As a result of the wide range of oxidation states for some of the actinide ions, as shown in Table

9.4, the control of the oxidation state may play an important role in chemical separations, particularly if plutonium is involved. The oxidation-reduction chemistry is also important to the behavior of actinide elements and compounds if released to the environment.

The selective extraction of plutonium from uranium or fission products depends on proper adjustment of the valence state of plutonium relative to the other ions from which it is to be separated. For instance, in decontaminating plutonium by extraction with TBP, plutonium must be oxidized to the tetravalent state, without bringing cerium into the tetravalent, ceric state. Again, to separate plutonium from uranium and the fission products in the tributyl phosphate extraction process, plutonium must be trivalent and uranium hexavalent.

A typical oxidation-reduction reaction of the type met in processing plutonium is the reduction of Pu4+ to Pu3+ by ferrous ion:

FeJ+ + Pu4* ->■ Fe3+ + Pu3* (9.8)

In dealing with groups of such equilibria it is convenient to break them up into two half-reactions, or oxidation-reduction couples, which indicate the mechanism by which electrons are transferred from the reducing agent to the oxidizing agent.

The two couples for the above reaction are

Pu3+ -*• Pu4+ + є~ (9.9)

and Fe2+ -* Fe3+ + e~ (9.10)

The electrons from each couple may be thought of as exerting an electromotive force, which is the oxidation-reduction potential of the couple. When equilibrium is reached in a solution containing a number of oxidation-reduction couples, the potentials of all couples must be equal, otherwise electrons would be transferred from one couple to another and further reaction would take place. Thus, if we can evaluate the dependence of oxidation-reduction potentials on concentration, we can determine equilibrium concentrations in solution of ions of mixed valence.

The oxidation-reduction potential of a couple E, in volts, is related to the change in free energy AG when 1 g-equiv of electrons is produced:

AG = — SE (9.11)

where 7 is Faraday’s constant, 96,487 J/(Volt-g-equiv). The minus sign is used because of the negative charge on the electron. When the components of a couple are in their standard states of unit activity and electrons are in their standard state, the free-energy change is the standard free-energy change ДG°, and the potential is the standard oxidation potential E°. These are related by

AG°=-3E° (9.12)

The standard state for electrons is customarily taken to be that corresponding to equilibrium in the reaction

^ H2 (?, 1 atm) -*• H*(unit activity) + e ~ (9.13)

whose standard oxidation potential by definition is zero.

where Kjrp is the equilibrium constant for the overall iron-plutonium reaction (9.8). At 25°C, we have for A’:

Afp = e3»-M(£F-£p)

For the plutonium-iron case, the values are [Al]

Pu3* -*■ Pu4+ + e~ Ep = —0.9819 V

Fe2+ ->■ Fe3+ + e~ Ep = — 0.7701 V

The equilibrium constant for the reaction (9.8) is

A:fp = e38.93(-0.7701+ 0.9819) = 3gl0

This shows that reduction of plutonium from tetravalent to trivalent can be made substantially complete with only a slight excess of ferrous ion.

A large negative oxidation-reduction potential means that the first member is a strong oxidizing agent. A positive potential means that the first member is a strong reducing agent (and could, in fact, reduce water) and the second member is a very weak oxidizing agent.

The general equation relating the equilibrium constant for a reaction involving the transfer of n electrons and standard oxidation-reduction potentials is

Equation (9.21), as written, involves the transfer of 2 equivalents of electrons from tin to plutonium, so that the equilibrium constant is

Three examples of pyrochemical processes that have been developed for purifying uranium or plutonium oxides are listed in Table 10.2.

Skull-reclamation process. The skull-reclamation process was developed by Argonne National Laboratory and used at the Idaho EBR-II Fuel Cycle Facility in the 1960s to recover uranium

from the crucible oxide residues, or skulls, remaining after the partial oxidation, melt-refining process described above. The process involved selective reduction and extraction of the oxides by magnesium-zinc alloys at controlled temperatures and reductant metal concentrations, followed by removal of the magnesium-zinc solvents by distillation.

