insertion accidents (Papin, 2003; Vitanza, 2007).

The nuclear industry has responded to these issues by improving the original Zry-4 (which itself is improved over Zry-2 regarding corrosion and hydrogen pick-up) and developing new alloys with better corrosion resistance.

228 Nuclear fuel cycle science and engineering

The majority of PWR designs use unshrouded fuel assemblies. In these the core effectively consists of an open array of fuel rods. The fuel assemblies consist of square arrays of pins. The number of pins and the pin diameters vary but a typical 17 x 17 fuel assembly is illustrated in Fig. 10.9. The fuel rods are held on 17 x 17 grids. This provides 289 positions. The central position is reserved for in-core instrumentation and another 24 positions are occupied by thimble tubes. These provide part of the skeleton of the fuel assembly and are structural elements, which are joined to the top and bottom nozzles as well as the spacer grids. The reactor control rod cluster assemblies (see Fig. 10.9) are inserted into these tubes in some of the assemblies. The control rod positions are fixed by the design of the plant but as all the fuel assemblies can accommodate RCCAs this allows flexibility in the placement of the assemblies in the core loading pattern.

The fuel rods themselves consist of zirconium alloy tubes into which are inserted pellets of enriched UO2. The rods incorporate a gas plenum to accommodate fission gases produced during irradiation. In general two lengths of fuel rod are commonly used, corresponding to active core heights of 12 foot (‘standard’) or 14 foot (‘XL’), or their metric equivalent. In the Westinghouse design these cores can be accommodated in the same vessel. The XL core needs a longer core barrel, which extends into the lower head region. PWRs can also use mixed oxide (MOX) fuel where plutonium oxide provides the initial fissile loading rather than 235U. Because plutonium has a lower delayed neutron fraction than uranium the shutdown margin is affected and if more than about 30% of the core is loaded with MOX then additional RCCAs are required.

The enrichment used depends on the target burnup and the fuel cycle length. Originally PWRs were designed to run on a 12 month cycle and achieve a burnup of about 30 GWd/t. Each fuel assembly stayed in the core for three cycles and a third of the core would be changed at each refuelling outage. The factor determining the burnup and hence dwell time was largely the performance of the fuel cladding. For all plants which refuel periodically rather than on-load, the reactor must be designed to cope with a number of failed fuel rods, since it would be uneconomic to shutdown, to replace fuel failures, on an individual basis. At one time the design basis was specified in terms of being able to operate with a very small percentage (~0.25%) of fuel failures. In practice it is now controlled by limiting the maximum allowable activity levels in the primary coolant and operators demand very high fuel reliabilities. Operating with failed fuel makes maintenance more complex and increases operator radiation exposure. Cladding materials have been developed to give greater resistance to radiation effects, which has allowed burnups to be extended to more than 60 GWd/t. Modern cladding materials are based on zirconium alloys. Stainless steel has been used in the past, in some fuel, but the performance of zirconium seems to be superior.

Within the constraints provided by the fuel design, the actual fuel cycle used is largely a matter of economics. Long fuel cycles mean that the ratio of generation time to refuelling time is increased. However, increasing the cycle length increases the enrichment required and also increases the number of fuel assemblies which have to be replaced at each refuelling. Small batch sizes tend to allow more efficient fuel utilisation. The replacement of individual assemblies when they reach optimum burnup is ideal, but can only be practically achieved by on-load refuelling, which is not possible for PWRs. Thus the fuel costs tend to be higher for long fuel cycles, but the average outage costs are lower. In addition there is a practical limit on the maximum enrichment currently used. Most fuel fabrication plants and transport containers are designed for enrichments of up to 5%.

Cycle lengths of between 6 months and 2 years are currently used. The short cycles are used by some German Konvoi plants, which were designed to allow rapid refuelling. The commonest cycle lengths are 12 and 18 months. Eighteen months is widely used, particularly in countries which have peaks of demand in both summer and winter, since the outages can be alternated between spring and autumn when demand is lower.

Safety case issues have much in common with those of the AGR, discussed further in Section 12.4. Issues specific to Magnox plant include:

• air ingress (leading to oxidation of the fuel and core) and loss of cooling capacity following failure of boiler ducts (not possible where boilers are contained within the pressure vessel)

• brittle fracture of the pressure vessel

• single — or multi-channel fires involving the oxidation of cladding and fuel (and potentially channel blockage)

• reactivity faults, including local removal of control rods

In general, gas-cooled thermal reactor designs are much less sensitive to loss of coolant sequences than are water-cooled reactors due to the combination of the inherently low heat transfer between fuel and coolant combined with the very high thermal capacity of the core including moderator. This makes rapid response much less of an issue than in PWR and BWR designs.