Oxidation states. In aqueous solution plutonium can exist in the five oxidation states Pu(III), Pu(IV), Pu(V), Pu(VI), and Pu(VII), which occur as hydrated ions Pu3+, Pu4+, Pu02+, Pu022+, and РиОг34. Although the most stable oxidation state in solution is usually considered to be Pu(IV), the differences in oxidation potentials of Pu(III) through Pu(VI) are small enough that all of the first four states can exist simultaneously in aqueous solution. Although Pu(V) is unstable to disproportionation, it is less so than pentavalent uranium. The Pu(IV) state also disproportion — ates to a measurable extent, in part due to radiolytic decomposition. Unless the oxidation poten­tial is controlled, solutions that contain other than the pure Pu(III) or pure Pu(VI) will react to form stable mixtures containing appreciable concentrations of all four states through Pu(VT). Pu(IV) usually predominates in nitric acid solutions free of oxidizing and reducing substances. In Purex reprocessing nitrite ion is added to the nitric acid-plutonium solution to oxidize Pu(III) to Pu(TV) and to reduce Pu(VI) to Pu(IV).

The heptavalent state Pu(VII) is observed when Pu(VT) in a basic solution is oxidized by ozone, persulfate, or electrolytically. The resulting Pu023+ is highly unstable and readily reverts to Pu(VI) in a less oxidizing environment or in acidic solutions.

In the Purex aqueous separation process [C6] plutonium is maintained as Pu(IV) for decon­tamination from fission products and as Pu(III) for partition from uranium. The stability of these oxidation states in nitrate solutions depends on rate phenomena, because the potentials of the nitrate-reduction couples N03′-N0(g-) and N03^-N204(?) are near those of the plutonium — oxidation couples Pu(III)-Pu(IV) and Pu(rV)-Pu(VI). The rate is in part determined by the con­centration of the nitrate-reduction products NO(g) and N204(g) and their rate of removal from solution. Experimental data summarized by Seaborg [SI ], show that Pu(IV) in nitric acid solu­tions at room temperature undergoes essentially no disproportionation or oxidation for several days. Addition of sulfuric acid to the solution further represses the oxidation. Data taken at temperatures of 75 to 100°C indicate no detectable oxidation of Pu(IV) in 14 M HN03, a half-life for Pu(IV) of about 4 h in 2 M HN03, and a half-life of only about 10 min for 0.25 M HN03. The high acid concentration represses the equilibrium of the hydrolytic oxidation of Pu(IV) to Puvi022+.

The disproportionation of Pu(IV) proceeds by means of two reactions, first the slow,

2PU4* + 2H2 О -* Puv02+ + Pu3* + 4H+ (9.41)

followed by rapid establishment of the equilibrium

Puv 02+ + Pu4+^Puvi022* + Pu34

Hie calculated equilibrium percentages of plutonium in the various oxidation states in 0.5 M HD at 25°C, assuming an average oxidation state of IV and assuming no complexing, are listed in Table 9.19.

The rate of the Pu(IV) disproportionation reaction (9.41) is proportional to the square of the concentration of Pu(IV), with a rate constant [SI] of 0.75 liter/(mobh) at 25°C in 0.481 M HC104. The rate is faster the lower the acid concentration and the higher the temperature.

The practical consequences of these findings is that plutonium may be stabilized in the Pu4* state by keeping the solution at 25°C, by having a high ЮГОэ concentration to retard reaction (9.41) and to complex the Pu44 ion, and by keeping the solution dilute in plutonium.

Seaborg [SI] states that the rate of oxidation of Ри(ПІ) to Pu(IV) is slow in dilute nitric acid at room temperature but proceeds rapidly in dilute nitric acid at 100°C and in concentrated nitric acid (16 M) at room temperature. The suggested mechanism is the oxidation of Ри(ІП) by nitrous acid HN02 formed by the reaction of dissolved NO with N03" or with N02.

Another complication in plutonium solution is the gradual, spontaneous reduction of Pu(VI) to Pu(IV), and Pu(IV) to Pu(III), caused by ionization products of alpha particles emitted in radioactive decay [SI]. The rate of alpha reduction is slow, however. For example, the observed rate of reduction of Pu(VI) in 0.5 M HC1 at 25°C is 0.0035 g-equiv/day per mole of plutonium, which corresponds to a half-life of 199 days for reduction of Pu(VI) to Pu(IV). From these rates and the known alpha-decay rate and decay energies of plutonium, it is estimated that approxi­mately 80 eV of dissipated alpha energy in this solution brings about the addition of one electron in reducing plutonium ions. After several hundred days the plutonium reaches an average oxida­tion state intermediate between Pu(III) and Pu(IV).