Salt-transport process. The salt-transport process was studied by Argonne, with the objective of reprocessing short-cooled, high-burnup LMFBR fuel oxide with nonaqueous systems in which radiation damage of solvents would not be a problem. In this process, stainless steel cladding is removed from the fuel by solution in molten zinc at 850°C. The U02-Pu02 fuel is then reduced by a copper-magnesium-calcium alloy with a CaCl2-CaF2 flux at 800°C. This produces a salt solution of the more stable fission-product oxides (Cs2 0, SrO, BaO, and some rare earth oxides), a copper-magnesium solution of plutonium, the rest of the rare earths and the more noble fission-product metals (ruthenium, molybdenum, palladium, etc.), and a solid phase consisting mostly of uranium metal. Plutonium in the liquid copper-magnesium phase is purified by countercurrent extraction with 50 w/o MgCl2, 30 w/o NaCl, 20 w/o KC1, which extracts the rare earths selectively. Finally, the plutonium is separated from the noble fission-product metals by transport through a second 50 w/o MgCl2 salt phase to a 95 w/o Zn-5 w/o Mg Pu-acceptor alloy. This last salt-transport step, suggested by Chiotti and Klepfer [C7], gave the process its name. Argonne [V2] tested individual steps of this process, but did not conduct a complete demonstration with full-bumup fuel.

Salt-cycle process. The salt-cycle process was developed by Battelle Northwest Laboratory [H4] with the following objectives:

To permit reprocessing short-cooled fuel at the reactor site

To handle U02 and U02-Pu02 fuel without requiring conversion to other chemical forms To recover 99 percent of the plutonium and remove at least 80 percent of the neutron­absorbing fission products

To permit control of the plutonium/uranium ratio in recovered fuel

In this process, oxide fuel is dissolved in a molten chloride salt mixture through which C12-HC1 gas is flowing. Dissolved uranium and plutonium are then recovered as oxides by cathodic electrodeposition at 500 to 700°C. The process was demonstrated with kilogram quantities of irradiated fuel, with production of dense, crystalline U02 or U02-Pu02 reactor-grade material. Difficulties were experienced with process control, off-gas handling, electrolyte regeneration, and control of the plutonium/uranium ratio. Development has been discontinued.

For voloxidation to remove tritium completely from (U, Pu)02 fuel, it has been found necessary that the oxygen content of the fuel be increased to that corresponding to mixed U308 and PuOj. With LWR fuel, containing 1 percent or less plutonium, oxidation of U02 to U308 proceeds relatively rapidly and completely at temperatures around 600°C, with almost quantitative release of tritium. This favorable result is attributed to the swelling and disintegration of the fuel accompanying the phase change from cubic U02 to less dense orthorhombic U308.

With the Pu/(U + Pu) ratio of 0.20 or 0.25 proposed for the core of an LMFBR, the voloxidation process is more complex, very dependent on method of fuel fabrication, and sometimes incomplete. Oak Ridge National Laboratory [09] found that cubic dioxide fuel with a Pu/(U + Pu) atom ratio of 0.2 was oxidized to the desired mixture of U3Og and Pu02 in a two-step process only within a narrow temperature range of 500 to 600°C. First oxidation takes place rapidly to tetragonal U2 4 Ри0,бО7. Little swelling or disintegration occurs in this step, and little tritium is released. The second, slower oxidation step results in conversion of the fuel to a less dense mixture of orthorhombic U3Og and cubic Pu02 (oxygen/metal ratio of 2.53), with crumbling of the fuel and release of 98 percent or more of the contained tritium [FI]. The rate at which this second phase change occurs is strongly dependent on how the uranium and plutonium were homogenized, how the fuel was made, or possibly some other factor. The second change takes place most rapidly with coprecipitated U02-Pu02, next most rapidly with sol-gel fuel, and slowest with mechanically mixed fuel. Furthermore, Oak Ridge reported [09, p. 17]: “Under some conditions as yet undetermined, even fuel with Pu/(U + Pu) = 0.2 will not readily oxidize to oxygen/metal ratios in excess of 2.4,” and hence will not release tritium.

