Mechanical design codes are used to evaluate the mechanical behaviour of the fuel assembly. They often also have the functionality to be used for thermal analyses if the temperatures of the assembly structural components (i. e. everything excluding the fuel pins) are of interest. The standard technique employed is the finite element method, where each component of the assembly is discretised into a number of volume elements. Given suitable material properties, and any (potentially time dependent) external loading, the (potentially time dependent) stresses, strains and displacements applicable to each finite element are then determined by solution of the underlying matrix equations for the force balances, stress-strain relations and strain-displacement relations.

The primary application of mechanical design codes is to calculate the stresses imposed by the loads applied to the various assembly components (during normal operation, anticipated operational occurrences and accidents). Other uses include vibrational mode and harmonic response analysis, and buckling assessments.

Mechanical design calculations have historically been performed using in-house codes; more recently, ‘off the shelf’ commercial finite element software packages have tended to be employed.

418 Nuclear fuel cycle science and engineering

The separation is based upon hydrometallurgical techniques of liquid-liquid extraction in which material is transferred between two immiscible liquids: aqueous and organic phases. Extraction consists in passing the materials to be separated from the aqueous phase to the organic phase (solvent). The reverse

operation is called re-extraction. These operations are the core of the PUREX process, Fig. 16.6, whose main steps are:

• uranium and plutonium separation from fission products by extracting jointly uranium and plutonium from the clarified aqueous solution (Fig. 16.7)

• separation of uranium and plutonium (U/Pu partition)

The nitric acid solution that contains the nuclear material is mixed with the solvent (TBP diluted in a C12 alkane with a formula close to kerosene). TBP is selected for its very low miscibility in aqueous solution, its selectivity with regards to uranium and plutonium, its resistance to hydrolysis in a nitric acid environment, its resistance to radiolysis and its ability to be regenerated.

Uranium and plutonium migrate to the solvent while fission products stay within the aqueous phase (nitric acid). Settling and decanting then allows the two phases to be separated. This operation is repeated several times. The fission products that remain dissolved in the nitric acid are then concentrated through evaporation and stored in tanks prior to vitrification. Traces of fission products and minor actinides are still present in the solvent containing uranium and plutonium; they are removed through additional washings with nitric acid.

In this case the multi-recycling of TRUs in FRs is considered as the most appropriate strategy, due to the possibility of increasing U utilization by a factor of >50 and of using a large variety of fuels loaded with TRUs. The scheme in Fig. 17.4 summarizes the features of this scenario and indicates some of the most outstanding issues at each step of the fuel cycling. As indicated, both homogeneous and heterogeneous recycling options can be implemented in this scenario. For homogeneous recycling, it would be sufficient to chemically recover TRUs without separating out the MAs. For heterogeneous recycling, however, it would be necessary to separate Pu (or Pu+Np) from MAs. In this case the MAs could be kept together or separated into Cm and Am to allow Cm to decay in a dedicated facility.

Homogeneous recycling requires the TRU-bearing fuel to be fabricated at an industrial scale, as required by a fast reactor fleet deployment, albeit at a low MA content (< 5%). The alternative, heterogeneous recycling, concentrates the MA/ TRU-bearing fuel in a separate and, in principle, smaller fuel cycle, although with higher MA content and, consequently, higher post-irradiation neutron emission and decay heat (see Section 17.3.3).

It has been shown that this scenario (whatever the recycling mode) would allow the waste radiotoxicity in the repository to decrease to the level of the original uranium ore after 200-300 years.2’ 25

In designing repositories for HLW and spent nuclear fuel, an important consideration is heat output. During the operational period, temperatures can be controlled by ventilation but, after backfilling and closure, this is no longer possible and temperatures will rise. To avoid damage to the engineered barriers, it is usual to specify a limiting peak temperature on the surface of the waste canister (often 100 °C). This is implemented by increasing the spacing between adjacent waste packages so that the volume of rock that is available to absorb and conduct away the heat is sufficiently large to keep temperatures at acceptable levels. Increasing the size of the repository in this way is expensive and may also rule out some otherwise acceptable sites. Both are reasons why heat-producing waste may need to be cooled for decades before it can be disposed.

