Uranium Dioxide

Uranium Dioxide (UO2) has more design heritage than Uranium Nitride (N2N3).

From: Functionality, Advancements and Industrial Applications of Heat Pipes, 2020

Chapters and Articles

Oxide power reactor fuels

Lian C. Wang, Matthew H. Kaye, in Advances in Nuclear Fuel Chemistry, 2020

Abstract

Uranium dioxide (urania) is widely used in the nuclear industry for various kinds (e.g., pressurized water reactors, boiling water reactors, and Canadian Deuterium Uranium) of nuclear power reactors. Currently urania is used in over 90% of power reactors, but other oxides, in conjunction with the traditional urania fuel matrix, show promise as potential alternatives to a pure UO2 reactor core. For example, mixed oxide fuels are of interest since they incorporate plutonium into the fuel matrix and increase uranium utilization. Because thorium is three times more abundant than uranium, thorium-based nuclear fuels show promise despite thorium being a fertile not fissile material. To adapt these new nuclear fuel sources to future generations of reactors, dopants, such as chromia (Cr2O3), alumina (Al2O3), titania (TiO2), niobia (Nb2O5), vanadia (V2O5), are used, or are being or have been considered. This chapter serves as an introduction to the basic physical, chemical, and thermodynamic properties of UO2-, PuO2-, and ThO2-related oxides. The latest phase diagram for systems incorporating these oxides is presented.

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Advances in fuel fabrication

Elizabeth Sooby Wood, ... Paul Demkowicz, in Advances in Nuclear Fuel Chemistry, 2020

10.2.1.2 Additives to UO2

Uranium dioxide has been deployed in all commercial LWRs over the past 50 years and is therefore a likely uranium compound to consider for advanced reactor concepts, given its documented success in the nuclear industry. Incremental progress to this commercial fuel form has been in the form of minor alloying additions/dopants to improve the fission gas retention as well as creep behavior at operating temperatures to minimize detrimental pellet–cladding interactions (PCI) [8]. An exhaustive list of dopants has been examined in the literature, including niobium [9,10], titanium [11], silicon [12], aluminum [12,13], and chromium [14,15]. Each investigation studies the impact of these dopants on the microstructure as well on the operating parameters of the fuel, in particular grain size, thermal conductivity, ionic conductivity, mechanical properties, etc.

Industry teams have each put forth concepts for fuel additions that attempt to improve upon the aforementioned properties and, at a minimum, do not greatly degrade the performance of UO2 in-pile with respect to other performance metrics (e.g., oxidation behavior and washout). To this end, Cr2O3 was first proposed, which enhances grain growth during sintering of the fuel pellets when added near the solubility limit of Cr2O3 in UO2 of 0.16 wt.% [14,16]. Chromia additions above 0.16 wt.% rely on higher temperatures during sintering, which reduces Cr cations to a +2 valency state and is suspected to form liquid phases to increase grain growth [17,18].

Chromium additions increase the grain size of the native UO2 from ~10 µm to 50–60 µm with pellet densities in excess of 96% [17]. Enlarged grains increase the diffusion length of fission product gasses, thus improving the retention of these gases in the fuel as well as slight improvements to the sintered density, benefitting the core economics. However, irradiation tests on Cr2O3-doped UO2 have also shown increased diffusion of fission gases (e.g., 133Xe) from the enhancement of defect cites of the UO2 lattice leading to similar or even accelerated gas release [12,15]. It is noted that the referenced two studies were conducted at temperatures greater than standard reactor operating conditions, which is anticipated to give rise to higher defect concentrations and thus higher fission gas diffusion [18]. Current studies are evaluating if dopants provide a beneficial impact on fission gas retention at standard operating conditions. Given that Cr2O3 is a known neutron absorber, teams have also investigated other cations that can substitute Cr3+ cations while also contributing to grain growth and minimizing the Cr3+ neutronic penalty within the fuel. Al2O3 additions with <1000 ppm by weight Cr2O3 are added to UO2, which yields densities >97% with slightly smaller grain sizes of 40–55 µm [19]. Note that grain growth can be facilitated by the higher Cr2O3-doped UO2 as reported by [17] and is displayed in Fig. 10.2.