Dioxide fuel with a Pu/(U + Pu) atom ratio of 0.25 oxidizes readily to a cubic phase of empirical formula (и, Ри)4094, but with little swelling or tritium release. Further oxidation to a mixture of U308 and Pu02, with the desired swelling and tritium release, proceeds only with great difficulty.

Another difficulty with voloxidation of mixed U02 — Pu02 fuel is conversion of some of the Pu02 to a form insoluble in nitric acid. For these reasons the German workers Baumgartner and Ochsenfeld concluded [B6] that “voloxidation is no longer considered as a treatment step preceding dissolution of LMFBR elements.” However, work on voloxidation of mixed U02- Pu02 fuel is continuing at Oak Ridge.

Apart from the more exotic approaches to waste disposal that have been mentioned before, shallow burial and sea disposal are widely used. Disposing of liquids into isolated aquifers or exhausted oil lenses has been mentioned as a special technique for tritium waste.

Burial grounds have become quite common, mainly in those countries where nuclear activities have a long history and originated in weapons research. The safety of this disposal technique is largely dependent on the type of soil, particular groundwater occurrence, and on the type of land use. Presently, a volume of over 200,000 m3 containing approximately 2 X 106 Ci of radioactivity including 80 kg of plutonium are disposed of in commercial burial grounds in the United States and about the same order of magnitude in burial grounds established by the former U. S. Atomic Energy Commission. Although this technique cannot be considered unsafe when properly conducted, some incidents of radionuclide migration resulting in off-site contamination have occurred. It is fair to say that shallow burial of non-high-level, non-alpha waste may be safe in remote areas, but these usually do not exist in Europe. The overall policy of establishing burial ground needs reconsideration.

The term sea disposal includes two basically different techniques, namely, disposal into coastal waters and deep-sea disposal. Deep-sea disposal may be perfectly safe if handled

2000 4000

Distance from center of repository, ft. (1000 ft. = 304.8 m)

Vertical Cross Section of Repository

(2)b: surface of container

End View of Burial Zone Side View of Burial Zone

for High-level Waste for High-level Waste

Figure 11.27 Schematic cross sections of proposed HLW repository. (From Cheverton and Turner [C1J.) responsibly. For certain types of waste that are difficult to deal with on land, such as bulky parts from decommissioning, it may even be the most appropriate technique. Deep-sea disposal has been practiced mainly under the supervision of international agencies. Disposal into coastal waters as practiced with non-high-level liquids from European reprocessing plants, however, is highly debated.

7.1 Requirements

The principal functional requirements of a solvent extraction contactor are as follows:

1. To develop sufficient interfacial area to promote transfer of extractable components between phases

2. To facilitate countercurrent flow of the two phases, without excessive entrainment Additional considerations in selecting contacting equipment are as follows:

3. It should have flexibility to operate under varied conditions of flow ratios and concentra­tions.

4. It should be mechanically dependable and easy to operate and maintain.

5. It should be compact and have low holdup of process materials.

6. Initial cost and operating cost should be low.

The importance of these different factors varies with the application. Although reliability is important for any application, it is particularly important when processing highly radioactive materials, as in reprocessing discharged reactor fuel, where the intense radioactivity makes normal methods of maintenance difficult or impossible. Equipment handling such highly radioactive solutions must be enclosed in massive shielding and operated, and perhaps main­tained, by remote control. If concrete is used as shielding material, thicknesses of 2 to 3 m are often required. To control the cost of such massive shielding, it is important to reduce its bulk, and this means using compact solvent extraction contactors. If vertical column contactors are used, their height should be kept to a minimum. If a horizontal array of mixer-settler contactors is used, the layout of equipment and piping should be compact.

Those applications that require remote operation and maintenance dictate the use of simple, rugged equipment, with a minimum of moving parts and with little tendency to foul, rust, clog, or corrode.