For Pu(IV) in 0.481 M HCIO4 at 25°C, the rate of spontaneous reduction is 0.00106 g-equiv/ day per mole of plutonium, corresponding to a half-life of 653 days for reduction of Pu4+ to Pu3*. To keep plutonium in the hexavalent and tetravalent states over long periods of time, it is neces­sary to have an oxidant present to reoxidize lower valence states as fast as they form.

Trivalent plutonium. Solutions of trivalent plutonium salts are generally similar to the trivalent rare-earths. Like the rare-earths, the hydroxide, fluoride, oxalate, and phosphate. are insoluble. Plutonium forms double sulfates with alkalis of the form MPu(S04)2 -4H20, again like the rare — earths. In the absence of air, aqueous solutions of trivalent plutonium salts are stable against hydrolysis; they are readily oxidized by air to the tetravalent form.

Tetravalent plutonium. Solutions of tetravalent plutonium salts are generally similar to tetravalent cerium and uranium. The fluoride PuF4, potassium complex fluoride K2PuF6, iodate Ри(ІОз)4, and phosphate Риз(Р04)4 are insoluble. Excess soluble hydroxides precipitate Pu(OH)4. The

hydroxide is easily converted to Pu02 by heating. When hydrogen peroxide is added to acid solu­tions of Pu34, Pu44, Pu024, and PuOj24, a Pu(IV) peroxide is precipitated. The precipitate compo­sition is variable, such as Pu04 (N03 )2;c • 2-3H2 O. Tetravalent plutonium forms chelate com­pounds with thenoyl trifluoracetone (TTA) or acetylacetone, which may be extracted from aqueous solution into benzene. Tetravelent plutonium nitrate is the form of plutonium most readily extractable by TBP.

Pentavalent plutonium. Pentavalent plutonium salts have only limited stability in aqueous solu­tion. At pH above 1, they begin to hydrolyze, and at lower pHs they tend to disproportionate to Pu02 24 and Pu44 or Pu34. All the common salts are soluble.

Hexavalent plutonium. Pu0224 in acid solution is a much stronger oxidizing agent than U022+. The two ions also differ in that the solubilities of plutonium are greater than those of the corre­sponding uranium compounds. In most other respects the two ions are similar. Plutonyl nitrate is very soluble in water and is extracted by methyl isobutyl ketone and other oxygenated organic solvents. Soluble hydroxides precipitate plutonates, such as Na2Pu04. These dissolve in sodium carbonate solution as complex carbonates. Plutonyl phosphate, arsenate, and double sodium acetate, NaPu02(CH3C02)3, are relatively insoluble.

Plutonium complexes. Plutonium ions form complexes with many anions. The most important of the complexes are those that form with Pu44, some of which are listed in Table 9.20, in order of increasing stability.

At sufficiently high concentrations of HN03 or HC1, e. g., > 2.5 M HC1, plutonium forms anionic complexes that are strongly sorbed by anion-exchange resins. Because the complexing ability to form anions varies with the plutonium oxidation state, which can be preferentially ad­justed with respect to the other actinides, anion exchange is useful in the separation of plutonium from other actinide elements and in the separation from cationic impurities that do not easily complex. Because of its high ionic potential, plutonium is also readily adsorbed onto cation — exchange resins. Elution of sorbed Pu(III) from such resins by means of dilute nitric acid, or of sorbed Pu(IV) by a complexing acid such as HC1, is a means of concentrating plutonium in solu­tion.

Plutonium readily complexes with organic complexing agents, such as TBP, according to the overall reactions

Pu(N03 )4 (aq) + 2TBP(o) ^ Pu(N03 )4 • 2TBP(o) (9.43)

Source: S. Peterson and R. G. Wymer, Chemistry in Nuclear Technology, Addison-Wesley, Reading, Mass., with permission.

and Pu02(N03Maq) + 2TBP(o)^PuC>2(NC>3)2-2TBP(o) (9.44)

Weaker complexes are formed by Pu(III) and Pu(V). The use of these extractable complexes in fuel reprocessing to separate uranium and plutonium from fission products and to partition plu­tonium from uranium is discussed in Chap. 10.