Long-lived ILW, which also requires deep disposal, has a much lower heat output so that, even when the waste packages are stacked so as to completely fill the emplacement tunnels, the temperature increases are no more than a few tens of degrees. This reduces the cost of deep disposal for this type of waste.

Some ore, usually very low-grade (below 0.1%U), is treated by heap leaching. Here the broken ore is stacked about 5 to 30 metres high on an impermeable pad and irrigated with acid (or sometimes alkaline) solution over many weeks. The pregnant liquor from this is collected and treated to recover the uranium, as with ISL, usually using ion exchange. After the material ceases to yield significant further uranium, it is removed and replaced with fresh ore. Recoveries are typically 50-80%. The depleted material has the potential to cause pollution so must be emplaced securely so as not to affect surface water or groundwater. Usually this will be in mined-out pits.

If sulfides are present the main agent is the bacterium Thiobacillusferrooxidans, and this generates acid from the sulfide, so no further acid make-up may be required. Some aeration is required, however, so the broken ore must be coarse enough to allow this, and the process takes longer.

The initial driver for thorium fuel development was to provide an alternative fuel cycle in anticipation of a projected rapid growth in nuclear power and possible shortage of natural uranium. An added stimulus was thorium’s supposed abundance in nature, based on the fact that the average concentration in the earth crust is approximately three times that of uranium, as mentioned above. Further, by the mid-1970s, the uranium price reached $40.00/pound U3O8 and this resulted in a perceived shortfall of low-price uranium based, in part, on one large nuclear power plant vendor being unable to meet uranium supply commitments to its customers. Along with the abundance of thorium in nature and breeding U-233, there were a number of other reasons at that time for rising interest in the thorium fuel cycle. Some of them included:

• the absence of uranium resources but large amounts of identified thorium resources in some countries having an ambitious civil nuclear programme, such as India

• the good in-core neutronic and physical behaviour of thorium fuel under irradiation

• a lower initial excess reactivity requirement (higher thermal conversion factor) of thorium-based cores using particular configurations

Thus, as illustrated in Table 8.1, the feasibility of different types of reactors based on Th fuels has been successfully demonstrated and significant experience has been accumulated so far, theoretically as well as practical and engineering-wise.

By the early 1980s, a number of factors had essentially killed enthusiasm for alternative fuel cycles. First, interest in the nuclear option waned significantly, especially in the US where public support for nuclear power dramatically declined following the Three Mile Island event of 1979. This anti-nuclear trend intensified and was further exacerbated in Europe by Chernobyl, seven years later. Second, starting in the early 1980s, the price of uranium remained low for over two decades so that, again, there was less interest in developing alternative fuel cycles. A contributing factor was the introduction into the market of down — blended uranium obtained from nuclear weapon disarmament programs (e. g., the US’s collaboration with Russian in the Megatons to Megawatts Program). Third, by the end of the 1970s, the Ford and Carter administrations had put an end to commercial reprocessing in the US so that it no longer had the capability to recover the fissile material from any non-military used fuel, let alone thorium-based fuel. Finally, there were proliferation concerns because, at that time, the reference option for implementing the thorium cycle was to deploy it with HEU. Not only is HEU chemically separable from thorium (assuming seed and fertile material are combined), but some fuel designs completely separated the HEU driver fuel from the fertile thorium. Consequently, the infrastructure needed for large-scale commercialization of thorium fuels never came about.

In the last decade, however, there has been a revival of interest in thorium — based fuels. This seems to have been initially motivated by the development of a LWR proliferation-resistant fuel cycle (i. e. the Radkowsky Concept),[15] and also by the so-called nuclear renaissance and resource scarcity that it might entail. It was also stimulated by some of the same factors that were the drivers for thorium cycles development in the 1950s and 1960s. These new factors vary from country to country, of course, but they include:

• The potential for a low production of plutonium and minor actinides in thorium based-fuel cycles. This is explained by the lower position of thorium in the Mendeleev’s table.

