The UF6 produced by the enrichment process is subsequently converted into nuclear fuel. It is therefore important that the material supplied to the fuel manufacturer meets quality standards that allow the manufacturer to operate its process safely and successfully and to generate product that is suitable for onward use. Through the participation of industry experts, ASTM International has developed a series of standards relating to the nuclear fuel cycle, which have been widely adopted by the nuclear industry. There are standards relating to UF6 quality and others concerned with methods for sampling and analysis.

There are two main methods used to sample UF6:

1 As a gas during processing

2 As a liquid, from a batch

Gas sampling is relatively simple in that it involves tapping into the process line or container and diverting gaseous UF6 either to an on-line measurement instrument or to a sample point. The sample is withdrawn using low pressure (effectively sucking on the source) and may be collected by cooling a sample receipt container so that the UF6 freezes out. The operation of cooling also drives gas transfer and one common method for doing this is to immerse the container in liquid nitrogen. Sample containers can be of different designs and materials of construction depending on the analysis required. Stainless steel is perfectly acceptable for most analyses, including measurement of the proportion of uranium isotopes (the isotopic abundance). Nickel, monel (a high nickel alloy) or plastics such as polychlorotrifluoroethylene (PCTFE) may also be used.

Gas sampling works for isotopic abundance testing but it has limited value for checking the levels of contaminants in feed or product as they will not have the same volatility as the UF6 and may therefore be increased or decreased with respect to their true concentration during the sampling process. Proportional gas sampling is possible as a container is filled but requires considerable care in order to ensure that the sample is truly representative of the container contents. Measurement of contaminants in a batch of UF6 therefore typically requires liquid sampling to achieve homogeneity, which in turn requires that sampling be undertaken at pressures and temperatures above the critical point. The UF6 is sampled by pouring from a batch into a sample bottle. In a conversion plant the uranium hexafluoride is sampled at the end of the manufacturing process, where it is liquefied in large batches before being decanted into transport containers for delivery. In an enrichment plant the product container is heated to liquefy the contents in a purpose designed sampling autoclave and a sample decanted. In both cases the sample will be taken in industry standard sample containers known as 1S or 2S bottles. These are basically a monel or nickel bulb with a tube and valve arrangement fitted at one end. The 2S bottle is a little larger than the 1S bottle and may be used to take samples of up to 2.21 kg, while the 1S bottle is used for samples of up to 450 g.

The 1S or 2S bottle requires sub-sampling to provide samples that can be prepared for analysis. The bottle is again heated to liquefy the contents and to help ensure that the sub-sample is representative and a sample is tapped off into a smaller sample container, usually a P10 or P25 tube. These are again industry standard sample bottles made of PCTFE with a lid and seal arrangement. A P10 tube holds approximately 10 g of UF6 whilst a P25 tube holds around 25 g. Some sub-samples may be sealed at this point to be used for independent analysis or arbitration, whilst other samples will be immersed in water to produce a hydrolysed solution of uranyl fluoride and hydrofluoric acid. This solution is used for subsequent analysis.

The three main things that a customer is interested in is how much uranium they are getting, what enrichment it is and what contaminants are present that could interfere with the next stage of the fuel cycle. Measurement of uranium content is typically carried out using a redox titration while the proportion of 235U and 234U in the uranium is measured using a mass spectrometer. High precision is normally required for this measurement so that a sophisticated magnetic sector instrument is likely to be used. A number of different techniques may be used for measurement of contaminants depending on the equipment available within the laboratory and the contaminant being measured. Mass spectrometry can be used for many contaminants with a standard quadrupole unit likely to offer sufficient capability. ASTM International has published a number of standards for routine measurement methods.

Mixed oxide or MOX fuel as used in present light water reactors means a mix of uranium and plutonium oxides. Thorium can be a constituent of mixed oxide fuel as well, but is presently not used on a commercial basis.

The pellet fabrication steps are similar to those of UO2 fuel, but the powder preparation must necessarily differ. The entire fuel powder, pellet and rod fabrication has to occur in shielded facilities. By and large, two methods are in use for powder preparation and mixing. The technology developed by former BNFL (Weston et al., 2001) is called the ‘Short Binderless Route’ (SBR). The UO2 and PuO2 are mixed in the right quantities to obtain the desired Pu content of the final product in one step. The blend is attrition-milled to break down agglomerates and to produce an intimately mixed and micronised powder feed for pellet pressing.

