Uranium is very active chemically and rapidly reacts with most environments (air, oxygen, hydrogen, water, water vapor, and others). Freshly polished uranium has a dull silvery color. However, when exposed to air for a few minutes, the surface shows straw-like color and darkens to a blue-black color within a few days. The oxide films formed are not quite protective. At elevated temperatures as the film thickens with time, the characteristic black color of UO2 develops, and it starts cracking and crum­bling exposing fresh uranium metal from underneath the oxide film to be attacked.

Unalloyed uranium reacts with water almost readily. Figure 7.4 shows the corrosion behavior of uranium in aerated distilled water. At 50-70 °C, an ini­tially formed UO2 film provides corrosion protection to the metal with consid­erable incubation period. However, beyond this temperature regime, corrosion rate picks up as the surface oxide film becomes porous and the protection of the oxide film is lost. Also, no incubation period is noted at these

temperatures. Conversely, in hydrogen-saturated or degassed water, the corro­sion rates remain linear with respect to time in the moderate temperature regime. It is postulated that hydrogen diffusion takes place through the thin oxide film to form uranium hydride (UH3) between the oxide and the metal and also into the grain boundaries of uranium.

This depends on the alloy composition (Cu, P, Ni, etc.), fluence, and irradiation temperature. Copper forms very fine (1-3 nm) coherent precipitates; nickel and manganese amplify the hardening effect of copper. Phosphorus segregates to the grain boundaries (contributes no more than 10-20% of the total effect) and the rest to the particle-matrix interfaces (this contribution is very important). Figure 6.32a and b shows the effect of copper content on the change of DBTT and upper shelf energy, respectively.

Recent advances in atom probe technique have positively identified the presence of these Cu — and P-enriched clusters. An atom probe map of a neutron-irradiated KS-01 weld is shown in Figure 6.33.

Thorium (atomic number 90) is a soft, silvery-white metal when present in pure form. It retains its luster for a long period of time. However, given a chance to oxidize to thorium oxide (ThO2), it quickly loses its luster turning into gray and

finally black in color. It is another nuclear fuel that has not been tapped to its full potential to date. As we have already discussed in Chapter 1, Th232 is a fertile iso­tope that could produce fissile U233 isotope upon capturing a neutron and then fis­sion to produce energy needed for electric power. The relevant reaction has been shown in Eq. (1.3). Thus, thorium is an important breeder material. Thorium exists in only one isotope form (Th232) in nature and decays very slowly (its half-life is

14.5 billion years, that is, thrice the age of the earth). Other isotopic forms of tho­rium (Th228, Th230, and Th234) occur as decay products of thorium and uranium at some point, but they are present in trace amounts only.

Thorium is far more abundant than uranium in nature. Most rocks and sands on earth surface contain minute amount of thorium. Monazite sand that is rare earth phosphate mineral is an important source of thorium, and almost two-third of this high-quality deposit (thorium content of 6-7 wt% on average) is found in southern and eastern coasts of India. Other thorium resource also includes thorite or tho­rium silicate (ThSiO4). A large vein deposit of thorium is present in the state of Idaho of the United States.

As noted before, radiation exposure generally leads to increased rates of corrosion and oxidation. Three basic effects on materials such as zircaloys are (i) oxida — tion/corrosion, (ii) hydriding, and (iii) stress corrosion cracking and corrosion fatigue.

Oxidation

In LWRs, radiolytic decomposition of water can lead to the creation of various free radicals (such as hydrogen, oxygen, hydroxyl radicals, and hydrogen peroxide), which may accelerate the corrosion effect. Some of the reactions occurring in water during radiolysis are given below:

Figure 6.42 shows the schematic representation of zircaloy corrosion in the tem­perature range of 260-400 °C. In zircaloys, ZrO2 forms as protective layer following a cubic rate law, meaning that oxidation rate decreases as ZrO2 layer thickens. However, after a certain point (breakaway transition), the oxidation rate (as meas­ured in terms of weight gain) becomes constant following a linear rate law, as shown in Figure 6.42. The breakaway transition occurs due to the destruction of the ZrO2 film. It is thought that polymorphic phase transformation of ZrO2 takes place as the stress develops in the ZrO2 film as the thickening continues. This poly­morphic phase transformation leads to sudden volume expansion, thus resulting in

cracking/flaking of the protective film (Figure 6.43). This also means that the lon­ger the breakaway time, the better would be the component sustaining the chemi­cal environment. Recently, research has shown that Nb-containing zirconium alloys (like Zirlo) has greater longer term corrosion resistance in LWR environment.

