Objectives. The objectives of fuel dissolution are (1) to bring the uranium and plutonium in the fuel completely into aqueous solution; (2) to complete the separation of fuel from cladding; (3) to determine as accurately as possible the amounts of uranium and plutonium charged to reprocessing; and (4) to convert uranium, plutonium, and fission products into the chemical states most favorable for their subsequent separation.

Reactions. Because the Purex process requires that the elements to be separated be present in aqueous solution as nitrates, the dissolvent is always nitric acid. The principal reactions that take place are

3U02 + 8HN03 -+ 3U02(N03)2 + 2NO + 4H20

and U02 + 4HN03 -* U02(N03)2 + 2N02 + 2H20

Ordinarily, both reactions take place to some extent, with the first dominant at acid concentrations below 10 M and the second at higher concentrations [S8]. In principle, formation of gaseous reaction products could be avoided by addition of oxygen directly to the dissolver:

2U02 + 4HN03 + 02 ->• 2U02(N03)2 + 2H20

This process is known as “fumeless dissolving” and is used in European plants. Practically, small amounts of nitrogen, nitrogen oxides, and gaseous fission products are also formed. Reference [07] gives an example of fumeless dissolving.

Plutonium in oxide fuel dissolves as a mixture of tetravalent and hexavalent plutonyl nitrates, both of which are extractable with TBP. Neptunium dissolves as a mixture of inextractable pentavalent and extractable hexavalent nitrates.

Most of the fission products go into aqueous solution. However, at high bumups above

30,0 MWd/MT, some elements such as molybdenum, zirconium, ruthenium, rhodium, palladium, and niobium may exceed their solubility limits and be present as solids.

In the solution, americium, curium, and most of the fission products are in a single, relatively inextractable valence state. Iodine and ruthenium are important exceptions. Iodine may appear as inextractable iodide or iodate or as elemental iodine, which would be extracted by the solvent and react with it. Ruthenium may appear in any valence state between 0 (insoluble metal) and 8 (volatile ruthenium tetroxide) and, at valence 4, may form a number of nitrosyl ruthenium (Ru^NO) complexes of varying extractability.

An important objective of dissolution and the preconditioning of feed solution prior to extraction is to convert these fission-product elements into states that will not contaminate uranium, plutonium, or solvent in subsequent solvent extraction.

Dissolution rates. Uranium dioxide dissolves more rapidly than Pu02 or Th02. The time required for dissolving more than 99.5 percent of the U02 from unirradiated stainless steel-clad U02 pellets was found to be 40, 70, and 110 min for ;-, 1-, and 2-in chopped lengths, respectively [W4]. Irradiated fuel usually dissolves faster, probably because of cracking during irradiation. These tests were made with from 150 to 200 percent excess of 10 M HNO3 at a temperature just below boiling. The instantaneous dissolution rate in 8 M HN03 is about one-half that in 10 M acid. Because extensive foaming results when fuel is added directly to boiling nitric acid, the preferred procedure in a batch dissolver is to add U02 to cold acid, then bring the solution to just below the boiling point, with adequate cooling available to deal with heat evolved from chemical reaction and radioactive decay.

The rate of dissolution of Pu02 in nitric acid is slower than U02 and depends on the plutonium/uranium ratio, the methods used to fabricate fuel, and the conditions of irradiation. At one extreme, plutonium produced at low concentration in U02 by transmutation dissolves almost as rapidly as the associated U02. At the other extreme, plutonium present as Pu02 mixed mechanically with U02 without proper sintering dissolves much more slowly and less completely than U02. Plutonium present as a solid solution (U, Pu)02 at the concentration of 20 to 25 percent used in breeder-reactor fuel dissolves at an intermediate rate.^

In all cases, however, dissolution of irradiated fuel in nitric acid leaves some plutonium associated with undissolved fission products. This plutonium can be leached from the residue with mixed nitric and hydrofluoric acids or with mixed nitric acid and ceric nitrate, Ce(N03)4 [U2]. Residue from irradiated mixed U02-Pu02 fuel was 99.94 percent dissolved in 4 h by treatment with 4 M HNO3-0.5 M Ce(IV). Ceric nitrate is preferred to HF in the secondary dissolution step because cerium is already present as a fission product, and its addition does not complicate subsequent solvent extraction. Use of Ce(IV) in the primary dissolution step is undesirable because it would convert all plutonium to the less extractive hexavalent state and would volatilize much of the ruthenium as Ru04.