• The capability of destroying plutonium by fissioning it in a plutonium/thorium cycle in thermal reactors. These investigations include advanced reactor concepts based on thorium fuel cycles for future nuclear applications such as LWRs, HTRs, molten salt reactors (MSRs), accelerator-driven systems (ADS) and even fusion blanket systems.

• Transmutation of minor actinides.

• The possibility of breeding fissile isotopes (i. e. a conversion factor greater than one) with a thorium cycle in some thermal reactors such as MSRs, which is one of the concepts included for Generation IV systems.

• More recently, the dramatic increase in the price of uranium, which is closely tied to the perceived shortage of this material in light of a rapid growth of nuclear energy especially in Asian countries.

In this chapter we will look at some of these points in more detail.

Because of its long-term prospects, thorium continues to be studied. There is even currently a new upswing of interest in thorium both within academic institutions and R&D organizations but more importantly by industry. Indeed, some utilities as well as fuel vendors are revisiting thorium to investigate the industrially viable paths for the use of thorium as a complement to uranium/plutonium in LWRs, the main goal being savings in the usage of natural uranium when market conditions might render this thorium option viable. In Japan the HTTR could well be used in the future with thorium (as well as HTR-10 in China). Furthermore, India is still considering thorium as an industrial fuel for use in the not too distant future.

The PWR was originally developed as a submarine propulsion unit. As such it was designed to be small and responsive. Larger versions were developed for surface ships but these became the basis for the commercial PWR. The prototype was the Shippingport reactor (230 MW thermal, 60 MW electrical), which was developed as a joint AEC (US Atomic Energy Commission)/vendor (Westinghouse)/utility (Duquense Light Company) project. Over time other manufacturers have entered the market and there are a number of different designs but the basic concept remains the same. For the purposes of this chapter the modern Westinghouse four — loop design will mainly be used to illustrate the design characteristics.

10.3.1 Reactor Coolant System (RCS)

Figure 10.3 shows a typical Westinghouse four-loop nuclear steam supply system (NSSS). The NSSS consists of a reactor pressure vessel (RPV) containing the reactor core, which is connected by pipework to a number of steam generators, four in this case. Reactor coolant pumps (RCPs) located in the return (cold) legs of the circuit provide the circulation of water through the system. Connected to one of the RPV

outlet (hot) legs is the pressuriser. This is the only part of the circuit where a free surface exists.

The pressure in the circuit is maintained at a level such that boiling is virtually suppressed. (A small amount of nucleate boiling may occur in the top of the core.) This is achieved using the pressuriser, which is partly water filled with a steam ‘bubble’ above it. The pressure of the circuit is thus set by the saturation pressure of the water in the pressuriser, which is higher than the rest of the circuit. The pressure is controlled by means of electrical heaters in the bottom of the pressuriser, which can increase the saturation temperature, and sprays in the top, which spray cooler water from the cold legs into the steam space to reduce the pressure. The pressuriser is therefore at a higher temperature than the rest of the circuit and under normal operation a small spray flow is maintained to ensure a slow outflow of fluid into the circuit to establish a temperature gradient along the pressuriser surge line. The pressuriser is also fitted with power-operated relief valves to control larger pressure variations and safety relief valves to provide pressure protection for the reactor coolant system as a whole.

The discharge irradiation of Magnox fuel was determined by burn-up of the fissile U-235 component, and also by the potential for swelling or deformation of the uranium fuel.

Changes in the crystalline structure of the fuel under irradiation conditions lead to anisotropic changes in dimension: irradiation growth/creep. The effect is particularly important under conditions of low temperature and high stress, such as those found in the bottom of the fuel stack. Irradiation creep can be minimised by careful control of the crystalline structure of the fuel, which is influenced by the minor alloying components and by the manufacturing process.

As irradiation proceeds, fission gas builds up within the fuel. Diffusional processes result in microscopic bubbles forming within the fuel matrix, leading to swelling and possibly deformation of fuel elements. Swelling occurs primarily within a narrow band of temperatures centred around about 400 °C. This results in a typically annular region of porosity and swelling within a fuel element, which swells increasingly as irradiation (and so fission gas inventory) proceeds. The uranium metal was alloyed with small amounts of (principally) aluminium and iron to improve its resistance to swelling.

