Seven key technical domains have been identified,54 where important choices, feasibility assessments and new specific developments are needed: 1) choice of FR core conversion ratio, 2) reactor type, 3) heterogeneous versus homogeneous recycling, 4) transmutation fuels and targets, 5) MA content in the fuel, 6) recovery of grouped TRU versus MA/Pu separation and, possibly, selective MA separation, 7) aqueous versus pyrometallurgical reprocessing for transmutation fuels.

The corresponding ‘option tree’,54 is shown in Fig. 17.9.

In the following the options within the seven technical domains and associated challenges are briefly described.

National and international laws are based on the willingness of countries to commit themselves to a course of action. National laws typically require the passage of legislation or decrees and frequently require more detailed regulations to implement the very detailed requirements necessary in highly technical areas such as dangerous goods transport safety. International laws establish legal rights and obligations for the states that consent to be bound by them, and they may also require detailed implementation requirements. Both of these types of laws and their supporting regulations, standards and guidance are necessary to ensure a comprehensive system of transport safety for all dangerous goods, including for radioactive materials.

In 1998 the IAEA Secretariat prepared a report (IAEA, 1998), which provides additional detail about the international agreements in this area. It illustrates that, although both the IAEA Transport Regulations and the UN Model Regulations are viewed as recommendations, their provisions become binding through various international instruments. Through this process, a sound international transport safety regulatory regime has been established.

The attachment to GOV/1998/17 (IAEA, 1998), combined with updated information through the year 2003, identified a significant number of such binding international instruments and agreements that directly or indirectly apply to the safe transport of radioactive materials:

• 21 worldwide instruments in force

• five worldwide instruments that have been prepared but were, at that time, not yet in force

• 22 regional instruments in force

While the number of binding instruments is large, there are two worldwide modal conventions and several major regional conventions that provide the most comprehensive coverage of dangerous goods transport safety. The two worldwide conventions are:

1 For maritime transport of dangerous goods — The International Convention for the Safety of Life at Sea and three protocols (London, 1974) and other actions taken by the International Maritime Organization (IMO) required that state parties make the IMO International Maritime Dangerous Goods Code mandatory by 1 January 2004 (IMO, 2010).

2 For air transport of dangerous goods — The Convention on International Civil Aviation — Annex 18, the Chicago Convention (the Safe Transport of Dangerous Goods by Air (Chicago, 1945)), requires state parties to make the International Civil Aviation Authority (ICAO) Technical Instructions for the Transport of Dangerous Goods by Air mandatory (ICAO, 2010).

As international land transport (i. e. transport by rail, road and/or inland waterways) is de facto limited to continental traffic, there is no global convention governing the carriage of dangerous goods by these modes. However, regional agreements exist, including:

• the European Agreement concerning the International Carriage of Dangerous Goods by Inland Waterways (ADN), for which the UN/ECE serves as secretariat

• the European Agreement concerning the International Carriage of Dangerous Goods by Road (ADR) ECE/TRANS/175, for which the UN/ECE serves as secretariat

• Regulations concerning the International Carriage of Dangerous Goods by Rail (RID), for which the Intergovernmental Organisation for International Carriage by Rail (OTIF) serves as secretariat

Figure 19.2 shows how the IAEA Transport Regulations and UN Model Regulations are implemented into these binding instruments.

General

This chapter aims to illustrate the characteristic features of the economics of nuclear power and we do this through comparisons with coal, gas and onshore wind generation. The numerical results are produced by combining Eq. 5.3 with data that are (mostly) extracted from the European and North American data provided by PCGE for nuclear, coal, coal plus carbon capture (coal+CC), gas and onshore wind. Other schemes are omitted from the present discussion.

Discount rates

Equation 5.3 allows any discount rate to be used so long as it is constant with time. In PCGE, calculations are performed for two annual discount rates, namely 5% and 10%. These figures, it is suggested, may be viewed as broadly representing the cost of capital under two different market conditions. For the first, it is considered that the state is the prime mover in the investment. In the second, this role is taken by the private sector. It is argued that different rates apply because private investors will invariably demand a higher return on capital than the state. And, while some have argued that rates should be higher, these two values do appear to span the normally used range. Thus, recent analyses of electricity generation costs have used values of 7.5%6 and 10% ‘as advised by DECC’6 (UK Department for Energy and Climate Change). In the case of nuclear power, where capital costs dominate, the discount rate is one of the most important variables. Here, to avoid proliferation of the calculations, we assume a single constant discount rate of 7.5%. This is chosen to represent a position

half-way between the 5% for state and 10% for private finance examined by PCGE. Section 5.2.6 examines the sensitivity of the calculations to this parameter.