Separation from cladding. After reaction of fuel with acid has been completed, the resulting solution and any suspended fine particles are drained from the coarser cladding fragments. The cladding is washed, first with dilute nitric acid and then with water. The cladding is checked by gamma spectroscopy to establish removal of adherent fuel Tand then discharged for packaging as radioactive waste. The fuel solution, possibly containing suspended particles, is clarified by centrifugation. Centrifuged solids are accumulated and periodically leached as described above for recovery of plutonium and uranium.

Accountability measurements. The dissolver solution and washings are collected in a calibrated accountability tank and mixed thoroughly. The volume and density of the solution are measured as accurately as possible, and samples are taken for determination of uranium and plutonium concentrations. This is the first point in reprocessing at which a quantitative measure of input amounts can be made. Even here, measurement is difficult because of the intense radioactivity.

tDissolution of Pu02-U02 fuel is also discussed in Sec. 6.8.

Conditioning of feed. Before solvent extraction, the concentrations of nitric acid and uranyl nitrate are brought to the desired values by addition of water and/or nitric acid, as required. Preferred concentrations are HN03, 2 to 2.5 M U02(N03)2, 1.2 to 1.4 Af.

It is usually considered desirable to bring all plutonium to the most extractable, tetravalent state, although this step was not found necessary at West Valley. Sodium nitrite was formerly used for this purpose, but N204 or hydroxylamine is now favored because each adds no nonvolatile material to the aqueous phase. With N204, hexavalent plutonium is reduced:

Puvi022+ + N204 + 2H+ ->■ Pu4+ + 2HN03 Any trivalent plutonium that might be present would be oxidized:

4Pu3+ + N204 + 4H+ -*■ 4Pu4+ + 2NO + 2H20

Prevention of criticality. In dissolving fuel obtained from irradiating material more enriched than natural uranium, precautions must be taken to prevent accumulation of a critical mass in the dissolver. Three general methods are (1) use of subcritical geometry, (2) control of fissile material concentration, or (3) addition of a soluble neutron absorber with the dissolver solvent. For subcritical geometry, dissolvers have been built as thin slabs or long cylinders of subcritical diameter. A good example of subcritical geometry combined with concentration control is the dissolver used in the West Valley plant of Nuclear Fuel Services, Inc., shown in horizontal cross section in Fig. 10.6. Fuel baskets were 7 ft high and 8 in or less in diameter. The basket diameter selected for a particular fuel was one that limited the concentration in the 3-in

annulus and 10-in cylinders after dissolving to 60 percent of the critical value. Nuclear interaction between the cylinders was prevented by addition of 0.5 w/o natural boron to the concrete which provided 30-in separation between cylinders.

The Barnwell plant of Allied-General provides an example of use of soluble poison. There it is proposed that 5.6 g natural gadolinium, as nitrate, per liter be added to the nitric acid solvent. At the design concentrations of plutonium and uranium in dissolver solution, this will prevent criticality even with fully enriched 235 U (Prob. 10.1).

Dissolution equipment. Dissolution equipment, termed dissolvers, must provide for (1) adding fuel and dissolvent; (2) removing the product solution, undissolved solids, and gaseous effluents;

(3) maintaining proper contact between fuel and dissolvent; and (4) controlling the dissolution rate. Dissolvers may be characterized by the mode of fuel addition as either batch or continuous, or by their shape as column, slab, annular, or pot. The first three shapes are used for geometric control of criticality. Of the many types of dissolver that have been used, only a few examples can be described.

Batch pot dissolvers have been widely used, especially for low-enrichment fuel. The big advantage of batch operation is simplification of charging fuel and discharging residues. A disadvantage is the variable reaction rate, which is highest at the start when a large quantity of fuel is present, and which becomes much smaller toward the end when most of the fuel has been dissolved. The dissolution rate can be made more uniform by varying the concentration of dissolvent during the cycle. Dissolver product from a previous cycle, partially saturated with uranium, may be charged at the beginning of a cycle, to produce the most concentrated solution. When the reaction slows down, this solution may be replaced by fresh nitric acid, to produce a partially saturated solution to be used as solvent at the beginning of a subsequent cycle. Finally, at the end of a cycle, the residue in the dissolver would be washed with water to remove the remaining acid and fuel solution.