Another technology is the ‘Micronised Master Blend’ (MIMAS) (Vliet et al., 1996). In this case, a master blend with a plutonium content of about 30% is produced and micronised by attrition milling. This blend is then diluted with UO2 powder to the desired plutonium content and homogenised. The process is used

by AREVA (Arslan and Krellmann, 2006), has been selected for the US ex-weapons plutonium disposition, and is being implemented in Japan for recycling of spent Japanese fuel.

A number of mechanisms provide for reactivity control during normal operation and in postulated accidents in the CANDU reactor. Most of the reactivity control devices are vertically oriented and are located between fuel channels in the unpressurized, low-temperature moderator. (Liquid poison injection nozzles in the second shutdown system are located horizontally between fuel channels.) This removes high pressure as a possible driving force in accidently ejecting control rods from the reactor, and significantly reduces the environmental degradation of the mechanical components of the control and safety systems.

On-power refuelling maintains the core reactivity close to zero and compensates for the reactivity loss due to fuel burnup. In a CANDU 6 reactor, liquid zone controllers provide fine reactivity control. These controllers comprise vertical tubes containing light water, which is an absorber in a CANDU reactor core. The light water level in the tubes can be varied to provide bulk reactivity control and power shaping over 14 regions of the core, thus compensating for reactivity and local power changes as burnup proceeds and refuelling takes place, while also ensuring that bundle and channel powers are kept below specified limits.

The CANDU reactor employs a system of absorber rods (called adjuster rods), which provides a coarser range of reactivity control to compensate for the change in absorption in saturating fission products (primarily xenon) during power changes, such as during load following or reactor restart after a shutdown. The adjuster rods are made of stainless steel or cobalt and are normally in the core. Their withdrawal after an unanticipated shutdown, to compensate for build-up of xenon, provides a certain ‘decision and action’ time for restarting the reactor. The adjuster rods also shape the radial and axial power distributions with natural uranium fuel. In the CANDU 6 reactor, there are 21 vertical adjuster rods arranged axially in three rows.

Four mechanical control absorbers are parked above the CANDU 6 reactor core, and supplement the negative reactivity worth of the zone control system. They are inserted to achieve reductions in power at pre-determined rates, and compensate for the reactivity increase due to a reduction in fuel temperature at low power.

Soluble neutron absorbers (boron or gadolinium) can also be added to (and removed from) the moderator for additional negative reactivity.

CANDU reactors have two different, fast acting, independent shutdown systems, which are separate from the control system. The first shutdown system consists of spring-driven vertical shutoff rods, which can be dropped into the low — pressure moderator. The second shutdown system consists of a dissolved neutron absorber, which can be quickly injected into the moderator through horizontal nozzles located between the fuel channels that run the width of the core. Both systems act independently and both are capable of meeting safety requirements in shutting down the reactor. Coolant void reactivity is positive in a CANDU reactor with natural uranium fuel. Hence, a postulated large loss of coolant accident (LOCA) largely determines the requirements of the shutdown system and the emergency core cooling system. If a LOCA were to occur, the neutronic pulse would provide a fast trip signal, which would trigger the shutdown system and quickly turn over the power pulse, meeting safety requirements. In these circumstances, safety is aided by a large prompt-neutron lifetime (about 9 ms with natural uranium fuel, about 40 times longer than in a LWR), which slows down the reactivity (and power) transient. All reactors have postulated accidents that can insert positive reactivity, often without the mitigation of a long prompt — neutron lifetime (Meneley and Mujumdar, 2009; Mujumdar and Meneley, 2009) and normal practice is to utilize a combination of inherent features and engineered systems to cope with these.

The two other main safety systems in a CANDU reactor are emergency core cooling, in the event of the loss of the normal primary heat transport system cooling, and the containment building.

There are large volumes of water present in CANDU reactors, which would provide an ultimate heat sink in the case of a severe accident — heavy-water moderator surrounding the fuel channels in the calandria and light water in the shielding tank, which surrounds the calandria.