Figure 10.2. Optical micrographs of Cr2O3-doped UO2 sintered under the same conditions where the Cr2O3 concentration is (A) 0.1 wt.% and (B) 0.25 wt.% [17]. The provided scale bar is representative for both figures.

For boiling water reactors (BWRs) the mechanical behavior of the fuel is a key parameter since higher failure rates are attributed to PCI during burnup in comparison to pressurized water reactors. It should be noted that the primary design feature to mitigate PCI in BWR fuel is modified, liner cladding. Nonetheless, because of the known susceptibility to PCI, additives to UO2 are designed to improve creep resistance of the fuel during operation. Promising additives generally include nonsoluble aluminosilicates that coat the UO2 grain boundaries, allowing viscous flow between the grains at operating temperatures [13]. Grain growth and density is also improved with the viscous grain boundary phases, which promotes liquid-phase sintering, yielding densities ~95%–97%, and grain sizes ~30–60 µm for additions of Ai–Si–O of 2500 wppm [13,20,21].

Commercially, the additives are introduced to UO2 by solid-state synthesis, whereby the as-fabricated feedstock UO2 powder from the standard powder line is comixed with oxide additives to distribute the dopants. The combined powders are then pressed and sintered under a reducing atmosphere to yield sufficiently high density to meet fuel pellet design requirements.

Recent modeling efforts have suggested that other unexplored dopants to UO2 could yield improved grain growth and densification [22]. In particular, elements such as Fe3+ and Mn3+ at low concentrations could increase the defect concentration yielding improved sintering kinetics similar to the aforementioned cations. However, these concepts will require considerable experimental effort for acceptance in commercial reactors, including but not limited to fresh fuel property characterization, irradiation testing, and postirradiation examinations to fully realize opportunities with these new dopants.

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A REVIEW OF ACCIDENT RISKS IN LIGHT-WATER-COOLED NUCLEAR POWER PLANTS

Lynn E. Weaver, in Nuclear Power Safety, 1976

LOCATION AND MAGNITUDE OF RADIOACTIVITY

The fresh uranium dioxide pellets that serve as the fuel in the PWR and BWR reactors are only slightly radioactive. However, during reactor operation the fission process produces large amounts of radioactivity in the fuel. By far the largest fraction of the radioactivity is associated with the fission products resulting from the fission process. Some of the neutrons produced by the fission process are absorbed, to various degrees, by structural and coolant materials and thereby generate radioactivity. This radioactivity is generally referred to as induced radioactivity. The induced radioactivity is only a minute fraction of the total radioactivity that could potentially be released from the reactor in the event of a severe accident and is, therefore, not important.

While essentially all the radioactivity in the plant is initially created in the reactor core, transfer of spent fuel assemblies from the core results in considerable radioactivity being located in other parts of the plant. The radioactivity inventory, second largest in amount compared to the reactor core, is located in the spent fuel storage pool (SFSP) which holds fuel that has been removed from the reactor and is awaiting shipment to an off site fuel reprocessing facility. The average number of fuel assemblies in the SFSP constitutes about half of a full reactor core loading. Radioactive fuel assemblies in the plant may also be located in the spent fuel shipping cask which holds up to about 10 fuel assemblies. The refueling transfer from the core to the SFSP involves only a single fuel assembly at a time. In addition to the above, smaller sources of radioactivity are normally present at the plant in the waste gas storage tanks (WGST) and the liquid waste storage tanks (LWST). These latter sources result, for example, from leakage of a small amount of radioactivity from the fuel rods during reactor operation, as well as radioactivity induced in impurities in the reactor cooling water. Typical magnitudes of the radioactive inventories in the above noted plant locations are shown in Table III.