A batch dissolver typically is provided with heating coils to bring the solution to the desired temperature, cooling coils to remove the heat of reaction when it is most rapid, corrosion-resistant baskets or other containers to hold the fuel undergoing dissolution and retain cladding hulls at the end of the operation, and a cover to prevent escape of steam, nitric acid vapors, and volatile fission products and lead them to a condenser and fission-product traps. For ease of placement and removal, the cover may be sealed by a flanged ring that dips into a trough containing a sealing liquid. Recirculation of liquid through the dissolver is sometimes used to provide more uniform conditions and increase reaction rates. Product solution and undissolved sediment are withdrawn from the bottom. For dissolvers using nitric acid, heavy-gauge stainless steel has satisfactory corrosion resistance.

The volume of a batch of fuel to be charged to a batch dissolver may be evaluated from the product of the desired fuel reprocessing rate and the time required to complete the dissolving cycle. The cycle time may be estimated from small-scale experiments that simulate the geometry and time-temperature-concentration variations in the production dissolver, with a substantial allowance for inability to mock up accurately all relevant conditions of a production dissolver.

Continuous dissolution is especially advantageous when fuel and cladding are to be dissolved completely, as there is then no problem in removing undissolved solids from the dissolver. In such a case, fuel may be charged continuously at the top, dissolvent may be fed continuously, and dissolved solution removed continuously. The volume of undissolved fuel in the dissolver adjusts itself automatically so that the rate of solution balances the rate of addition. The big advantages over batch dissolution are smaller dissolver volume, more uniform product solution composition, steady gas evolution rate, and smaller and more efficient absorption system. It is estimated [B12] that the volume of a continuous dissolver may be from one-tenth to one-twentieth that of a batch dissolver of the same average dissolving rate.

This is especially advantageous for enriched fuels, where criticality limits dissolver dimensions.

The overall dissolution rate in a continuous dissolver is controlled by the metal feed rate, the temperature, the concentration of feed solution, and its flow rate. The metal composition of product solution at steady-state operation is just the ratio of the metal mass feed rate to the solution volume feed rate.

In a continuous dissolver with no solid residue, the solid necessarily flows downward. Liquid flow may be either down or up. With liquid downflow the dissolver is sometimes called a trickle dissolver. With liquid downflow it is preferable to remove off-gases at the bottom, to prevent flooding. With liquid upflow the dissolver is sometimes called a flooded dissolver. Off-gases separation from liquid is simpler with liquid upflow than with downflow.

When a solid residue such as cladding hulls remain, design of a continuous dissolver is much more complicated. Provision must be made for washing dissolver solution from the residue and for discharging the residue without escape of off-gases. A number of possible design concepts for continuous dissolvers have been tested by Oak Ridge National Laboratory [G15, 012].

Distribution coefficients of neptunium in 30 v/o TBP depend on neptunium valence, tempera­ture, and concentrations of uranyl nitrate, nitric acid, and other nitrates. At the nitric acid concentrations below 4 M usually used in Purex processes, the distribution coefficient of hexavalent neptunium is higher than that of tetravalent neptunium at the same nitric acid and uranyl nitrate concentrations. Both are much higher than that of pentavalent neptunium. Both tetravalent and hexavalent neptunium are extracted as the complexes with two molecules of TBP, Nprv(N03)4*2TBP and NpVI02(N03)2 -2TBP.

Table 10.23 lists principal sources of information on distribution coefficients of neptunium between 30 v/o TBP and aqueous solutions of uranyl nitrate and nitric acid.

Distribution coefficients of tetravalent and hexavalent neptunium can be correlated conveniently in terms of the separation factor from hexavalent uranium, i. e., the ratio of the distribution coefficient of neptunium to that of uranium.