The results of irradiation growth and fuel swelling can lead to breaching of the fuel cladding. Bowing of fuel under irradiation also occurs and could lead to the fuel becoming hard to withdraw from the channel, given the typical clearance of a few millimetres from fin tip to channel wall. Bowing of the fuel was exacerbated by each fuel element supporting the weight of all of the elements above it.

Cladding ductility is also an issue for Magnox fuel. In general, a ductile clad is good as it will accommodate changes in fuel dimensions. Fine grain structures aid ductility, but are weaker at high temperatures. As a result of this two types of fuel element, known as HT and LT variants, were produced for many reactors. HT (or HTA) fuel was annealed at a high temperature giving a coarse-grained structure suitable for high-temperature operations. LT fuel was annealed at a lower temperature giving a fine-grain structure. As maximum fuel deformation, driven by dimensional changes in the uranium bar, occurs at relatively low operating temperatures, LT fuel gave better performance for elements at the bottom region of the fuel channel.

Considerable research was undertaken by the main UK operator of Magnox reactors, the Central Electricity Generating Board (CEGB) during its lifetime from 1957 to 1990, on optimisation of fuel irradiation conditions and of fuel element design. The resulting improvement in fuel performance led to channel average irradiations doubling from 3.6 GWd/Te(U) for early operations to eventually 7.2 GWd / Te(U) and to peak element irradiations of 9 GWd/Te(U). At the later levels irradiation was limited by burn-up of the fuel leading to loss of reactivity. As permissible irradiations increased, ‘double dwelling’ of fuel became standard practice, in which lower irradiation elements from top and bottom of the channel stack were returned for a second irradiation period

Although, the extraction of U and Pu from SNF has been developed on an industrial scale, processes that meet Generation IV targets for proliferation resistance have only been demonstrated under laboratory conditions. The technical and economic feasibility of deploying them on an industrial scale has not been established. The fabrication of fuels containing high quantities of MAs and possibly LLFPs will require heavily shielded facilities. The wastes from these new processes have not been studied and will require new treatment systems.

13.9 Sources of further information and advice

Further information on the Generation IV initiative may be found on the official website (http://www. gen-4.org/) and in many publications on these matters currently published by the main international nuclear journals.

From the beginning of commercial nuclear power plants, wet stores of limited capacity were planned and built at-reactor (AR) sites since the strategy for spent fuel management involved reprocessing, as described in the previous section. When reprocessing fell out of favour and disposal became a preferred option it became obvious that spent fuel storage capacity would need to be increased as well as storage duration. Dry storage technologies were soon developed and away-from-the-reactor (AFR) stores were built. At the beginning of the twenty — first century, spent fuel storage technology is a mature industry that can respond to the needs of nuclear operators on a commercial basis. During over more than 50 years of nuclear power plant operations, significant experience on spent fuel storage, either dry or wet, has been collected.

15.5.1 Wet storage of spent nuclear fuel

All UO2 based water reactor spent fuel is, after being removed from the reactor, stored in wet storage at the reactor (called the spent fuel storage pool or the irradiated fuel bay in Canada). There are also some centralized wet storage facilities where fuel is transferred for long-term storage. One example of such a centralized storage facility is CLAB in Sweden.

These wet storage systems aim to:

• Cool the spent fuel (heat removal system).

• Provide a biological shield for workers.

• Contain and remove contamination radioactivity that may be released from the fuel (all wet stores have a water purification system).

• Maintain the high clarity of the water to maximize visibility for remote handling of the fuel.

A spent fuel cooling and purification system uses demineralized water and typically contains pumps, filters, ion exchangers and heat exchangers. Typically, all components have a minimum of 100% redundancy, which is placed in parallel to enable the independent operation of each component. In most cases there would also be skimmers for removal of oils or other substances floating at the surface of the water.

Figure 15.9 shows a typical spent fuel storage pool. Pools are usually lined for water tightness either with stainless steel plates, or are coated with

water-resistant paint. At the bottom of the pools there are storage racks, the design of which is dependent on the fuel type, facility type and the facility operator.