Table III. Typical Radioactivity Inventory for a 1000 MWe Nuclear Power Reactor

Empty CellTotal Inventory (Curies)Fraction of Core Inventory
LocationFuelGapTotalFuelGapTotal
Core(a)8.0 × 1091.4 × 1088.1 × 1099.8 × 10−11.8 × 10−21
Spent Fuel Storage Pool (Max.)(b)1.3 × 1091.3 × 1071.3 × 1091.6 × 10−11.6 × 10−31.6 × 10−1
Spent Fuel Storage Pool (Avg.)(c)3.6 × 1083.8 × 1063.6 × 1084.5 × 10−24.8 × 10−44.5 × 10−2
Shipping Cask(d)2.2 × 1073.1 × 1052.2 × 1072.7 × 10−33.8 × 10−52.7 × 10−3
Refueling(e)2.2 × 1072 × 1052.2 × 1072.7 × 10−32.5 × 10−52.7 × 10−3
Waste Gas Storage Tank--9.3 × 104--1.2 × 10−5
Liquid Waste Storage Tank--9.5 × 101--1.2 × 10−8
(a)
Core inventory based on activity 1/2 hour after shutdown.
(b)
Inventory of 2/3 core loading; 1/3 core with three day decay and 1/3 core with 150 day decay.
(c)
Inventory of 1/2 core loading; 1/6 core with 150 day decay and 1/3 core with 60 day decay.
(d)
Inventory based on 7 PWR or 17 BWR fuel assemblies with 150 day decay.
(e)
Inventory for one fuel assembly with three day decay.

The values given in Table III are typical for a 1,000 megawatt electric (MWe) plant operating at 3,200 megawatts thermal (MWt). In addition to the reactor power level, the plant radioactive inventory depends slightly on the length of power operation. For example, in the reactor core the total amount of radioactivity produced is directly related to the product of the power level and time at power. However, since the radioactivity decays to other isotopes, which are non-radioactive or less radioactive, an equilibrium amount of radioactivity occurs when the radioactive decay rate equals the production rate. For most of the radioactivity, equilibrium has occurred after several months of sustained operation. The reactor core radioactivity inventory shown in Table III is based on 550 days of sustained operation and represents the expected equilibrium radioactivity in an operating reactor. The inventory of radioactivity in the SFSP is based on a plant that has a common SFSP serving two 1,000 MWe reactors. The average number of spent fuel assemblies stored in the SFSP is based on assumed normal unloading and shipment schedules. The radioactive inventory in the shipping cask is based on a full load of fuel in the largest shipping cask currently licensed, and the shortest decay period (150 days) allowed for fuel shipped in the container. The refueling radioactivity represents that amount in a single fuel assembly at three days after reactor shutdown. This time is typical of the earliest time after shutdown that transfer of fuel from the reactor core to the SFSP begins.

Table III clearly shows that the reactor core contains by far the largest source of radioactivity in the plant. It also shows there is a relatively large inventory of radioactivity in the fuel in the SFSP and indicates that potential accidents of interest could result from melting of fuel initiated by a complete loss of water from the pool. While the spent fuel assemblies in a loaded shipping cask constitute a significant radioactive inventory, there is only a small potential for releasing a small fraction of this radioactivity in an in-plant accident. The radioactivity in shipping cask fuel has decayed long enough so that air cooling alone is sufficient to preclude fuel melting. However, the fuel clad temperatures reached may become high enough to cause cladding failures and the release of the small amount of gaseous radioactivity that collects in the fuel rod gap and plenum. The postulated accident related to refueling transfer is the inadvertent lifting of a fuel assembly completely out of the water-filled refueling canal or SFSP. Convective air cooling and heat radiation are also adequate to prevent fuel melting in this case, but cladding failures and a relatively small release of radioactivity (from the fuel rod gap and plenum) could result. The radioactivity in the waste gas storage tanks (WGST) and liquid waste storage tanks (LWST) are very small compared to the other sources. Accidents postulated for release of radioactivity from these tanks include tank ruptures as well as malfunctions that could involve release of the contents of the tank.

Although accidents that involve release of radioactivity from the shipping cask fuel, the refueling process, the WGST and the LWST would be troublesome, particularly to in-plant personnel, none of these could result in public consequences nearly as serious as accidents involving melting of the fuel in the reactor core or in the SFSP. Thus, although the study treats accidents involving all the radioactive sources listed in Table III, the ensuing discussion is directed at potential accidents involving fuel in the reactor core and the SFSP.