Tetravalent neptunium. Srinivasan et al. [SI8, SI9] measured distribution coefficients of tetravalent and hexavalent neptunium and hexavalent uranium as functions of nitric acid and uranyl nitrate concentrations. At 45 and 60°C, the ratio of the observed [S19] separation factor for tetravalent neptunium to that of hexavalent uranium can be correlated within an average deviation of 6 percent by Eq. (10.25),

j? NBgv) = 0.01129 exp (0.3208лгщо — + 0.03636r°c) (10.25)

^UfVI) 3

At 25°C the equation is less satisfactory, with an average deviation of 18 percent from observations by Srinivasan et al. [S18],

Hexavalent neptunium. Distribution coefficients of hexavalent neptunium at 25, 45, and 60°C measured by Srinivasan et al. [S18, S19] are simply related to measured distribution coefficients for hexavalent uranium by Eq. (10.26), with an average deviation of only 5 percent,

= 0.54 (10.26)

^U(VI)

at all uranium concentrations and at nitric acid molarities between 1 and 4 Af. Germain et al.’s observed [G6] Np(VI) distribution data at 22°C yield an average value of 0.47 for this ratio.

In the HA extracting and HS scrubbing sections of the Purex process, pentavalent neptunium is partially oxidized to the hexavalent state by nitrate ion,

2Npv02+ + N03* + 3H+ -*• 2NpVI022+ + HN02 + H20

when nitrous acid is present to act as catalyst. For the reaction to proceed at a useful rate, the

HN02 concentration of the aqueous phase must be over 0.00004 M. At equilibrium, neptunium in the aqueous phase is then divided between the hexavalent and pentavalent states. The ratio of hexavalent to pentavalent neptunium is given by Eq. (10.27), obtained from the equilibrium ratio ANp defined by Eq. (10.24), and plotted in Fig. 10.30.

[Np(VT)] = „ [H*]3/2 [N03 ~31/2

[Np(V)J Np [HN02 ]1/J

The total neptunium concentration in the aqueous phase X]sjp is

xNp = [Np(Vl)] + [Np(V)] = [Np(Vl)] (l + 0°.28)

Because pentavalent neptunium is essentially inextractable, the neptunium concentration in the organic phase yNp is related to the distribution coefficient of hexavalent neptunium ^Nptvi) by

^Np =^nP(vi)[Np(V1)] (10.29)

The apparent equilibrium distribution coefficient of neptunium, D&pp, defined as

(10.30)

is then given by

n ________ _____________ ^Nptvi)_____________

app 1 + [HN02] 1/2/A’np[H+]3/2 [N03_]1,1

Figure 10.31, calculated [G12] from Eq. (10.31), £>Np(vi), and the observed equilibrium ratios of Fig. 10.30, shows the dependence of Z? app on temperature and the concentrations of HN02 and HN03. Figure 10.31 is strictly valid only in the absence of nitrates other than nitric acid and traces of neptunium. When uranyl nitrate is present at appreciable molarity хи, DNp(VTj is given by Eq. (10.26), and the apparent equilibrium distribution coefficient for neptunium may be estimated from

Dv is given in Figs. 10.13 and 10.15 and ANp in Fig. 10.30. Complete ionization of HN03 and U02(N03)2 is assumed.

Another type of gravity-flow, vertical contactor with a rotating axial shaft is the rotating disk contactor developed by the Shell Development Company [Rl, R2], shown schematically in Fig. 4.28. It consists of alternate annular stator disks attached to the outer shell and circular rotor disks attached to the rotating shaft. Rotation of the central shaft, at peripheral speeds up to 6 m/s, provides controlled dispersion of the two phases and sets up a toroidal flow pattern within each stator compartment. There are no settling chambers, and the two phases drift past each other in countercurrent flow.

Dimensions of contactors for which test data are available [R2] are given in Table 4.15.

When the 20-cm-diameter column was tested in hexone-acetic acid-water, a height equivalent to a theoretical stage as low as 10 cm was observed, with a combined flow rate of both phases of 1.0 cm3 /(s-cm2) of column cross-sectional area. The holdup time per equivalent theoretical stage is only 10 cm т 1 cm/s = 10 s.