The intense radioactivity associated with the reprocessing of discharge fuel, and the degradation and decomposition of organic solvent when exposed to ionizing radiation, require that the amount of solvent exposed to radiation be kept to a minimum. Also, the length of time that the organic is exposed to radiation should be kept small. This places a premium on compact contacting equipment, with high throughput per unit volume.

Nuclear criticality places special constraints on the size of contacting equipment in fuel reprocessing. This is particularly important when reprocessing highly enriched uranium fuel or for the stripping-scrubbing contactors that separate plutonium from low-enriched uranium, as illustrated in Fig. 4.5. Both 23S U and plutonium fission. As shown by the criticality data in Table 4.11, as little as 760 g of 23SU or 510 g of plutonium can form a critical mass when dispersed at the optimum concentration through a hydrogenous medium, such as a nitric acid solution or organic, with relatively little fission products or nonfissile uranium [A2, Т1]. The sizes of the contactors and other process equipment must be kept small enough to promote neutron leakage and make criticality impossible [C4]. Limiting dimensions may be as small as 14 cm in diameter for a cylindrical column contactor or 4.6 cm in height for an array of mixer-settlers in horizontal slab geometry [A2, Т1]. Larger equipment sizes are acceptable for process operations that do not involve solutions of relatively pure fissile material, such as the extracting-scrubbing contactors that separate the fission products from low-enrichment uranium fuel. The allowable dimensions and throughput of criticality-limited process equipment can be increased by incorporating fixed neutron absorbers, i. e., “poisons,” such as boron or gadolinium, without the equipment.

Contactors with low inventory of process solutions are also important when the material pro­cessed is valuable, such as the plutonium recovered from irradiated fuel. Low inventory is also impor­tant in maintaining a close accountability of the total inventory of fissionable material processed. [10] [11] [12] [13]

As an example of uranium concentration by carbonate leaching, a brief description will be given of the uranium mill owned by United Nuclear-Homestake Partners at Grants, New Mexico fJ1, М3]. Mill capacity in 1977 was 3500 short tons ore per day. The ores treated are primarily sandstones, whose silica or limestone grains are cemented by intergranular uranium-bearing minerals such as coffinite, uraninite, tyuyamunite, and carnotite. Average ore composition is 0.21 w/o и308, with small amounts of vanadium, molybdenum, and selenium.

In this mill, uranium is recovered from carbonate leach liquor by precipitation with sodium

1. Alkaline leach

or ammonium hydroxide. Another U. S. mill employing carbonate leaching recovers uranium by anion exchange (Table 5.19).

Crushing and grinding. Figure 5.4 is a schematic flow sheet of the principal process steps in this mill. Ore is crushed dry to particles smaller than 0.5 in. Crushed ore is ground in two conical ball mills with recycled filter washings containing sodium carbonate and bicarbonate and some recycled uranium. Ball mills are operated in closed circuit with a spiral classifier which returns oversize particles to the mill. Classifier overflow is 16 to 20 percent solids, with 95 percent finer than 48 mesh and 65 percent coarser than 200 mesh. This is finer grinding than used in the Kerr-McGee acid-leach mill (see Sec. 8.5).

Leaching. Ground slurry is divided between two similar leaching circuits. The slurry is thickened to 52 to 54 percent solids in 20-in (0.5-m) diameter cyclones and 100-ft (30-m)

Figure 5.4 Schematic flowsheet, United Nuclear-Homestake Partners uranium mill.

diameter settling tanks, with overflow recycled to grinding. The slurry is preheated to 63°C, and sodium carbonate is added up to 35 to 37 g Na2C03/liter, with about 7 g NaHC03 /liter. Slurry is pumped at a pressure of about 4.5 atm in series through two lines of four stirred autoclaves 12 ft (3.7 m) in diameter and 16 ft (4.9 m) high, which are heated to about 95°C, with a residence time of about 4.5 h. Air is pumped through the autoclaves to oxidize tetravalent uranium.

Less soluble ores are leached for an additional 36 h at 85°C and atmospheric pressure in nine large tanks 19 ft (5.8 m) in diameter by 38 ft (11.6 m) deep, stirred, and supplied with additional air. Enough excess sodium carbonate is used to bring the pH up to 11.