Generally, materials chosen to serve in nuclear reactors should have the ability to form self-healing, passivating oxide films. But prior experience has shown that it is quite difficult to maintain the passivating oxide film in the long term in a reactor environment. Various flaws can be introduced as a result of stress and irradiation, which may turn into crack initiation sites. Hence, both uniform corrosion and nod­ular corrosion have been observed to be an issue. Evidence of nodular corrosion on zircaloy fuel cladding tube is shown in Figure 6.44.

Plutonium-based ceramic compounds are mainly plutonium dioxide (PuO2), pluto­nium monocarbide (PuC), and plutonium nitride (PuN). But they are generally used with UO2, UC, and UN, respectively, in the form of mixed oxide fuels. These are considered fast breeder reactor fuels. Mixed oxide fuels have drawn the most interest because of the long experience with them. The physical, mechanical, and chemical properties of UO2 and PuO2 are quite comparable. The nuclear properties of ceramic plutonium fuels are considered better than ceramic uranium fuels. However, plutonium fuels must be diluted. Mixed ceramic uranium-plutonium fuels have certain benefits: (i) high melting temperatures, (ii) well-developed fuel fabrication technology and operation experience (obtained from UO2 fuel), and (iii) good thermal and irradiation stability. Given the lack of their near-term appli­cation as nuclear fuels, we will not discuss this further. Readers should refer to appropriate literature to gain information in this area.

7.3.5

Alloying of uranium is done to improve the mechanical properties, dimensional stability and corrosion resistance of uranium. However, selection of alloying ele­ments should not adversely affect the neutron economy; hence, a lot of emphasis was placed on the alloying elements like Al, Be, and Zr. The alloying elements like Ti, Zr, Nb, and Mo have extensive solid solubility in uranium at higher tempera­tures, V and Cr have moderate solubility, and Ta and W are further less soluble in C-U. Figure 7.5 shows the equilibrium-phase diagram of U-Mo system. U-Zr and U-Pu-Zr fuels in EBR-II were used as the alloying raised the alloy solidus

temperature, enhanced dimensional stability under irradiation, and reduced fuel­cladding material chemical interaction. Furthermore, uranium-fissium/fizzium (U-Fs or U-Fz) alloys are being utilized in LMFBRs. U-Fs alloys can be developed during the reprocessing of spent fuels in which part of the fission products such as Mo, Nb, Zr, Rh, Ru etc. are left in the uranium matrix. These types of alloys (e. g., U-15 wt% Pu-10 wt% Fs or U-5 wt% Fs) show better irradiation stability.

Addition of alloying elements to small concentrations in uranium can improve the high-temperature strength of the alloy. This is beneficial since the strength of uranium falls drastically at elevated temperatures. For example, addition of Cr to the tune of 0.5 wt% or Zr to 2.0 wt% can increase the yield strength by four-five times. Addition of Si and Al may also improve strength when added in small amounts. However, addition of larger amounts may result in the formation of brit­tle intermetallics, adversely affecting the ductility and fabricability of the alloy. Mar­tensitic transformation is another way of hardening the uranium alloys. The addition of Zr to the tune of about 5-10 wt% can be water quenched from the gamma-phase regime to produce supersaturated metastable alpha-phase (alpha — prime) regime. The as-quenched U-5 wt% Zr alloy (900 °C at 1 h and quenched) is very hard (~535 VHN). Upon tempering at 650 °C for 2h, it loses its hardness (~315 HVN). However, a range of microstructure can be developed by manipulat­ing the tempering parameters. Similar martensitic transformations also occur in U-Mo, U-Ti, and U-Nb alloy systems.

Uranium alloys exhibit better corrosion resistance by forming and retaining a protective oxide film up to 350 °C if the alloy is in the form of (i) metastable gamma — phase, (ii) supersaturated alpha-phase, and (iii) intermetallic compounds.

The first type of alloys contains 7 wt% or more Mo or Nb. The alloying elements remain dissolved in the gamma-matrix (BCC) by cooling at moderate or rapid rates from the gamma-phase regime. As long as the gamma phase is retained in the alloy, the corrosion rate remains low. It can be noted here that U-Mo alloys are being developed under the ‘Reduced Enrichment for Research and Test Reactors (RERTR)’ program. The RERTR program was started by the US Department of Energy in 1978 to develop technologies essential for enabling the conversion of civilian nuclear facilities using high enriched uranium (HEU; > 20 wt.% U235) to low enriched uranium (LEU; <20 wt.% U235). By the end of 2011, over 40 research reactors have been converted from HEU to LEU.