The CANDU reactor has, for a long time, included sophisticated digital computer control of the reactor, and a system of in-core, self-powered neutron flux detectors that provide an accurate and fast measure of the flux distribution in the reactor core. Different types of detectors are used in the reactor control and shutdown systems. Many other system parameters are also monitored in the control and safety systems.

The supercritical sater-cooled reactor (SCWR) is a high-temperature, high — pressure water-cooled reactor, which operates above the thermodynamic critical point of water (above 374 °C, 22.1 MPa) (Fig. 13.10) . SCWRs are based on existing advanced Gen III water-cooled reactors as well as developments in supercritical water power cycle technology in such sectors as the coal industry. SCWRs are basically LWRs operating at higher pressure and temperatures with a direct, once-through cycle. The coolant remains single-phase through the system. There is no need for recirculation using jet pumps, pressurizer systems, steam generators, separators and driers, which results in plant simplification, albeit at the expense of increased temperature and pressure. The major components of the power conversion cycle external to the reactor vessel are similar to supercritical fossil-fired boilers.

The SCWR uses either pressure vessel or pressure tube boundaries for the supercritical water in the core (Starflinger et al., 2008) (see Fig. 13.11). The higher of the two outlet temperatures, 625 °C, affords a thermal efficiency approaching 50%, which compares very favourably to the ~33% efficiency of today’s LWRs. The high-pressure single-phase coolant provides another advantage over current technology, because it circumvents the need for steam generators and allows the use of an off-the-shelf advanced power turbine. Combined, these factors could potentially reduce capital costs by up to 40%. The main advantage of the SCWR is thus a lower operating cost because of higher thermal efficiency and a simpler

design made possible by the use of a well-established, high-temperature, single­phase coolant.

It is hoped that SCWRs will support the next generation of baseload electricity suppliers. The overall GIF plan for the SCWR is to complete the operation of a fuelled loop test by about 2015, with a view to construction of a prototype sometime after 2020. The SCWR can be designed as a fast or thermal reactor with a closed or once-through fuel cycle. In addition, pressure-vessel or pressure-tube designs offer a number of design options that have the potential to meet the GIF’s criteria (Khartabil, 2009) (see Table 13.7).

Research in developing the technology will need to focus on areas such as safety, sustainability, proliferation resistance and physical protection (Khartabil, 2009). A key challenge is the selection of materials for the core components (a replacement will be needed for Zircaloy cladding for example) and in demonstrating core power stability; thermal-hydraulics data will be needed to design and license the reactor (Khartabil, 2009; Abram and Ion, 2008). Material selection for the reactor core (fuel cladding and other components) will need to take account of creep, oxidation and stress corrosion data.

13.11 SCWR pressure vessel baseline alternative (Starflinger et al., 2008).

Z. L O VA SIC, International Atomic Energy Agency (IAEA), Austria

Abstract: This chapter considers the types of nuclear reactor and the spent fuel waste produced. The various technologies for spent fuel storage, reprocessing, recycling and disposal are reviewed. The vitrification of high-level radioactive waste (HLW) is described.

Key words: spent nuclear fuel storage, vitrification, high level nuclear waste.

15.1 Types of nuclear reactors and nuclear fuel arisings

Since 1960 when the first commercial pressurized water reactor (PWR) of 250 MWe, Yankee Rowe, designed by Westinghouse, and the boiling water reactor (BWR), designed by General Electric, were put into operation in the USA, the quantities of spent fuel occupying the spent fuel pools of commercial reactors have steadily increased. While, from the beginning, PWR and BWR designs used enriched uranium fuel, Canada adopted a different approach using natural uranium fuel with heavy water moderator and coolant; its first commercial CANDU unit started operating in 1962. The Soviet Union also developed a pressurized water reactor, known as the VVER, and later (1973) the RBMK high power channel reactor. In contrast, the magnesium-clad fuel from the first generation of gas-cooled reactors, a number of which were also built between 1958 and 1971, was generally sent for reprocessing recognizing that it was unsuitable for long-term storage in water. In addition, over this same period, several experimental or prototype fast reactors operated. All but one of these (BN-600 in Russia) are now shut down. Many countries have chosen light water designs so that today 60% of the world nuclear capacity is PWRs and 21% BWRs.