6.1.2 Tailings management and mine rehabilitation

From open cut mining, there are substantial volumes ofbarren rock and overburden waste. These are placed near the pit and either used in rehabilitation or shaped and revegetated where they are.

Uranium minerals are always associated with other radioactive elements such as radium and radon in the ore, which arise from the radioactive decay of uranium over a few million years. Therefore, although uranium itself is barely radioactive, the ore which is mined, especially if it is very high-grade such as in some Canadian mines, is handled with some care, for occupational health and safety reasons.

Mining methods, tailings and run-off management and land rehabilitation are subject to government regulation and inspection. For instance in Australia the code of practice and safety guide: Radiation Protection and Radioactive Waste Management in Mining and Mineral Processing was published in 2005, updating previous versions.

Solid waste products from the milling operation are tailings, ranging in character from slimes to coarse sands. They comprise most of the original ore and they contain most of the radioactivity in it. In particular they contain all the radium present in the original ore. At an underground mine they may be first cycloned to separate the coarse fraction, which is returned underground and used for underground fill. The balance is pumped as a slurry to a tailings dam, which may be a worked-out pit as at Ranger and McClean Lake, or an engineered structure.

When radium undergoes natural radioactive decay one of the products is radon gas. Because radon and its decay products (daughters) are radioactive and because the ground rock comprising the tailings is now on the surface, measures are taken to minimise the emission of radon gas. During the operational life of a mine the material in the tailings dam is often kept covered by water to reduce surface radioactivity and radon emission (though with lower-grade ores neither pose a hazard at these levels). This water needs to be recycled or evaporated since it contains radium, which is relatively soluble. Most Australian mines and many others adopt a ‘zero discharge’ policy for any pollutants.

On completion of the mining operation, it is normal for the tailings dam to be covered by some two metres of clay and topsoil with enough rock to resist erosion. This is to reduce both gamma radiation levels and radon emanation rates to levels near those normally experienced in the region of the orebody, and for a vegetation cover to be established. At Ranger and Jabiluka in North Australia, tailings will finally be returned to the mine pit or underground, as was done at the now-rehabilitated Nabarlek mine. In Canada, ore treatment is often remote from the mine that the new ore comes from, and tailings are emplaced in mined-out pits wherever possible, and engineered dams otherwise.

Radon occurs in most rocks and traces of it are in the air we all breathe. However, at higher concentrations than are likely to be found at any modern mine it is a health hazard. Underground, especially in high-grade orebodies, it is removed by mechanical ventilation.

After mining is completed at established ISL operations, the quality of the remaining groundwater must be restored to a baseline standard determined before the start of the operation so that any prior uses may be resumed. Usually this is potable water or stock water (usually less than 500 ppm total dissolved solids). Contaminated water drawn from the aquifer is either evaporated or treated before reinjection.

In contrast to the main US operations, the water quality at the Australian sites is very poor to start with, and it is quite unusable. At Beverley the original groundwater in the orebody is fairly saline and orders of magnitude too high in radionuclides for any permitted use. At Honeymoon the original water is even more saline, and high in sulfates and radium. When oxygen input and leaching are discontinued, the water quality reverts to its original condition over time.

Upon decommissioning, ISL wells are sealed or capped, process facilities removed, any evaporation pond revegetated, and the land can readily revert to its previous uses.

Mining is generally considered a temporary land use, and upon completion the area with any waste rock, overburden and covered tailings needs to be left fit for other uses, or its original use. In many parts of the world governments hold bonds to ensure proper rehabilitation in the event of corporate insolvency.