The rotating disk contactor has had extensive application in petroleum refining and organic chemical separations. A modified version, with holes in the horizontal stator disks to promote countercurrent flow, has been used in the recovery of uranium from solutions used in cleaning process equipment [D4, L2].

The principal effluents carrying radioactive material from a uranium mill are the following:

1. Airborne effluents, carrying radon gas (222Rn) and radioactive dust particles

2. Liquid effluents, carrying water-soluble radionuclides

3. Solid effluents, in mill tailings

In present mills, radioactive liquid effluents are held in storage ponds with mill tailings and eventually evaporate to a solid.

The amounts of these radioactive effluents have been estimated [S2] for two model uranium mills, each with a capacity of 2000 short tons of ore containing 0.2 percent U3Oe/day. One mill uses a carbonate leach, sodium hydroxide precipitation flow sheet such as that described in Sec. 8.4. The other mill uses an acid-leach, amine extraction flow sheet such

as described in Secs. 8.5 and 8.6. Each mill is assumed to be associated with a storage pond and tailings pile in which 20 years of mill effluents have accumulated. Two alternative sites were studied for each mill type, one in New Mexico in an arid region with average wind speed of 7 mi/h and the other in Wyoming in a region with more vegetation and an average wind speed of 10 mi/h.

For each model mill a number of cases were examined, with progressively better retention of airborne radioactive effluents. The results to be summarized here are for case 1, at the New Mexico site. Case 1 represents 1975 practice, with least complete removal of airborne dust and no holdup of gaseous effluents to permit 3.8-day 222 Rn to decay.

Airborne effluents. Table 5.24 lists the annual emission rate of airborne radionuclides from the two types of model uranium mill, each after 20 years’ accumulation of tailings.

Yellow cake. Table 5.25 gives the percent of the uranium, thorium, and radium in the ore assumed [S2] to be recovered in the yellow cake uranium mill product, and the activity of the yellow cake due to the thorium and radium.

Tailings. The first two columns of Table 5.26 give the percent of the uranium, 230 Th, and 226 Ra and its daughters assumed [S2] to be recovered in the tailings sand and the tailings slime and liquid effluents for the acid-leach and carbonate-leach processes, and the calculated [S2] concentration of the principal radionuclides in the two classes of tailings. The third column gives the resulting nuclide concentration in the composite tailings. The fourth column gives the calculated total curies of each radionuclide in the tailings after 20 years of mill operation. The potential hazard from insecurely impounded tailings is suggested by the large amount of radioactivity.

Differences among uranium mills and the ores they process will cause the amounts of radioactivity in individual mills to vary considerably from the estimates given above.

Thorium concentrate produced by the processes described in Sec. 8 is too impure to be used as nuclear fuel. Especially objectionable impurities, which frequently are associated with thorium in its ores, are neutron-absorbing rare earths and uranium, the latter because it would dilute isotopically 233 U formed in thorium during subsequent neutron irradiation. The objective of thorium purification is removal of these and other impurities to concentrations below a few parts per million.

Solvent extraction with TBP has become the standard procedure for purifying thorium, just as for uranium. Processes used in different countries differ, however, in details such as the solvent used to dilute TBP, its concentration, and the means used to strip thorium and coextracted uranium from TBP. Table 6.20 summarizes the main features of processes used for purification of thorium on an industrial scale in the principal thorium-producing countries. Wylie [W5] gives more detail on early pilot-plant thorium-purification runs. Most of the published U. S. work on thorium purification on an industrial scale deals with irradiated thorium rather than natural; this will be described under the Thorex process, in Sec. 5 of Chap. 10.

Here, a summary will be given of Callow’s [C2] description of a process used in England for purifying thorium concentrate and separating it from associated uranium. Figure 6.8 shows relative flow rates and nitric acid and thorium concentrations in this process. Feed is a nitric acid solution of thorium concentrates containing about 200 g Th02/liter of nitrate, a smaller concentration of uranyl nitrate, and considerable amounts of nitrates of other metals, such as iron and rare earths (RE). In the first contacting unit, consisting of five extracting stages and five scrubbing stages, one volume of feed is extracted with four volumes of recycle solvent, 40 v/o TBP in kerosine. At the 4 A nitric acid concentration of the feed, this solvent extracts effectively all of the uranium and thorium in the feed and a little of the associated impurities. Counterflow of 0.8 volume of 4 A HN03 in the scrubbing section removes these impurities from the solvent. Rare-earth content of extracted thorium is less than 5 ppm if the rare earth-to-thorium ratio of feed is less than 1:4.