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Material Properties/Oxide Fuels for Light Water Reactors and Fast Neutron Reactors

D.D. Baron, L. Hallstadius, in Comprehensive Nuclear Materials, 2012

2.19.6 Conclusions

The performance of uranium dioxide or MOX fuels in LWR nuclear reactors is very well established. These fuels have demonstrated a very good behavior during irradiation, favored by their high melting temperature, giving large operating temperature margins. Some progress is still needed in the characterization of the mechanical properties of irradiated fuel samples. Experimentally, this is a tricky problem to solve, accounting for the drastic fuel pellet fragmentation and the steep radial gradient in the nuclear fuel properties evolution. Microindentation and focused acoustic techniques, able to characterize micrometric-sized fields on irradiated fuel samples, are now potentially available in the ITU hot laboratories.

Due to the mild operating conditions compared to FBR reactors, the uranium dioxide matrix retains most of the fission products, even at very high burnup, up to 300 MWd kgM−1.48 The local fuel power variability is higher in BWR than in PWR, due to an axial gradient in the neutron spectrum and control blade operation. For these reasons, BWR fuel has evolved with the use of a PCI liner barrier. The addition of minor quantities of softening elements to the fuel (chromium, niobium, alumino-silicates) is expected to increase PCI margins in both BWR and PWR fuel. Moreover, these additives favor manufacturing large grain fuels in order to delay the release of corrosive volatile fission products during power transient. The new and improved zirconium alloys first introduced in the 1980s allowed achieving higher discharge burnups due to a reduction of the cladding oxidation and hydrogen pickup.

As the use of uranium dioxide is universal, very efficient international projects such as NFIR, SCIP, HBEP, HBRP, and HRP, etc., have provided the modelers in the past with useful basic data or brought good insights for mechanisms' comprehension. With the incredible progress in computer capability and the large improvement in the knowledge base regarding the mechanisms involved in the fuel during irradiation, very efficient tools are now available for fuel design. Modeling is still in progress, shifting progressively to multiscale approaches, or fundamental atomic calculations to evaluate parameters not attainable, up to now, by experimental characterization (see Chapter 1.08, Ab Initio Electronic StructureCalculations for Nuclear Materials; Chapter 1.10, Interatomic Potential Development; Chapter 1.11, Primary Radiation Damage Formation; Chapter 1.12, Atomic-Level Level DislocationDynamics in Irradiated Metals; Chapter 1.13, Radiation Damage Theory; Chapter 1.14, Kinetic Monte Carlo Simulations of Irradiation Effects; Chapter 1.15, Phase Field Methods; Chapter 1.16, Dislocation Dynamics; Chapter 1.17, Computational Thermodynamics: Application toNuclear Materials; and Chapter 1.18, Radiation-Induced Segregation).

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Reaction kinetics and chemical thermodynamics of nuclear materials

Anna L. Smith, Rudy J.M. Konings, in Advances in Nuclear Fuel Chemistry, 2020

The vapor phase above condensed uranium dioxide is rather complex and constituted of several species, namely, O(g), O2(g), U(g), UO(g), UO2(g), and UO3(g). The predominant species depends on the uranium dioxide stoichiometry, that is, UO(g) for an hypostoichiometric sample (O/U<~1.95), UO2(g) for a sample close to stoichiometry, and UO3(g) for a hyperstoichiometric composition. Regardless of the starting composition, the uranium dioxide sample will tend to its azeotropic composition at equilibrium.

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FeCrAl—iron–chromium–aluminum monolithic alloys

Raul B. Rebak, in Accident Tolerant Materials for Light Water Reactor Fuels, 2020

Resistance to crud deposition under normal operation conditions

In the zirconium and uranium dioxide fuel rod system, it may be likely that a scale buildup would happen on the wall of the cladding in contact with the coolant owing to the presence of iron, chromium, and nickel ions in the water. The buildup on the OD of the cladding could be linked to the heat transfer across the cladding wall (Olander, 2009). Crud deposition is undesired because it decreases the heat removal rate from the rods by the coolant. The crud deposition in power plants has been curtailed in the last couple of decades by the strict control of the chemical composition of the coolant. At this moment it is not known how the magnetic ferritic alloys APMT and C26M would behave regarding crud deposition from the coolant side. Testing is necessary to determine whether FeCrAl materials would be more or less prone to crud deposition than the current zirconium alloys. In reactors with zirconium alloy fuel rods, the main dissolved elements in the water that later contribute to crud deposition are nickel and iron. If the cladding material is changed to FeCrAl, besides nickel and iron, it is expected that the water may also contain dissolved molybdenum and aluminum. At this moment, it is not known how these newer elements dissolved in the coolant may impact the environmental performance of the metallic components in the reactor.