Table 15.1 shows the types of nuclear reactors that make up the current total of 441 operating nuclear reactors (another five reactors are in a long-term shutdown condition). As PWR and BWR spent fuels are predominant in the quantities of arising spent fuel, most examples in this chapter are related to these two types of fuel. Yearly discharge from all these reactors was about 10 200 tonnes of heavy metal in 2010 and was about 10 500 t (HM) during several years prior to that year.

Uranium is separated from plutonium using the selective reduction of plutonium to the trivalent state (Fig. 16.8). This is done by adding four-valent uranous nitrate to the aqueous feed to produce the reaction

2Pu4+ + U4+ + 2H2O ^ 2Pu3+ + UO22+ + 4H+

In the trivalent state, plutonium has little ability to make neutral complexes with nitrate ions so that it has a higher affinity for the aqueous phase. The uranium, on the other hand remains in the organic phase in the six-valent state.

This plutonium reduction from valence IV to valence III is a delicate one within a nitric acid solution because plutonium III can be oxidized to plutonium IV by nitric acid through complex reactions where autocatalysis of nitrous acid plays a major role. This is prevented through the addition of hydrazine nitrate, which stabilizes uranous nitrate (U(NO3)4) but prevents parasitic re-oxidation of plutonium by nitrous acid:

HNO2 + NH2NH3+ ^ HN3 + 2H2O + H+

possibly followed by the reaction:

HNO2 + HN3 ^ N2 + 2N2O + H2O

The aqueous phase containing plutonium is scrubbed by a small flow of solvent to remove the fraction of uranium that is stripped with the plutonium (uranium scrubbing). The organic and aqueous phases containing, respectively, uranium and plutonium can be separated using a mixer-settler and a pulsed column. Uranium is then concentrated through evaporation. Uranium and plutonium are then transferred to the purification stages.

The objective here is the management of spent fuel inventories, as a legacy of previous operation of nuclear power plants, both for a strategy of continuation of nuclear energy, based only on LWR reactors, and for one of nuclear energy phase-out. As for reprocessing, a ‘grouped TRU recovery’ (i. e. TRU recovery without separation of Pu from MAs) is likely. As dedicated reactors, both ADSs and low CR critical FRs can be envisaged, with a MA/Pu concentration ratio -0.1 in the fuel.

The scheme of this scenario is presented in Fig. 17.6.

In the phase-out case, the scenario would lead to a reduced repository burden and a low final inventory of TRUs. If implemented by a country in isolation, however, this strategy would require a substantial deployment of new installations (fuel reprocessing and fabrication, ADS, etc.). Moreover, after ~100 years of operation ~20% of the initial TRU inventory (i. e. the last transmuter inventories) would be left in the wastes.27 A better approach would be to conceive ‘regional’ P&T scenarios,28 with fuel cycle facilities shared among different countries, possibly with different strategies in terms of future nuclear energy utilization.

A number of countries, including Australia, Russia, South Africa and the USA, have practised and, in some cases, continue to practise, disposal of radioactive waste in boreholes or shafts of various sizes. In the case of the USA, this even extends to transuranic waste. In addition, the prospect of using very deep boreholes for disposal of spent fuel has continued to attract attention over a period of more than 30 years (e. g. Ref. 39). More recently, however, of boreholes have come to be seen as a potential solution for the long-term safety of disused sealed sources. In particular, the IAEA has developed the so-called BOSS system, 1 0 which aims to provide a safe, permanent solution for those many countries that use sealed sources but have no nuclear infrastructure that would enable them to be managed safely after use. A particular problem has arisen with radium sources that, until 30 or 40 years ago, were widely used in medical therapy. Because the half-life of radium-226 is 1600 years, the radioactivity of these sources has hardly diminished in the interim and they now constitute a dangerous waste and a potential security hazard.

The BOSS system provides double containment of disused sources in stainless steel, burial at more than 30 m depth and encasement in concrete. Under anaerobic conditions (commonly found at depth), the combination of stainless steel and concrete is capable of providing absolute physical containment for at least 200 000 years, which is more than sufficient to allow radium-226 sources to decay to below exemption level. The system is small scale and economical. Considering the deaths that have occurred from loss of control of disused sealed sources over the years, it has the potential to provide significant benefits in terms of public safety and security.