Filtration. Uranium-bearing solution is separated from leached solids in three filter stages with countercurrent washing. Each stage consists of five 650-ft2 (60-m2) and two 570-ft2 (53-m2) rotary-drum vacuum filters, operated batchwise. Filter cake from the fust stage is washed with filtrate from the third stage; filtrate and washings from the first stage constitute the leach liquor, which contains 3 to 3.5 g U308/liter. Filtrate and washings from the third stage are used in the first stage to wash the filter cake and then to reslurry it and move it to the second stage. Filter cake from the second stage is washed and reslurried with sodium carbonate solution; the filtrate and washings containing some uranium are returned to the ball mills. Filter cake from the third stage is washed with dilute carbonate solution from the tailings pond; filtrate and washing? are used to wash and reslurry filter cake in the first stage. Washed filter cake from the third stage is reslurried and pumped to the tailings pond.

Leaching and washing reduce the U308 content of tailings to around 0.01 percent.

Precipitation. Leach liquor from the first stage filters is clarified in a thickener, mixed with about five times its content of recycled yellow cake,1*” heated to 74°C, and held about 5 h in stirred tanks. Recycle increases the soluble uranium content from about 3.5 to 7.5 g U3Os/liter and was found needed to improve the completeness of uranium precipitation in the next step. Precipitation of uranium by the reaction

2Na4U02(C03)3 + 6NaOH -* Na2U207 + 6Na2C03 + 3H20

is accomplished by adding sufficient sodium hydroxide to leave an excess of 5 g КаОНДкег after reaction. The mixture flows through eight heated and stirred tanks in series for an 8-h residence time at 74°C. Reaction product is filtered to produce an impure solid Na2U207 and a solution of NaOH and Na2C03. The solution is recarbonated with flue gas and returned to the leaching circuit via the second-stage filter wash.

Purification. The impure Na2U207 contains 5 to 6 percent V205, probably coprecipitated with the uranium. This impurity is converted to soluble sodium vanadate, NaV03, by roasting the impure yellow cake with about an equal mass of Na2C03 at 860°C for a half-hour, cooling the mass, and leaching it with water. The leached product is filtered and washed, producing a solution from which vanadium is recovered, and a yellow cake that, after drying, contains about 85 percent U308, 0.2 to 0.8 percent V205, and up to 7.5 percent Na.

Although the principal constituents of monazite are rare earth and thorium phosphates, its composition varies widely within a given deposit and from place to place. Table 6.16 gives the composition of monazite concentrates from different locations.

1.13 Processes for Opening Up Monazite

Monazite is chemically very inert. Two general methods that are used for opening up monazite and making its constituents sufficiently reactive to permit extraction and separation of thorium, uranium, and rare earths are

1. Reaction with hot, concentrated caustic soda solution

2. Dissolution in hot, concentrated sulfuric acid

The caustic soda process has been used on a large scale in Brazil [B6] and India [Dl] and has been investigated on a pilot-plant scale at Battelle Memorial Institute [B2, Cl] in the United States. A brief description is given in Sec. 8.4.

Sulfuric acid has been used to dissolve monazite in Europe, Australia, and the United States. The numerous processes used to separate thorium from the acid leach liquors are listed

in Sec. 8.5, with a more detailed description of one process, developed at Iowa State College [В1].

Solvent extraction processes for recovering thorium from monazite sulfuric acid leach liquor are described in Sec. 8.6.

The chemical composition of fission products in discharge fuel is controlled by the long-lived and stable species. The amounts of most of the fission-product chemical elements change but little for thousands of years after discharge. Those elements that do change significantly in amount over long decay periods include:

1. Cesium, which includes appreciable quantities of 30-year 137Cs

2. Hydrogen, which consists entirely of 12.3-year 3H

3. Niobium, which consists almost entirely of 35-day 9sNb

4. Promethium, which consists entirely of 2.62-year 147Pm

5. Strontium, which contains appreciable quantities of 27.7-year 90Sr

6. Technetium, which consists entirely of 2.12 X 105-year "Tc

The elemental composition of the fission products in fuel discharged from the uranium-fueled PWR (Fig. 3.31) is listed in Table 8.2 and is plotted in Fig. 8.2. The composition expressed in elemental atoms per fission-product pair is the effective long-term elemental yield per fission, so the sum over all elements is equal to 2.