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Fuel-Rod Design

R.H.S. WINTERTON, in Thermal Design of Nuclear Reactors, 1981

Effect of porosity on UO2 thermal conductivity

The main fuel material is uranium dioxide, UO2. The starting-point for uranium dioxide manufacture is uranium hexafluoride since both the diffusion and centrifuge enrichment processes require gaseous uranium hexafluoride and rely on the very slight difference in molecular weight between U235F6 and U238F6. After a series of reactions with different gases UO2 is left in the form of a fine powder. The powder is sintered at about 1700°C to make the fuel pellets which inevitably have a residual porosity. A porosity of around 5% is used for reactor work, that is, about 5% of the pellet volume is occupied by spaces. To produce 100% dense material by melting the UO2 would be difficult since the melting-point is 2865°C [2,3]. Also the fuel tends to swell under irradiation and a small amount of porosity is helpful in accommodating this swelling. The extremely high melting point of this ceramic fuel is the reason that it can be used, because, as was seen in the example above, the thermal conductivity is very low.

The presence of these voids in the fuel naturally reduces the thermal conductivity. For the idealised case of porosity P that consists of isolated spheres of zero conductivity in a matrix of conductivity k0, the expression derived by Maxwell [4] reduces to

(3.10)k=k01P1+0.5P.

In practice the voids cannot be taken to be isolated and spherical, and they may well vary in shape from one sample to another. To allow for this various empirical expressions have been suggested, such as

(3.11)k=K01P1+αP

and

(3.12)k=k0(1βP)

where the values of the geometrical factors α and β have to be determined experimentally. Concentrating on equation (3.12) which for small values of P is essentially the same as equation (3.11) anyway, and comparing it with equation (3.10), the value of β for isolated spherical pores is seen to be 1.5. Experimental values of β show a wide scatter but average to around 2.5 [5,6]. ∫kdT values can also be corrected using equation (3.12).

A number of more sophisticated methods of allowing for the effect of porosity have been proposed, including theoretical ones that require an accurate knowledge of the shape of the pores and whether or not they are interconnected [7], and empirical ones based on experimental results that have a temperature dependent porosity correction [8].

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Thermal aspects of conventional and alternative nuclear fuels⊛

Wargha Peiman, ... Mohammad Hosseiny, in Handbook of Generation IV Nuclear Reactors (Second Edition), 2023

18.3.1.1 UO2

As a ceramic fuel, Uranium diOxide (UO2) is a hard and brittle material due to its ionic or covalent interatomic bonding. In spite of that, UO2 is currently used in PWRs, BWRs, CANDU reactors, and other nuclear reactors due to its favorable properties. Oxygen has a very low thermal-neutron absorption cross-section, which does not result in a serious loss of neutrons. UO2 is chemically stable and does not react with water within the operating temperatures of these reactors. UO2 is structurally very stable such that the crystal structure of the UO2 fuel retains most of the fission products even at high burn-ups (Cochran and Tsoulfanidis, 1999).

The thermal conductivity of the fuel is an important thermophysical property in the computation of the fuel temperature. The thermal conductivity of 95% theoretical density UO2 can be calculated using the Frank correlation, shown as Eq. (18.1) (Carbajo et al., 2001). In Eq. (18.1), T is the temperature in K. This correlation is valid for temperatures in the range of 25–2847°C. Even though UO2 has a high melting point, its thermal conductivity is very low compared to those of high thermal-conductivity or composite fuels. The properties of other fuels are discussed in the following sections.