In the second contacting unit, thorium is stripped from the rich solvent by 0.1 A HN03; uranium is scrubbed from the thorium product by additional solvent.

Uranium in solvent leaving the second contacting unit is stripped into an aqueous phase by 5% sodium carbonate solution. Stripped solvent is washed and reacidified with 4 N nitric acid for recycle to the process.

At the high thorium and nitric acid concentrations used in this flow sheet, two solvent phases may form, one rich in thorium and TBP and the other, lean. Callow [C2] states that formation of the two solvent phases does not interfere with operation of a mixer-settler cascade, whereas difficulty would be experienced with a pulse column. Conditions at which a

second solvent phase forms are sketched in Sec. 5 of Chap. 10. Distribution equilibrium data for thorium, nitric acid, uranium, and impurities are also given there.

If the plutonium recovered from discharge fuel by fuel reprocessing is stored for long periods, there is a loss of fuel value due to the radioactive decay of fissile 241 Pu. Even during storage periods as short as a few months, 241 Am, the beta-decay daughter of 241 Pu, builds up. Its decay is accompanied by gammas that increase the shielding fequired in the fabrication of fuel from recycled plutonium. Small quantities of 237U, formed by the alpha decay of 241 Pu, also increase the gamma activity. The decay of 2.85-year 236Pu forms 232U, 228Th, and short-lived decay daughters that also contribute to the shielding requirement. The growth of radioactive daughters in plutonium recovered from the fuel discharged each year by the uranium-fueled 1000-MWe LWR of Fig. 3.31 is shown in Fig. 8.6 [PI]. The radioactivity of the 228Th daughters, which will be in secular equilibrium with 228Th, is not included.

The properties of metallic americium are listed in Table 9.24.

Alternative processes for preparing metallic americium are the reduction of AmF3 with barium vapor in high vacuum at about 1300°C, reduction of AmF4 with calcium, and reduction of АтОг with lanthanum or thorium at about 1500°C in high vacuum. The vapor pressure of ameri­cium is much higher than that of lanthanum or thorium, so that pure americium is condensed in the colder parts of the apparatus [K2, L2]. Metallic americium dissolves readily in mineral acids.

Sources of data. Knowledge of distribution equilibria in Purex systems is useful in designing the solvent extraction contactors for a Purex reprocessing plant and in predicting the change in performance of an existing plant when operating conditions are changed. The first experimental measurements were those of Codding et al. [CIO] at the Knolls Atomic Power Laboratory. These were for an aqueous phase containing only water, uranyl nitrate, and nitric acid; an organic phase consisting of a 30 v/o solution of TBP in a commercial solvent (Gulf ВТ or Amsco 123-15); and for a temperature of 25°C. Distribution equilibria were represented graphically as plots of molarities of nitric acid and uranyl nitrate in the organic phase as functions of the corresponding molarities in the aqueous phase. After TBP became generally accepted as the preferred solvent for fuel reprocessing, many additional studies were made of distribution equilibria between aqueous nitric acid and TBP dissolved in a hydrocarbon diluent. These extended the early work to other hydrocarbon diluents, to temperatures other than 25°C, to TBP concentrations between 5 and 100 v/o, and to additional distributed components including plutonium, neptunium, and thorium.

It was found that distribution equilibria are not very sensitive to the composition of the hydrocarbon diluent, provided that it consists mostly of saturated (paraffinic or naphthenic) hydrocarbons containing about 12 carbon atoms per molecule. However, Purex plants now usually specify a synthetic, straight-chain, saturated hydrocarbon made by polymerizing and hydrogenating lower olefins, which contains an average of 12 carbon atoms and is mostly n-dodecane.