In the manufacture of fuel containing U-233, some handling issues arise that are beyond those associated with fabrication of fuels containing recycled plutonium, which is generally mixed with depleted uranium (MOX fuels). This is mainly due to the fact that U-233 is always associated with U-232, which, as mentioned in Section 8.1.2, has a high specific activity because of its radioactive daughters, especially Tl-208. Hence, U-233 handling and processing require additional biological shielding (compared with plutonium) so that fuel manufacture and all handling operations through to reactor charging must be done remotely. Thus U-233 fuel fabrication and handling is a major technical hurdle and constitutes one of the main drawbacks of the Th/U-233 closed fuel cycle, because it generates a significant cost penalty. It must be added that reprocessed thorium also contains Th-228 and Th-234, the first of which also has the hard gamma emitter, Tl-208, as a daughter, preventing direct handling for tens of years.

It may be possible to reduce the shielding requirements if fuel fabrication is started promptly after reprocessing before there has been significant ingrowth of the U-232 daughters. However, this strategy might be risky in the case of potential delays in the fabrication process as these might allow the gamma dose rate to reach unacceptable levels. This topic has to be studied in detail to assess the real impacts of various recycling strategies on handling issues.

The reactor operates as a recirculation boiler with boiling occurring within the core. The RPV of a BWR6 is shown in Fig. 10.11. The BWR RPV is taller than

that of a PWR to accommodate the steam separators and driers. The pressure (nominally 7.2 MPa) is lower than for a PWR and the vessel is constructed from both forgings (in the lower regions nearer to the core) and welded plate. The reactor internals are supported by the lower head and/or vessel walls.

The overall fluid flow is illustrated in the schematic (Fig. 10.2). The coolant water boils in the core producing a two-phase steam/water mixture, which exits the core, is separated and the steam dried. Steam (>99.9% quality) leaves the vessel and enters the turbine. The condensate is pumped back into the vessel entering the interspace between the vessel wall and the core shroud. Flow through the core is controlled by the means of recirculation and jet pumps. There are two recirculation loops each of which supplies ten jet pumps in a semicircle around the lower part of the downcomer. The later ASEA BWRs used directly powered internal jet pumps with motor assemblies mounted on the lower head of the RPV. This design was adopted for the Advanced BWR (ABWR), see Section 10.11. The core design is discussed in 10.8, below.

Water quality is very important in the BWR because the steam produced is directly used to power the turbine generator. A reactor coolant cleanup system takes water from the recirculation system and RPV bottom head as illustrated in Fig. 10.12 . This is passed through filters and demineralisers to remove fission products, corrosion products and other impurities. The purified water is then returned to the feedwater system.

BWR6

Reactor assembly

1. Vent and head spray

2. Steam drier lifting lug

3. Steam drier assembly

4. Steam outlet

5. Core spray inlet

6. Steam separator assembly

7. Feedwater inlet

8. Feedwater sparger

9. Low pressure coolant injection inlet

10. Core spray line

11. Core spray sparger

12. Top guide

13. Jet pump assembly

14. Core shroud

15. Fuel assemblies

16. Control blade

17. Core plate

18. Jet pump/recirculation water inlet

19. Recirculation water outlet

20. Vessel support skirt

21. Shield wall

22. Control rod drives

23. Control rod drive

24. In-core flux monitor

GENERAL ф ELECTRIC

In line with the designs of the more modern Magnox reactors, all AGRs are constructed with a pre-stressed concrete pressure vessel (PCPV).

The reactor core is contained within a single cavity in the centre of the PCPV. It consists of a 16-sided stack of graphite bricks on a square lattice, interconnected with graphite keys to provide stability. The graphite acts as the core moderator (see Section 12.4). The bricks have central holes through them into which fuel assemblies are loaded and control rods can be inserted. Different AGRs have different numbers of fuel channels. Hinkley Point B and Hunterston B have 308 channels per reactor, Hartlepool and Heysham 1 have 324, Heysham 2 and Torness have 332 but Dungeness B has 408.

There are graphite reflectors (to minimise neutron escape) around the core. Outboard of the reflectors there are shields, usually of steel and graphite although Hinkley Point and Hunterston use calcium hydroxide, to reduce neutron radiation levels such that access to the outer parts of the reactor cavity is possible when it is shut down. The shields also limit radiation damage to reactor components made of steel, which cannot be removed, and to the PCPV.