Supersaturated alpha-phase alloy is formed by adding a small amount of niobium (up to 3 wt%) and letting it cool rapidly leading to martensitic transformation. As long as the martensitic structure is maintained, the corrosion resistance property is retained. Further improvement in corrosion resistance can be achieved by adding zir­conium. For example, a ternary uranium alloy with 1.5 wt% Nb and 5 wt% Zr has good corrosion resistance. However, the alloy is susceptible to embrittling hydrogen attack.

Uranium-based intermetallic compounds may provide better corrosion resist­ance as typified by uranium silicide (U3Si). This class of uranium-based materials includes a range of intermetallics such as UAl2, UAl3, U6Ni, U6Fe, and so on. The main advantage of these compounds is that they provide corrosion resistance at elevated temperatures at which the first two types of alloys (metastable gamma and supersaturated alpha) cannot.

Higher irradiation temperature results in less embrittlement. This effect occurs due to the in-reactor annealing out of many radiation-produced defects before they can become stable defects.

According to Cottrell-Petch theory of radiation embrittlement [11],

where Ao; is the change in friction stress, oy is the yield stress, ky is the unpinning parameter, dky/dT is the change in the unpinning parameter as a function of tem­perature, and doy/dT is the yield stress as a function of temperature. Thus, the change in DBTT can be estimated by knowing these various parameters and their changes with temperature and fluence (Figure 6.34).

Thorium is mainly extracted from the monazite ore via a multistage process. The first stage is the process of digestion that involves dissolving monazite sands in concentrated sulfuric acid (93-98%) at 120-150 °C for several hours [11]. As an alternative, alkaline digestion process can also be followed. In the digestion pro­cess, thorium, uranium, and rare earth metals pass into solution to form sulfates

in phosphoric acid. Following the digestion process, the resulting solution is diluted to pH of 1 using ammonium hydroxide, and all the thorium from solution gets precipitated out of the solution along with some rare earths. But subsequently increasing the pH to ~2.5, the rest of the rare earth metals and uranium also get precipitated. The precipitate residue is then collected and treated with nitric acid (solvent extraction process) and thorium compound is separated.

There are several methods to obtain thorium from thorium compounds. Metallic thorium can be obtained by reduction in a sealed container or bomb-reacting thorium tetrachloride (ThCl4) or tetrafluoride (ThF4) with calcium, sodium, or mag­nesium. Because thorium has a high melting point, zinc is often added to create a low-melting eutectic from which Zn is later distilled off under vacuum to obtain the so-called “bomb-reduced” thorium. A relevant chemical reaction is shown below:

ThF4 + Zn + 2Ca! Th-Zn + 2CaF2 (7.12)

Highly pure thorium (>99.9%) can be obtained by using an iodide treatment (DeBoer process).

The morphology of the bomb reduced thorium is sponge-like and iodide thorium is loosely packed crystals of highly pure thorium. That is why they need to be consoli­dated through ingot or powder metallurgy process. In ingot metallurgy, two types of methods are generally employed: induction melting/casting under vacuum and arc melting/casting. If thorium is low in oxygen, silicon, nitrogen, and aluminum impurities, it can be fabricated by various deformation processing techniques like extrusion, hot and cold rolling, hot forging, and swaging. However, wire drawing presents challenges since thorium has great tendency to stick with the drawing dies. In the powder metallurgy, thorium can be fabricated into cold compacts (with 95% of theoretical density) from powders produced by hydride method. Then, hot pressing under vacuum at 650 °C at a nominal pressure can produce almost full density metal. The machining of thorium has been found to be easier, especially with greater tool feed rate and low spindle speeds.

This hydriding effect is particularly related to the use of zirconium alloy as fuel cladding. Hydrogen is produced due to reaction between water and zirconium in LWR environment. But moisture can also attack from within the fuel pellets if the pellets are not dried properly before insertion into the reactor. Hydrogen can then be absorbed and diffuse in the interior of zirconium cladding. The low-temperature a-zirconium (HCP) phase has a very low solubility of hydrogen, leading to any excess hydrogen-forming zirconium hydride precipitates. This leads to embrittle­ment, delayed hydride cracking, and hydride blistering, all of which diminish the lifetime of fuel rods and cause safety concerns in spent nuclear fuel rod reposito­ries. It has been known that crystallographic texture of the zirconium alloy fuel cladding tube can be tailored to minimize the effect of hydride formation. With a suitable texture, the hydrides form along the hoop direction in place of radial direc­tion, making it less susceptible to fracture. Figure 6.45 shows a microscopic cross section of hydrides oriented along the hoop direction in a zircaloy tube (the usual

texture consists of majority of grains with c-axis close to the radial direction: +/ — 30° from the radial direction toward the hoop).