The main by-product source of uranium today is at Olympic Dam in South Australia, where low concentrations of uranium (0.025 to 0.050%U) occur with copper grading about 1.8%Cu. Present production there is about 3500 tonnes of uranium, but there are plans to increase annually this to 16 000 tU/yr.

Following primary crushing underground, the ore is ground and treated in a copper sulfide flotation plant. About 80% of the uranium minerals remain in the tailings from the flotation cells, from which they are recovered by acid leaching as in a normal uranium mill. The copper concentrate is also processed through an acid leach to recover much of the other 20% of the uranium. The pregnant liquor is then separated from the barren tailings and in the solvent extraction plant the uranium is removed using kerosene with an amine as a solvent. The solvent is then stripped, using an ammonium sulfate solution and injected gaseous ammonia. Ammonium diuranate is then precipitated from the loaded strip solution by raising the pH, and removed by centrifuge. In a furnace the diuranate is converted to uranium oxide product.

However, after the secondary recovery by acid leaching, some uranium remains in the copper concentrate as it proceeds to be smelted. Typically it would have 45% Cu and up to 0.15% uranium, and the uranium is recovered in the further copper processing. This creates a safeguards problem if the smelting and electro­refining is not done at the mine site.

In the past some uranium has been recovered as a by-product of phosphate production, and this is set to be revived, with new technology, on an increased scale. Phosphate rock (phosphorite) is a marine sedimentary rock, which contains 18-40% P2O5 , as well as some uranium and all its decay products, often 70 to 200 ppmU, and sometimes up to 800 ppm. The phosphate rock is treated with sulfuric acid to give gypsum and phosphoric acid, and the uranium is normally recovered from the phosphoric acid by some form of solvent extraction (SX). A new process — PhosEnergy — uses ion exchange (IX) and promises to reduce recovery costs significantly.

The potential amount of uranium able to be recovered from phosphoric acid streams is over 11 000 tonnes U per year (global P2O5 production in 2010 was 33.6 Mt). The economic benefit will be both in the value of the uranium and in reduced regulatory demands on disposal of low-level radioactive wastes arising from the process. Estimated uranium production costs will put the new process in the lowest quartile of new uranium production.

6.2 World uranium production and demand.

8.1.3 The front-end of the thorium fuel cycle

Mining and milling

Since no thorium-based fuel is being used at an industrial or commercial scale in the world today, there is no international market for thorium and it can be assumed that a supporting mining industry will not develop until the demand for thorium fuel increases. Nevertheless, thorium is still being used today for various specific applications,[16] and is generally obtained as a by-product of uranium and, more especially, of rare-earth mining. As such, there is no real need to develop a specific thorium mining industry given that enough thorium is, for the time being, generated as a by-product. The available mining experience allows one to make some of the following observations.

The primary source of thorium is the rare-earth and thorium phosphate mineral, monazite. Without demand for the rare earths, monazite would probably not be recovered for its thorium content. Other ore minerals with higher thorium contents, such as thorite, would be more likely sources. However, mining of monazite deposits is easier than that of uranium-bearing ores. Very little overburden is removed as monazite is produced from beach sands or placer deposits.

Thorium is found in a number of minerals. Hence there are several process alternatives, like physical and magnetic separation and heavy-metal chemical extraction. Monazite, the chief commercial ore from which thorium is extracted, is chemically inert and any chemical treatment for extracting thorium must initially be very severe to achieve the complete dissolution necessary for the separation of the rare-earth elements, uranium and phosphates. The most common dissolution processes use highly concentrated sulphuric acid or highly concentrated sodium hydroxide.

Monazite is pulverized and leached in a 50-70% solution of hot sodium hydroxide and undergoes solvent extraction, stripping operations and ion exchange to obtain thorium nitrate, which is ultimately converted to thorium oxide powder. Those operations are necessary to obtain the level of purity required for nuclear applications. Like uranium, thorium is naturally radioactive but the ‘radon impact’ from processing thorium ores is easier to handle because its radioactive daughter thoron (Rn-220) is shorter lived (its half-life is 56 s) than its radon counterpart from uranium milling operations (Rn-222 with a half-life of 3.8 days).