The SEPHIS+ computer program was developed by Gronier [G16] for Purex equilibria in 15 v/o TBP. The program was adapted to the conventional 30 percent TBP Purex process by Richardson at Hanford [R7], and was further modified and generalized by Watson and Rainey [W5] at Oak Ridge. The SEPHIS code predicts the equilibrium distribution of uranium, tetravalent plutonium, nitric acid, and water between an aqueous phase containing these components and an organic phase containing TBP at any concentration between 2.5 and 100

tSolvent Extracting Processes Having Interacting Solutes.

v/o, at temperatures between 0 and 70°C. The SEPHIS code may be used for uranium concentrations up to 2 M, Pu(IV) up to 0.2 M, and nitric acid up to about 6 M.

Scotten [S5] at Savannah River developed a similar program, SOL VEX.

Distribution coefficients. The equations used in the SEPHIS code to correlate distribution equilibria are too complex for hand calculation or for graphic representation in a few figures. To provide a semiquantitative basis for stage-to-stage calculation of the separation performance of Purex solvent extraction contactors described in Sec. 4.14, Figs. 10.13 through 10.16 have been plotted from computer printouts from the SEPHIS code kindly provided by Vaughen [VI]. These give distribution coefficients for nitric acid and uranyl nitrate at 40 and 55°C between an aqueous phase and 30 v/o TBP in normal dodecane.

Plutonium and fission-product and other inextractable nitrates are present in significant amounts in contactors HA and HS. Their effects on distribution coefficients of nitric acid and uranyl nitrate may be taken into account approximately by reading distribution coefficients of nitric acid or uranyl nitrate given in these figures at a value on the horizontal, x, axis equal to

the combined molarities of Pu(N03)4 and U0j(N03)2, from the curve whose designated nitric acid molarity equals the sum of the actual HN03 molarity plus the normality of inextractable nitrates. Problem 10.3 requires application of these adjustments.

Distribution coefficients for Pu(IV) may be obtained from Fig. 10.17. This plots the ratio of distribution coefficients of tetravalent plutonium to hexavalent uranium.

In the SEPHIS code [W5] this distribution coefficient ratio DpJDv is evaluated from

where F = volume fraction of TBP in dry solvent *no3′ = total nitrate molarity in aqueous phase

= 2дги + 4xpu + xH + xs (10.2)

Xy = uranyl ion molarity in aqueous phase xpu = molarity of Pu4+ in aqueous phase xh = hydrogen ion molarity in aqueous phase and Xg = molarity of N03" associated with other nitrates in aqueous phase

= *no3 ■ — 2xu — ^Xpu — -*h (10.3)

The coefficients of Xu, xPu, xH, and Xg are so nearly proportional to the charges of the respective cations that this equation may be simplified to

for F = 0.30 (30 v/o TBP). For uranium molarities under 0.5, the difference between Eq. (10.1) and Eq. (10.4) is less than 0.1 percent. Figure 10.17 is a plot of Eq. (10.4) for temperatures of 25, 40, and 55°C.

4.15 Example of Use of Purex Equilibrium Data

Use of these Purex equilibrium charts will be illustrated by calculating the number of equilibrium extracting and scrubbing stages needed in the uranium decontamination unit 2D of the Barnwell Nuclear Fuel Plant, data for which were given in Fig. 10.11 and Tables 10.7 and 10.8.

Table 10.10 gives flow rates and compositions for the streams to and from this unit. Process quantities taken directly from Tables 10.7 and 10.8, or calculated from them in

equivalent units, are in italics. The remaining quantities in Table 10.10 have been adjusted as stated in the footnotes for two reasons. (1) Component flow rates in some streams have been changed slightly to provide exact material balances. (2) Volume flow rates, which change by a few percent in Fig. 10.11, have been held constant to simplify calculation; compositions were adjusted where necessary to keep component flow rates unchanged.

Figure 10.18 shows the solvent extracting system to be analyzed and the nomenclature to be used. Input and output flow rates and concentrations are from Table 10.10. Three extracting and two scrubbing stages are shown, because the calculation next to be described indicates that between two and three theoretical extracting stages and between one and two scrubbing stages would be sufficient for the specified separation.