Plutonium oxides. The phase diagram of the plutonium-oxygen system is shown in Fig. 9.2. The observed compounds are the stoichiometric Pu203 and Pu02 and the nonstoichiometric Pu01S2 and Pu161. PuO has also been shown to exist, but only under extreme conditions. No oxide of higher oxidation state than Pu02 has been formed.

Plutonium dioxide is the form of plutonium most commonly specified for fuel for power reactors. It has the same general features already described for pure U02 fuel, such as high melting point, irradiation stability, compatability with metals and with reactor coolants, and ease of preparation. In most designs of plutonium-fueled power reactors the fuel is a mixture of uranium and plutonium oxides.

Pu02 is formed when plutonium or its compounds, except the phosphates, are ignited in air. The most common starting materials are the nitrate or oxalate. Heating Pu(III) or Pu(IV) oxalate at 1000°С in air results in pure crystalline Pu02. The physical appearance of the dioxide depends on its origin, ranging from yellow-black to green and from powder to shiny particles. The Pu02 crystalline density is 11.46 g/cm3. The melting point varies from 2280°C in helium to about 2400°C in air [Cl].

The only other binary oxide of plutonium of practical importance is the peroxide, which is the basis of a process for the purification of plutonium and its conversion to the metal. Addition of H202 to an aqueous plutonium solution first converts plutonium ions to the tetravalent state.

Further addition of peroxide precipitates the plutonium peroxide complex, a nonstoichiometric compound whose composition and crystalline form depend on precipitating conditions. Manage­able hexagonal precipitates are promoted by adding sulfate ions or by precipitating at acidities as high as 4.7 M. The dry peroxide is unstable, decomposing rapidly and sometimes explosively, especially when iron is present. Plutonium peroxide is a stable solid in acid of concentrations up to 5 N. Dry sulfate-free plutonium peroxide can be fluorinated directly at 600°C in HF containing small quantities of oxygen, yielding PuF4 which can be readily reduced to the metal [C1,C2,M1].

Most of the designs for power-reactor fuel utilizing recycled plutonium involve the use of the mixed oxides of plutonium and natural or depleted uranium. The mixed-oxide fuel is formed either from mechanically mixed powders of the individual Pu02 and U02 binary oxides or by calcining a coprecipitated uranium-plutonium compound. A portion of the phase diagram of the uranium-plutonium-oxygen system at 20°C, in the region of U02-U3 08-Pu2 03-Pu02, is shown in Fig. 93 [II, K2]. The phase boundaries deduced for the same region of the uranium-plutonium — oxygen system at 400,600, and 800°C [II] are shown in Fig. 9.4. The mixed uranium-plutonium oxides with the stoichiometric dioxide composition form a continuous solid solution from U02 to Pu02, with the fee fluorite structure, which is stable also at high temperature.

Oxidation of mixed oxides to overall oxygen-to-metal ratios greater than 2, and subsequent cooling to 20° C, results in a two-phase region M02+x + M409 up to the oxygen-to-metal ratio of 2.20 and up to a Pu/(U + Pu) ratio of 0.30. For overall oxygen-to-metal ratios of 2.20 to 2.27, a single phase M409 exists that is stable up to 1000°C. Oxidation of mixed uranium-plutonium oxides containing more than 39 percent plutonium results in the oxidation of uranium from U(IV) to U(V). An equimolar uranium-plutonium oxide forms a single phase of overall composition M409 when all uranium has been oxidized to U(V). At 1400°C a single fluorite phase exists for all plutonium concentrations and for oxygen-to-metal ratios up to 2.27.