Stress Corrosion Cracking

In Section 5.3, we have learned about corrosion basics, including stress corrosion cracking. Inside the reactor, these effects manifest themselves in various ways with respect to different components. Here, we introduce an important effect that occurs in the fuel rod known as pellet-cladding interaction (PCI). Figure 6.46a shows a schematic of the PCI effect. PCI is associated with local power ramps at start-up or during operation and occurs due to the effect of fission products like I, Cs, Cd, and so on, resulting in stress corrosion cracking. The crack generally initi­ates in the inner wall of the cladding and then progresses outward. Figure 6.46b shows an actual example of PCI failure in a fuel cladding system in a commercial power reactor. PCI minimization/elimination can be achieved by the following: (i) reduced ramp rates (flux variation or thermal gradient), but it is not an easy solu­tion, (ii) coated (barrier) fuel, that is, surface coated with proper lubricant, and (iii) barrier cladding, that is, inner surface coated with graphite, copper, or pure zirconium to minimize stresses at the ID surface of the cladding; in BWRs, crystal bar Zr and zircaloy-2 are coextruded to form a thin Zr liner on the ID. The latter is discussed in detail in the following.

A modified Zr-lined barrier cladding known as TRICLAD™ has been developed by GE by adding a thin layer of corrosion-resistant zircaloy-2 bonded to the inner surface of the Zr barrier [40, 41]. In addition, the outer surface is made resistant to nodular corrosion by heat treatment that results in small second-phase particles, while major­ity of the parent zircaloy-2 material contained characteristic large SPP size distribu­tion for improved crack growth resistance (toughness) [42, 43]. Figure 6.47 is a schematic of the Triclad for BWR service with the following four layers from ID to OD: (i) an inner layer of corrosion-resistant zircaloy-2 to slow oxidation and hydrogen generation, and to delay local hydride formation in the case of rod perforation, (ii) a Zr barrier for PCI resistance to blunt cracks nucleated at the inner surface, (iii) bulk

Corrosion-Resistant Outer Zircaloy-2 Surface

parent zircaloy-2 tubing with good toughness during irradiation, and (iv) an outer layer of zircaloy-2 processed for high resistance to nodular corrosion. An important feature of the final product is the integrity of the metallurgical bonding between the Zr barriers with both the inner zircaloy-2 liner and the bulk zircaloy-2 tubing.

Thorium dioxide (ThO2) is undoubtedly the best-characterized ceramic com­pound of thorium. Although this partly stems from its study for nuclear pur­poses, the majority of information exists because of the non-nuclear

usefulness of the material. Since thoria has the highest melting point (~3300 °C) and is the most stable to reduction of all the refractory oxides, it is a superior crucible material for the melting of reactive metals. Thoria is gener­ally prepared in powder form by the thermal decomposition of a purified salt, generally the oxalate. This powder can be consolidated by usual ceramic fabri­cation techniques, such as slipcasting, pressing, and sintering, or hot pressing. The fabricability and ceramic properties can often be related to conditions of preparation of the starting salt and firing.

Thorium dioxide exists up to its melting point as a single cubic phase with the fluorite (CaF2 type) crystal structure and is isomorphous and completely miscible with UO2 to a measurable extent. Therefore, it is stable to high temperatures in oxidizing environments. In vacuum, it darkens with loss of oxygen, even though the loss is insufficient to be reflected in chemical analysis or lattice constant mea­surement. Unlike UO2, thoria does not dissolve oxygen even on prolonged heating to 1800-1900 °C. By reheating in air to 1200 or 1300 °C, the white color can be brought back.

When uranium dioxide is incorporated in thoria, the lattice can take up extra oxy­gen in proportion to the uranium content. Table 7.6 summarizes some important physical and mechanical properties of thoria, along with analogous properties of uranium dioxide taken from the compilation by Belle.

Thorium carbide and thorium mononitride fuels may have potential for use as nuclear fuels, but have not been thoroughly studied.