In this example, uranium and ruthenium are the key components whose compositions in feed, aqueous waste, and organic extract are specified. Nitric acid concentration in feed is specified, but its distribution between the two product streams must be found by trial, as in the zirconium-hafnium separation example in Sec. 6.5 of Chap. 4. The HN03 concentrations of 0.02 M in organic extract and 1.658 M in aqueous waste were found by trial to require the same number of scrubbing and extracting stages as the specified uranium and zirconium separation, as will now be shown, and hence represent the calculated distribution of nitric acid.

Table 10.11 gives steps in calculating concentrations of nitric acid, uranyl nitrate, and ruthenium as a function of stage number in the extracting section. Starting from the given aqueous concentrations xf, marked with a f, the calculation proceeds through alternative distribution-equilibrium and material-balance calculations. No iterations are required, as the distribution coefficients of uranyl nitrate and nitric acid are available in Figs. 10.13 and 10.14 as functions of the first calculated aqueous concentrations.

Table 10.12 gives the steps in calculating concentrations of these three components in the scrubbing section, starting from the specified composition of the organic extract stream vf, marked with a $. As the distribution coefficients of Figs. 10.13 and 10.14 are given as functions of the to-be-calculated aqueous composition, these must be found by the successive approximation procedure shown in the table. In this, distribution coefficients are assumed, trial aqueous concentrations are calculated, distribution coefficients are obtained from Figs. 10.13 and 10.14, and the process is repeated until calculated distribution coefficients are the same as assumed.

Figure 10.19 is a plot of the concentrations of ruthenium (bottom, circles) and nitric acid (top, squares) versus uranium concentration in the organic streams leaving the designated stages of the extracting section (filled symbols) and scrubbing section (open symbols). The inter­section of the two bottom lines shows that the specified ruthenium-uranium separation would be obtained at a value of n = 2.4 (TV = 2.4 theoretical extracting stages) and m = 2.1 (M = 1.1 theoretical scrubbing stages, because organic stream from stage m flows into scrubbing stage m— 1). The intersection of the two top lines shows that the assumed nitric acid-uranium separation would be obtained at the same values of m=2.1 and л = 2.4 thus establishing that the assumed nitric acid concentrations in aqueous waste and organic extract streams are correct. Concentrations at the intersections of the two curves are the calculated values for the organic stream flowing from the extracting to the scrubbing section.

“Quantities in italics evaluated from Tables 10.7 and 10.8. b Adjusted to close material balance.

“Adjusted to keep constant volume flow rate.

d Adjusted to keep mass flow rate in residue the same as evaluated from Table 10.7.

“Nitric acid distribution cannot be specified in advance, but is confirmed by calculation of Tables 10.11 and 10.12.

Liquid HLW is the concentrate of the aqueous raffinates from the reprocessing extraction cycles. This means that up to 1 percent of the uranium and plutonium and practically all of the

t For brevity, the Barnwell Nuclear Fuel Plant is referred in this chapter as the AGNS plant, an abbreviation for the plant owner, Allied General Nuclear Services.

fission products, neptunium, and the transplutonium elements produced in nuclear reactors end up in the HLW. It is therefore a reservoir of radioactivity that will not reach a steady state as long as nuclear power is generated, and its hazard potential will last much longer than the use of nuclear energy. Therefore, a reliable technology for long-term isolation of high-level wastes from the environment is a key to environmental protection against the consequences of nuclear power, and it is also a key to the public acceptance of nuclear power.

HLW arising in solid form, mainly cladding hulls, has activity concentrations more than two orders of magnitude lower than liquid HLW. It presents somewhat different technical problems, which will be discussed briefly in this chapter.

The density of uranium metal changes markedly with temperature. Table 5.6 summarizes densities inferred from x-ray diffraction data. The large density change at 662°C accompanying transition to the beta phase makes it undesirable to operate uranium metal reactor fuel above this temperature. Even at the lower temperatures at which the alpha phase is stable, its large density change with temperature, and its unequal temperature coefficient of thermal expansion along the three crystal axes (+19, —0.8, and +20 X 10’6/°C at 25°C) cause severe distortion and elongation of fuel assemblies during temperature cycling unless the fuel is given special treatment prior to irradiation.

Table 5.5 Phases of uranium metal

Transition

temperature, °С Phase Crystal system

Solid a Orthorhombic Solid jS Tetragonal

Solid 7 Cubic

Liquid Vapor