(18.1)kUO2T=1007.5408+17.692×103T+3.6142×103T2+6400103T5/2exp16.35/103T

The thermal conductivity of the fuel varies with temperature and is affected by manufacturing methods, the percentage of the porosity of the fuel, burn-up, fission gas release and deviation from stoichiometry. As such, there are uncertainties in the reported thermal conductivities. For UO2, the uncertainty is about 10% for temperatures below 1727°C (2000 K), while the uncertainty increases up to 20% for temperatures between 1727°C (2000 K) and 2847°C (3120 K) (IAEA, 2006). Figure 18.2 shows the thermal-conductivity profiles of UO2 as a function of fuel temperature for various percentages of theoretical fuel density, manufacturing, stoichiometry and irradiation. Figure 18.3 shows the impact of porosity and irradiation on thermal conductivity of UO2. Thermal conductivity is shown for un-irradiated UO2 and irradiated UO2 with a neutron flux of 1.16 × 1019 neutrons/cm2 at 527°C (800 K) before testing. In addition, Figure 18.4 shows the uncertainty associated with the thermal conductivity of UO2 for various percentages of fuel porosity. In these figures, the theoretical density of UO2 is considered to be 10,960 kg/m3.

Figure 18.2

Figure 18.2. Thermal conductivity of UO2 as a function of percentage of theoretical density, manufacturing, stoichiometry and irradiation

Figure 18.3

Figure 18.3. Impact of porosity and irradiation of thermal conductivity of UO2

Figure 18.4

Figure 18.4. Uncertainty in thermal conductivity of UO2

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Current headend technologies and future developments in the reprocessing of spent nuclear fuels

Chris J. Maher, in Reprocessing and Recycling of Spent Nuclear Fuel, 2015

5.3.2.2 Fuel pretreatment using high- and low-temperature processes

Oxides

High-temperature pretreatments that oxidize uranium dioxide to higher oxides have been under development for decades but not deployed in large-scale industrial reprocessing plants. These processes have been developed with different end purposes in mind:

Volox—voloxidation, a method of removing volatile fission products from fuel to allow abatement and simplify the dissolution cycle (Goode, 1973).

OREOX—oxidation reduction oxidation process is a development of the Volox process that uses alternate oxidation reduction processes to achieve higher efficiency of fission product removal (Strausberg et al., 1960).

AIROX—atomics international reduction oxidation process is an OREOX process to allow pulverization of fuel in fuel rods with holes punched in them. This allows fuel and cladding separation before dissolution (Majumdar et al., 1992).

DUPIC—direct use of PWR fuel in CANDU process is not strictly a headend to reprocessing; it is a pyrochemical method for fuel reprocessing. The concept allows the removal of some fission products prior to refabrication into fuel (Lee et al., 2012).

These processes essentially use the same chemistry, the oxidation of uranium dioxide to higher oxides, Equation 5.10, with the accompanying decrease in density (~ 11-8.4 g cm 3), thus allowing fuel pulverization. During this process, depending upon the process conditions, various amounts of tritium, carbon-14, iodine-129, kryption-85, and semivolatiles, including caesium and ruthenium, are evolved. The fuel expansion process can also be used to break open cladding that has been punctured (AIROX process), thereby avoiding the need for a shear process, or the process could be carried out upon sheared fuel.

(5.10)3UO2+O2viaU3O7,U4O9U3O8

From experiments with unirradiated and irradiated materials, it is known that the presence of elements that cannot be oxidized in the solid state affects the process chemistry. Examples include the presence of plutonium (Rance and Beznosynk, 2005), thorium (Anderson et al., 1954), MOX or alkaline and rare earth elements (SIMFUEL or irradiated fuels) (Kang et al., 2007) with uranium. These reduce the material reactivity as there is a smaller percentage of uranium to provide the driving force for pulverization and release of fission products. The presence of high contents of these dopant elements can stabilize intermediate phases, increase temperature or oxidant conditions to achieve oxidation, and ultimately at high contents prevent oxidation (Thomas et al., 1993; McEachern et al., 1998).

Experiments oxidizing SIMFUEL with varying portions of simulated fission products show that the presence of > 7% fission products (equivalent to ~ 70 GWd tHM 1) greatly stabilizes the (U,Fp)4O9 phase1 at 400 °C (Cobos et al., 1998). To achieve oxidation, an increase in temperature to 800 °C is needed.