The boilers on all AGRs, excepting the Hartlepool and Heysham 1 sister stations are also contained within the PCPV cavity. Part way through construction of the earlier stations, accelerated corrosion tests on some steels used for boiler tubing showed that they might be life-limiting. Removal of the boilers through the PCPV is not impractical but would be very difficult and time consuming. Because of this, the Hartlepool and Heysham 1 reactors were designed so that the boilers could be replaced if corrosion became an issue. This design is such that the boilers are set into deep, circular pits, called pods, within the walls of the PCPV. The closures above the pods were removable to allow access to and replacement of the boilers. However, in practice, the boilers have not been corroding as quickly as the accelerated tests predicted and the Heysham 2 and Torness boiler designs reverted to the Hinkley Point/Hunterston arrangement. Furthermore, safety concerns expressed by the UK regulator over the possibility of failure of the closures above the boiler pods in the Heysham/Hartlepool arrangement led to modifications, which would make it very difficult to remove a boiler.

Figure 12.3 shows a cross section of a Hinkley Point B type design with integral boilers. In the pod boiler design used at Hartlepool and Heysham 1, the boilers are set into vertical cylindrical ducts (the pods) in the concrete of the PCPV.

High pressure CO2 is used to cool all AGRs. Most AGRs operate at 4 MPa (40 bar) pressure, Dungeness B being the exception with a pressure of 3 MPa. The gas is pumped around the reactor by large circulators, contained in penetrations through the PCPV at the bottom of the reactor. From the gas circulator outlet, the gas is discharged into a lower plenum below the core where the flow divides with approximately half going directly up the core over the fuel with the residual (called re-entrant flow) being directed up an annulus outside the core, returning downwards through passages in and between the graphite bricks and thence to the fuel channel inlets. This gas cools the neutron shield for the boilers (the boiler shield wall) and the graphite core.

At one stage, as the AGR design was developed, it was thought that oxidation by the CO2 would limit the operation of the plant to below design output. As a consequence, a study was initiated to consider the possible use of helium as an alternative coolant. However, it was concluded that the helium in an AGR would still require the addition of a gas such as CO2 in order to produce an oxide layer on steel surfaces to reduce the friction between moving parts. The idea was therefore abandoned. A conceptual design was also proposed for a helium-cooled

fast reactor using essentially AGR PCPV technology, but was abandoned due to the size of the core and potential difficulties with the helium leak rate.

Gas emerging from the top of the core passes down the boilers and is pulled into the circulators. The typical design gas outlet temperature from the core is 650 oC and the outlet from the boilers is 300 °C.

The boilers are a once-through design consisting of re-heater, super-heater, economiser and evaporator sections. There are penetrations through the PCPV for each boiler section through which the steam passes. Boilers in most AGRs were of serpentine-wound sections in mild steel, low-alloy austenitic or stainless steel depending on operating temperature. In the podded Heysham/Hartlepool design, helically wound boilers were used.

There is a steel gas baffle between the neutron shielding and the boilers. On all AGRs except the Hartlepool and Heysham 1 pair, the baffle consists of a vertical cylindrical wall topped by a welded dome, which is provided with holes that align with the channels in the core. The dome is in the plenum above the top neutron shield and below the underside of the top cap of the PCPV. The baffle is effectively the boundary between the hot and cold gas. At Hartlepool and Heysham 1, with their pod boiler design, the gas baffle is a curved plate, which spans the upper plenum above the top neutron shield and is fixed, around its circumference, into the PCPV concrete.

The weight of the core is taken on a plate (the core support plate) beneath the lower core reflector. This plate is held in place by the diagrid, which is an open lattice steel structure to allow open coolant flow to the core channels. The weight of the diagrid is taken by supporting struts, which are anchored into the PCPV.

The PCPV is lined with insulation covered by steel plates, which are bolted to the concrete. The liner serves as a gas-tight containment membrane. There are a number of penetrations through the PCPV, which serve the boilers, the circulators and the fuel and control rod channels. Below the liner are cooling water pipes, which maintain the concrete temperatures at acceptable levels.