The complete miscibility of the stoichiometric uranium-plutonium dioxide results in the simple liquidus-solidus melting-point curves of Fig. 9.5. The curves are consistent with ideal-
solution theory for heats of fusion of 112 ± 13 kcal/mol for U02 and 16.8 ± 13 kcal/mol for Pu02 [II].

The preferred industrial process for manufacturing mixed-oxide uranium-plutonium fuel involves mechanical mixing of U02 and Pu02 powder, followed by compaction and sintering above 1200°C. At the temperatures normally used in commercial sintering of the mixed oxides only a small portion of the sintered material contains the solid solution of U02 — Pu02. Even at temperatures as high as 1400 to 1750°C long sintering times are required for complete homo­genization of the binary oxides.

It is important that the size of the remaining discrete particles of Pu02 be small enough so that fission heat generated in the particles, particularly during rapid power transients, is not suf­ficient to locally overheat the Pu02 particles. This requirement, which is most stringent for mixed-oxide fuel for fast-breeder reactors, is fulfilled by using Pu02 powder with particle sizes of less than about 0.01 cm [Bl].

Another important consideration is the problem of dissolving the mixed-oxide fuel for subse­quent reprocessing and plutonium recovery after the irradiated mixed-oxide fuel has been dis­charged from the reactor. When plutonium dioxide is in solid solution with uranium dioxide at low concentrations, as in the case of plutonium created during the irradiation of uranium dioxide

/0 0

fuel, the contained plutonium is soluble in the nitric acid normally used in fuel dissolution. How­ever, pure Pu02 is very difficult to dissolve in nitric acid, unless a fluoride catalyst is used, so problems with the dissolution of fuel containing crystallites of pure plutonia can be expected.

The fabrication and irradiation histories of uranium-plutonium mixed-oxide fuels strongly affect their solubility in nitric acid [L3, G3]. Fuel containing up to 28 percent Pu02 can be dissolved in a few hours in boiling 6 to 10 M HN03, provided that all of the plutonium is in solid solution with the uranium. However, poorly fabricated fuels consist of very small islands of essentially pure plutonia in a matrix of U02-Pu02 solid solution. Consequently, most of the fuel dissolves readily, leaving a refractory residue containing 1 to 10 percent of the plutonium. This residue is readily dissolved in 8 M HNO3-O. O5 M HF. Irradiation improves the solubility of poorly fabricated fuels, but it cannot be depended on to erase the solubility defect [01, L3]. Fuels completely soluble in HN03 have been prepared successfully by methods of coprecipitation, sol-gel, and mechanical mixing of the separate oxides followed by sintering [G3].

Plutonium carbides. Carbides of plutonium and uranium are of interest as high-performance fuel for advanced breeder reactors. As compared with oxide fuel, the higher density of the mono­carbide (U, Pu)C results in higher breeding ratio. Although the carbides cannot operate at as high a temperature as can the oxides, the much greater thermal conductivity of the carbides allows greater heat generation rates in the fuel. However, the technology for carbide fuel is not as far advanced as that for oxide fuel.

Plutonium carbides that have been prepared are PuC, PuC2, and Pu2C3, formed by the reac­tion of graphite with metallic plutonium or PuH3 at high temperature. The products are usually mixtures of PuC and Pu2 C3. Plutonium oxide can also be reduced by carbon, but the stability of Pu(0, C) prevents the formation of plutonium carbide of high purity. PuC2 exists only at tempera­tures above 1750°C. Plutonium monocarbide exists only as a substoichiometric compound, with a crystalline density of 13.58 g/cm3, in the presence of excess carbon. It undergoes peritectic decomposition to the metallic liquid and Pu2C3 at 1654°C [K2].

For reactor fuel, the ternary uranium-plutonium-carbon monocarbide is prepared by reduc­tion of (U, Pu)02 with graphite [FI], by melting a uranium-plutonium alloy with graphite, or by melting separately prepared individual carbides in an electric arc [K2]. Even though at low temperatures UC exists in the stoichiometric composition, the need for excess carbon for the existence of PuC limits the region of PuC-UC miscibility to a maximum of 35 a/о (atom percent) plutonium at room temperature. At higher plutonium concentrations the excess carbon is pre­cipitated as Pu2C3.