During the oxidation of mixed uranium plutonium oxide, the material oxidizes to a plutonium-rich dioxide phase and a uranium-rich higher oxide phase, Equation 5.11 (Rance and Beznosynk, 2005). During this type of oxidation test, the solubility of plutonium in the M3O8 phase has been observed to be ~ 3% Pu HM 1, which means that oxidation of most MOX will result in the formation of plutonium-rich oxide particulates. This reduces the proportion of the plutonium dissolved in nitric acid (Cadieux and Stone, 1980). However, at higher plutonium contents oxidation does not occur and experiments up to 900 °C have shown that unirradiated 20-25% Pu HM 1 MOX does not result in pulverization. Similar segregation is also observed for the oxidation of mixed uranium thorium oxide and, as thorium oxide does not dissolve readily in nitric acid (Anderson et al., 1954; Goode and Stacy, 1979), results in an increase in difficult to dissolve thorium oxide residues.

(5.11)3UPuO2+xO21xMO2+xM3O8

The oxidation of spent thermal and fast reactor fuels have been studied at the laboratory scale. These experiments have studied the effect of temperature, feed flow rate, and oxygen concentration upon the extent of fission product removal, for example (Goode, 1973). This type of experiment has shown that near-complete tritium removal is possible at moderate temperatures (480 °C) (Goode et al., 1980). Less complete removal of the noble gases and iodine-129 has been achieved (see Table 5.5).

Table 5.5. Effect of voloxidation conditions upon noble gas and iodine-129 removal (Goode, 1973)

T (°C)[O2] (%)Gas flow (cfh)Volatilised (%)
85Kr-133Xe131I
450750.219~ 8
750751.069~ 99
450250.273~ 85
750751.042~ 90
450751.036~ 33
750750.276~ 95
450251.011~ 38
750250.2~ 99~ 92

Oxidation of fuel at higher temperatures, for example, greater than 700 °C, increases the amount of the so-called semivolatile elements that are released (Goode, 1973). Examples of fission products that are evolved in significant quantities are caesium-137, ruthenium-106, antimony-125, and niobium-105. Experiments underpinning the DUPIC cycle development show that oxidation at higher temperatures quantitatively increases the release and number of semivolatile elements (Bateman et al., 2006). These experiments also observe that control of the oxidation conditions can suppress the distillation of caesium and other elements by formation of less-volatile oxides. Oxidation of spent fuel and distillation of fission product elements under high-temperature conditions can be compared with studies of fuel distillation under severe reactor failure, such as Kundsen cell mass spectrometry studies (Capone et al., 1996).

The oxidation of irradiated fast reactor 20% Pu HM 1 MOX has also been demonstrated (Goode, 1973). These experiments have shown that despite the difficulties in the oxidation of unirradiated 20% Pu HM 1 MOX (Rance and Beznosynk, 2005), the oxidation of irradiated MOX is achievable. Removal efficiencies of noble gases of up to 98% at 750 °C has been demonstrated, which is excellent considering a portion of the noble gases are likely to be trapped in plutonium-rich oxides and not released. Comparing these results with irradiated uranium oxide experiments suggests good removal efficiencies of tritium, carbon-14, and iodine-129 will also be possible. This type of test clearly shows oxidation of 20% Pu HM 1 MOX is possible.

More recent advanced voloxidation studies have focused on the use of strong oxidants such as nitrogen dioxide and ozone (DelCul, 2010). These oxidants are capable of oxidizing uranium dioxide to trioxide, for example, Equation 5.12, and show promise that near-quantitative removal of tritium, carbon-14, noble gases, and iodine-129 are possible at temperatures lower than air or oxygen.

(5.12)UO2+O3UO3+O2

Carbides and nitrides

The conversion of carbides and nitrides to oxides could be an important step to allow

Carbides—removal of carbon that prevents the generation of organics during dissolution.

Nitrides—evolution of nitrogen-15 predissolution for recovery.

Carbides and nitrides—the removal of fission products for abatement.

These processes can be carried out with a variety of oxidants and temperatures. The conditions necessary tend to be more moderate compared to oxide due to the reactivity of the materials. However, the reactivity also leads to concerns over pyrophoricity of the materials, which means careful control to prevent ignition is a key consideration.