Weld Fuel Rod

‘ Fuel Rod ‘ Characterization Acceptable? .

Yes

Figure 7.25 Process flowchart for fabrication of metallic fuels for fast reactors. Burkes et al., Ref. [19].

7.4

Summary

The topic of nuclear fuels is vast and is not easy to cover in a single chapter. None­theless, here a succinct review of both metallic and ceramic nuclear fuels is made and their various properties are discussed. Among metallic fuels, uranium, pluto­nium, and thorium are discussed. Among ceramic fuels, uranium dioxide, ura­nium nitride, and uranium carbide as well as plutonium-based oxide fuels and thorium oxide are covered. The metallic and ceramic fuels are found to have both advantages and disadvantages of their own.

Problems

7.1 What are the advantages and disadvantages of metallic nuclear fuels?

7.2 Describe various requirements imposed on a nuclear reactor fuel.

7.3 How many allotropic forms does uranium have? Discuss the effect of allo — tropic transformation on the properties of uranium.

7.4 Describe the plastic deformation mechanisms of alpha-uranium.

7.5 Discuss the beneficial effects of alloying on the properties of uranium as a nuclear fuel.

7.6 Distinguish between thermal cycling growth and radiation growth of alpha — uranium.

7.7 Martensitic transformation may occur in uranium system. Describe one example and its advantages and disadvantages.

7.8 How is plutonium produced?

7.9 Why is plutonium said to have fickle nature?

7.10 Discuss the origin of self-irradiation behavior of plutonium and its effects.

7.11 Discuss the origin of superplasticity in plutonium.

7.12 Can plutonium be alloyed with other metals? If so, compare it with uranium — based systems.

7.13 What are the main mineral sources ofthorium?

7.14 What are the main impediments to realizing thorium cycle on a wider scale?

7.15 State the advantages and disadvantages of ceramic nuclear fuels.

7.16 Describe the effects ofhypostoichiometry, stoichiometry, and hyperstoichiom­etry on the properties ofuranium dioxide.

7.17 Compare and contrast between UO2, UN, and UC fuels.

Bibliography

instability. Metallurgical and Materials Transactions A, 35 (8), 2207-2222.

9 Merz, M. D. and Nelson, R. D. (1970) Proceedings of the 4th International Conference on Plutonium and Other Actinides 1970 (ed. W. N. Miner), The Metallurgical Society of AIME, New York, p. 387.

10 Gschneider, K. A., Jr., Elliott, R. O., and Waber, J. T. (1963) Acta Metallurgica, 11 (8), 947-955.

11 Ray, H. S., Sridhar, R., and Abraham, K. P. (1985) Extraction ofNonferrousMetals, East-West Press, New Delhi, India.

12 IAEA (1997) Thermophysical Properties ofMaterials for Water Cooled Reactors/IAEA-TECDOC — 949, IAEA, Vienna, Austria.

13 Peterson, S., Adams, R. E., and Douglas, D. A., Jr. (1965)

Properties ofThorium, Its Alloys and Its Compounds, ORNL Report ORNL-TM — 1144.

14 World Nuclear Association, http://www. world-nuclear. org.

15 Webb, J. A. and Charit, I. (2012) Analytical determination of thermal conductivity of W-UO2 and W-UN cermet nuclear fuels, Journal of Nuclear Materials, 427, 87-94.

16 Webb, J. A. (2012) Ph. D. Analysis and Fabrication of Tungsten CERMET materials for Ultra-High Temperature Reactor Applications via Pulsed Electric Current Sintering, University ofIdaho.

17 Rondinella, V. and Wiss, T. (2010) The high burn-up structure in nuclear fuel. Materials Today, 13, 24-32.

18 Hayes, S. and Peddicord, T. (1990) Material properties correlations for uranium mononitride IV. Journal of Nuclear Materials, 171, 300-318.

19 Burkes, D. E., Fielding, R. S., Porter, D. L., Crawford, D. C., and Meyer, M. K. (2009) A US perspective on fast reactor fuel fabrication technology and experience. Part I: metal fuels and assembly design. Journal of Nuclear Materials, 389, 458^89.

Additional Reading

Allen, T., Busby, J., Meyer, M., and Petti, D. (2010) Materials challengesfor nuclear systems. Materials Today, 13 (12), 14—23.

Buckley, S. N. (1961) Irradiation growth, Atomic Energy Research Establishment, Harwel ARE-R 3674, UK.

Appendix A