The unirradiated mixed carbide (U, Pu)C readily hydrolyzes in water or acid, but neutron irradiation profoundly reduces the tendency toward hydrolysis.

Plutonium nitride. Unlike the corresponding uranium-nitrogen system, only the one plutonium nitride PuN exists. It is prepared by heating plutonium hydride in nitrogen at 250 to 400°C, by reacting plutonium metal with a hydrogen-ammonia mixture at 600° C, or by direction reaction of molten plutonium with nitrogen at 1000°C. Plutonium nitride forms solid solutions with UN. However, because of the appreciable volatility and dissociation of PuN at temperatures at about 1600°C and above, the ternary (U, Pu)N is less attractive as a nuclear fuel than pure UN [K2, S4].

Plutonium hydrides. Plutonium hydrides are made by reacting plutonium metal with hydrogen at atmospheric pressure and at temperatures between 50 and 300°C, forming a series of hydrides up to PuH3. Plutonium hydride is a useful intermediary in the formation of other plutonium compounds.

Plutonium halides. Table 9.18 lists plutonium halides together with some of their more significant properties.

PuF3 and PuF4 are important intermediates in the production of plutonium metal. The trifluoride is made by reacting Pu02 with a mixture of HF and H2 at 600° C:

Pu02 + 3HF + jH2 -+• PuF3 + 2H2 О (9.37)

The tetrafluoride is made by reacting oxide or oxalate in HF at 550°C, in the presence of oxygen to prevent reduction of tetravalent plutonium:

Pu02 + 4HF -*■ PuF4 + 2H2 О (938)

The volatile plutonium hexafluoride can be prepared by fluorination of the tetrafluoride at 550°C:

PuF4 + F2 -+ PuF6 (939)

or by direct fluorination of the oxide. In contrast to UF6, PuF6 does not sublime when heated at atmospheric pressure. It first melts, at 51.59°С, and then boils at 62.16°C. Its triple point is 51.59°C and 533 Ton. The vapor pressure of PuF6 above liquid PuF6) in the range from 51.59 to 77.17°C, is given by [Cl]

і 007 5

logiop(Torr) = — — 1.5340log, о T+ 12.14545 (9.40)

Unlike the stable uranium hexafluoride, which has a negative free energy of formation, plutonium hexafluoride is thermodynamically unstable. It dissociates to F2 and the relatively nonvolatile PuF4 , although the rate of thermal decomposition is very low at room temperature. If the specific alpha activity of plutonium is equivalent to that of239 Pu, the rate of decomposition of solid PuF6 at room temperature is controlled by radiolytic decomposition, amounting to 1.5 percent per day [W2].

Plutonium also forms the ternary oxyhalides PuOF, Pu02 F2, PuOCl, PuOBr, and PuOI.

Plutonium in the aqueous phase leaving partitioning contains around 1 percent of the feed uranium, an uncertain fraction of the feed neptunium, and fission products. This plutonium is usually purified by two additional cycles of solvent extraction. In each, plutonium is made tetravalent and, in the A contactor, is extracted by TBP, together with the uranium, thus separating it from most of the fission products and neptunium, here mostly pentavalent. In the В contactor, the plutonium is stripped selectively from uranium into the aqueous phase, either by use of 0.35 M HN03 or by reducing plutonium to the trivalent state prior to stripping. In some plants, final plutonium purification is by anion exchange from nitrate solution. Some others use precipitation as plutonium oxalate with oxalic acid. Optimum conditions are reported [S23, p. 449] to be as follows: HN03, 1.5 to 4.5 M; H2C204, 0.05 to 0.15 M. A problem in plutonium purification systems is heating and radiolysis of solvent or resin owing to the high alpha activity.

Many reprocessing plants are required to convert purified nitrates containing plutonium to oxides before shipment. Conversion processes are described in Chap. 9, Sec. 4.7.