Helium is produced in the fuel pellets/bars during irradiation via three mechanisms: (i) ternary fission; (ii) neutron capture by 1 6O (in oxide fuel); and (iii) alpha decay of transuranic isotopes, principally 242Cm (Kamimura et al., 1999). The generated helium atoms can then diffuse to intra-granular and grain face fission gas bubbles and to fuel surfaces in the same way as xenon and krypton fission gas atoms. However, due to the small size of the helium atoms, diffusion (at least in oxide fuels) is more rapid than that of fission gas, and so the fraction of the generated helium that is released is larger than that for fission gas. In plutonium-free fuels the helium generation due to (iii) is negligible and so helium generation and release can generally be ignored. In contrast, in plutonium-bearing fuels the helium generation due to (iii) can be significant, and helium release can provide an important contribution to the pressurisation of the fuel pin.

One of the side results of reprocessing activities is generation of high-level radioactive waste. This is the concentrate of the aqueous side streams from nuclear fuel reprocessing. Besides this major source of HLW there may be some other wastes that fell into this category12 but their volume contributions would be lower and the conditioning and packaging may differ in each specific case. The reprocessing waste consists of fission products and minor actinides but contains also the remnants of U and Pu isotopes (up to 1%). 1 3 Reprocessing waste is typically vitrified and stored until final disposal facilities become available. Figure 15.17 shows an

example of vitrified HLW conditioned and packaged in a metal canister for disposal. Until final disposal the HLW is stored in storage facilities. A photo of such storage facility in La Hague is shown in Fig. 15.18. A schematic of the vitrification process and interim storage of HLW in La Hague is shown in Fig. 15.19.15 The canisters with vitrified waste are stored stacked on top of each other in vaults that contain a number of channels. The photo in the Fig. 15.18 shows the tops of the vaults.

One of the advantages of vitrified waste is that the leaching rate is very low even if the containers and overpack were damaged. Disposal of HLW will be in the deep geological repositories similar to or using the same technology as for spent fuel disposal. There are several design concepts for disposal of HLW at various levels of development but they are in most cases tied to the concept for spent fuel disposal as they require a similar geological environment.

As the majority of major actinides are removed from the HLW its radiotoxicity is also lower and it takes a shorter period of time to reach the baseline radiotoxicity of uranium ore as shown in Fig. 15.20. This level of radiotoxicity of uranium ore may be considered as a reference to the natural background. The radiotoxicity of spent fuel is almost an order of magnitude larger to begin with and in addition it takes almost two orders of magnitude less time for HLW to reach the radiotoxicity of uranium ore.

As the radiotoxicity of HLW is after some 400 years dominated by actinides (tails of major actinides from reprocessing and minor actinides), efficient removal of actinides by partitioning and transmutation would simplify the disposal requirements but those technologies are still under development.

Air cooling

Dust removal

Heat and activity will decrease naturally over time

Sampling

Interim storage

Decontamination

Supply

15.19 Vitrification process and interim storage for HLW in La Hague, France.

15.20 Relative radiotoxicity of spent fuel and HLW over time.

458 Nuclear fuel cycle science and engineering

In Japan, the Tokai-Mura plant began reprocessing in 1975 (200 t/year); the Rokkasho plant is currently undergoing hot commissioning (800 t/year) and is scheduled to go into operation at the end of 2012 depending upon the successful start of the vitrification process. Tokai, JNC (now JAEA)2 has operated a 90t/yr pilot reprocessing plant using PUREX technology and treated 1116 tonnes of used fuel between 1977 and its final batch early in 2006. It processed over 1000 tonnes of used fuel, with a Pu-U mixed product. The plant will now focus on R&D, including reprocessing of MOX fuel. JAEA operates spent fuel storage facilities there and is proposing a further one. It has also operated a pilot high-level waste (HLW) vitrification plant at Tokai since 1995. Tokai is the main site of JAEA’s R&D on HLW treatment and disposal.

Until a full-scale plant was ready in Japan, the reprocessing of used fuel was largely undertaken in Europe by BNFL and AREVA (4200 t and 2900 t respectively), with vitrified high-level waste being returned to Japan for disposal. AREVA’s reprocessing finished in 2005, and commercial operation of JNFL’s reprocessing plant at Rokkasho-mura was scheduled to start in 2008. Used fuel has been accumulating there since 1999 in anticipation of its full-scale operation (shipments to Europe ended in 1998).

Reprocessing involves the conventional PUREX process, but Toshiba is developing a hybrid technology using this as stage 1 to separate most uranium, followed by an electrometallurgical process to give two streams: actinides (plutonium and minor actinides) as fast reactor fuel, and fission products for disposal.