After cooling the fuel for 5 to 10 years, it can be transferred to another wet store or to a dry storage facility (such as a centralized AFR store). The purpose of wet storage is the same as for AR spent fuel storage.

The fuel rod is the primary barrier for radionuclide containment purposes so that it is essential that fuel failures are reduced to the utmost. However, retention of the overall fuel assembly structure integrity is, if anything, even more important given that the fuel will at some point need to be retrieved. Consequently, the chemistry in spent nuclear pools must be carefully controlled to prevent corrosion of the fuel cladding and the structural elements of the fuel assembly.

There are variations in the chemistry of spent fuel pools and we will mention some key parameters that are typically controlled. The pH varies from the acid range (pH 4.5) to the basic range (5) depending on whether borated or demineralized water is used in the pools. For Magnox fuel, the pH is typically in the basic range (11.5-13) to prevent corrosion of the magnesium alloy by maintaining a magnesium hydroxide film on the cladding, which would dissolve in pure water. For stainless steel cladding, the pH is also maintained in the basic region. Although stainless steel is more susceptible to general corrosion than zirconium alloys, it is still low enough that is not expected to create problems during 100 years of storage. Conductivity is also a controlled parameter wherever possible. It has to be as low as possible under specific chemistry parameters (depending on the chemical additives; i. e. nuclear fuel from PWRs is stored with boric acid as a neutron poison. The acidity of boric acid affects the pH and conductivity that can be maintained). Chlorides, sulphates andfluorides are also maintained at a level as low as possible as they may trigger some specific types of corrosion. Sodium and calcium are also controlled in some pools.5 The level of radioactivity in pools is also monitored, primarily as a means of monitoring fuel failure and to indicate whether the purification system needs to be brought into action. Usually, the total radioactivity of the water is measured in addition to monitoring characteristic radionuclides like Cs-137 and Co-60.

One of the problems encountered in spent fuel storage pools comes from the growth of algae and bacteria. These can sometimes cause problems with the clarity of the water, which is essential for handling spent fuel during storage. Monitoring of water turbidity can provide an early indication of problems with bio fouling. Various chemicals are available that can remedy or eliminate problems of bio-logical activity in spent fuel pool water.

A number of potential degradation mechanisms have been investigated for fuel cladding under wet storage conditions: [27]

• Uniform (aqueous) corrosion — the corrosion of zirconium alloys in the spent fuel pool conditions is extremely slow.5

• Galvanic corrosion — zirconium alloys are near the noble end of the galvanic series and corrosion could occur through contact with Al. In this contact Al would be oxidized and Zr hydrided. Nevertheless, galvanic corrosion is prevented by the passive effect of the oxide layer on, which zirconium is generated during reactor operation (or even in some cases deliberately deployed as a thin layer). Galvanic corrosion due to contact between the Zr alloy and passivated stainless steel has not been observed.

• Pitting, and microbially induced corrosion — these are only possible if some specific conditions are present and that is avoided by the pool chemistry control.

• Hydriding — hydriding is to some extent prevented by the passive effect of the oxide layer on the zirconium. Hydrogen taken up by the cladding during reactor operation would under normal storage conditions precipitate as hydride platelets. Redistribution of those platelets could cause some loss of strength and trigger damage to the fuel element, but investigations so far indicate that this cannot occur under the conditions in wet storage.5

Another, safety concern in wet spent fuel storage is the reaction of Zr alloys with oxygen and steam considered for hypothetical accidents when the level of water in spent fuel pools decreases, leaving some of the surface of the spent fuel in contact with air. This would cause temperatures in the fuel assemblies to rise, accelerating the corrosion of the zirconium alloy cladding. The following chemical reactions with Zr can occur:

reaction in air Zr + O2 ^ ZrO2 reaction in steam Zr + 2H2O ^ ZrO2 + 2H2

Both reactions are strongly exothermic, which means they release large quantities of heat that can further raise cladding temperatures. These reactions can then become autocatalytic at high temperatures and explosive. Fortunately such autocatalytic reactions can happen only at temperatures that are very much higher than the temperature of boiling water (900-1000 °C).