The oxidation of uranium carbide and nitrides with oxygen or air leads to the formation of triuranium octoxide. The accompanying density change is even greater than that for uranium dioxide oxidation, so although these reactions have not been studied in as much detail as the oxidation of uranium dioxide, the uranium dioxide oxidation literature can be used as a starting point. This leads to two conclusions:

Efficient removal of volatile fission products should be achievable.

Segregation of uranium and plutonium phases will lead to incomplete plutonium recovery during dissolution in nitric acid. The extreme case for carbide is illustrated in Equation 5.13.

(5.13)UPuC+xO2yU3O8+zPuO2+O2

However, there is a significant density change in the conversion of carbides or nitrides to dioxides, 13.6 or 14.3-11.0 g cm 3, respectively. This density change is sufficient to cause pulverization of pellets. This could be used to as a pretreatment process to convert mixed uranium plutonium carbides or nitrides to oxides, with high plutonium solubility. This type of process has been demonstrated for unirradiated mixed uranium plutonium carbides using carbon dioxide (Murbach and Turner, 1963) and steam (Flanary et al., 1964). These processes proceed with two steps: oxidation to the dioxide and carbon followed by reaction of carbon. Of the two reactants, carbon dioxide is often preferred as water is prone to leave residual carbon and produces hydrogen as a product. However, the use of carbon dioxide or water as an oxidant for irradiated carbons would result in the abatement of very large quantities of carbon dioxide or water during the abatement of carbon-14 and tritium.

The conversion of mixed uranium plutonium nitride to solid-state mixed dioxide may provide a method of removing nitrogen-15 without reducing the plutonium solubility.

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Plant life management (PLiM) practices for pressurized light water reactors (PWR)

Ph. G. Tipping, in Understanding and Mitigating Ageing in Nuclear Power Plants, 2010

18.3.1 Nuclear fuel and core and void coefficient of reactivity

The PWR nuclear fuel is enriched uranium dioxide (UO2) that contains about 3 wt% of fissionable 235U, or mixed uranium-plutonium oxide (MOX) in pellet form. A typical MOX fuel core consists of around one-third MOX, and two-thirds normally enriched UO2 fuel. The sintered UO2/MOX fuel pellets are sealed into thin cylindrical tubes (e.g. ‘Zircaloy’ cladding) to make rods, and depending on the design, 200–300 such rods are assembled to make a fuel assembly. A PWR can have 150–250 fuel assemblies (e.g. a 17 × 17 configuration) and an inventory of around 80 tonnes of uranium metal equivalent. Fuel cores have typical dimensions of around 3.5 m diameter and 3.5 m high. Refuelling is done according to burn-up, but 12-, 18- or 24-month cycle intervals can be used, or when about a third of the core needs to be replaced with fresh fuel.

It is convenient to introduce here the principle of the void coefficient of reactivity (VCR), and particularly, the negative VCR, which is a very important safety feature associated with the reactor system and fuel physics of PWRs. Since PWRs use normal (light) water as a neutron moderator (and coolant), any incipient boiling/vapour generation caused by operational instabilities or power surges/transients will reduce the neutron moderating capacity (degree of slowing down/thermalization) and the fuel fission processes become less efficient/probable and thus less heat is produced. (Note: A representative (average) core power density for currently operating commercial PWRs is about 100 MW/m3 and thus it is nearly twice as high as that in BWRs (56 MW/m3), and it is a factor of about 12.5 times more than that typical of PHWRs (CANDUs) (8 MW/m3).)

A positive VCR is a feature in channel-type reactors (e.g. Russian-design RBMK) where the combination of graphite blocks as the neutron moderator, and light water for cooling may, under very specific conditions, cause fuel fission processes to actually increase when local boiling/vapour occurs. As a consequence, increasingly more heat is rapidly added to the system, potentially leading to an event as typified by the major Chernobyl/Ukraine NPP accident on 26 April 1986 [16]. It is noted here that those RBMK reactors still in service today have undergone significant safety upgrades and have been robustly modified to greatly increase their resistance to power surges and their consequences. More manual control rods have been added and several operating practices and procedures have been changed to greatly enhance safety and more fexibility in operation. The heavy water cooled and moderated CANDU type of reactor also has a positive VCR, but this is much less than in RBMK types and the operational margins and control systems in place are much greater, thus any power excursions are readily managed [17].

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