Hydrolysis. Hydrolysis is one of the most important reactions in the chemistry of plutonium in aqueous solutions. The tendency of plutonium ions to hydrolyze decreases in the order:

Pu4+ > Pu02 2+ > Pu3* > Pu02+

As would be expected from the relative sizes of ions listed in Table 9 .3, the plutonium ions in any oxidation state are more readily hydrolyzed than their larger neptunium and uranium analogues.

The hydrolysis of Pu3+ is not extensive, forming the hydrolyzed species PuOH2+ according to the reaction

Pu3+ + H2 О ^ PuOH2+ + H+ (9.45)

In solutions of lower acidity Pu(OH)3 precipitates, with a solubility product of about 2 X 10’20 [СЗ].

Tetravalent Pu4+ hydrolyzes more readily than any other plutonium species. In hydrogen ion concentrations of less than 0.3 M, the hydrolysis is initiated by the reversible reaction

Pu4+ + H20- PuOH3+ + H+ (9.46)

In Pu4+ solutions of low acidity the highly insoluble hydroxide Pu(OH)4 precipitates, with an estimated solubility product of 7 X 10"“ [C3].

The hydrolysis of Pu(V) is very slight, because of the low charge on the Puv02+ ion. The hydrolyzed Pu(V) rapidly disproportionates. Hydrolyzed monomers and polymers of Pu(VI) are also formed.

Polymers. In weakly acidic solutions the reversible hydrolysis may be followed by an irreversible formation of a colloidal product polymerized to a high molecular weight, quite similar to the hydrolytic behavior of Th4+ and U4+. These polymers consist of very small, discrete, amorphous particles, which invariably convert to the crystalline form when aged [L4]. The rate of formation of plutonium polymers is greater at high temperatures. Once formed, these plutonium polymers do not readily disperse or dissolve in solutions of acidities sufficiently high to have prevented their formation. The hydrolysis and polymerization of Pu(IV) is suppressed in a sufficiently acid solu­tion and in the presence of some complexing agents. The Pu(IV) polymer is destroyed only slowly by highly concentrated acid at room temperature, but it is destroyed rapidly at 90°C. Although the presence of fluoride ion appears to have little effect on the formation of Pu(IV) polymers, its presence does accelerate the depolymerization. The plutonium polymers cannot be extracted by organic complexing agents, such as TBP. The polymers can also form in the organic solvents.

The tendency toward Pu(IV) polymerization is of considerable practical importance in pro­cess operations involving plutonium solutions. Dilution of an acidic plutonium solution with water can result in polymerization in localized regions of low acidity, so plutonium solutions should be diluted instead with acid solutions. Polymerization can result from leaks of steam or water into plutonium solutions or by overheating during evaporation. Polymer formation can clog transfer lines, interfere with ion-exchange separations, cause emulsification in solvent extraction and ex­cessive foaming in evaporation, and can result in localized accumulation of plutonium that may create a criticality hazard [C3].

The hydrolytic chemistry of Pu4+ is important in that it affects the behavior and mobility of plutonium in the environment [A2] and in geologically isolated radioactive wastes that may be subjected to slow leaching by ground water. The absorption spectra of the Pu(IV) polymer is similar to that of the plutonium hydroxide precipitate Pu(0H)4 [L4]. Experimental data in Fig.

9.6 show that the solubility of Pu(OH)4 as a function of acidity (pH) is quite similar to the plu­tonium concentrations at which Pu(IV) polymer formation has been detected [R1 ]. The compari­son suggests that if the total plutonium concentration at any given pH falls above the line of Pu(OH)4 solubility, the solution will be saturated with respect to Pu(OH)4 or to the Pu(IV) polymer, so that these species can form. If the plutonium concentration falls below the Pu(OH)4 solubility line, the Pu(IV) polymer or precipitated Pu(0H)4 will be absent. In the environmental pH range of 4 to 8 the concentration of soluble and unpolymerized Pu(IV) is so low that the environmental chemistry may be governed by the possible oxidation or reduction to other valence states of plutonium, as well as by the presence of complexing agents.

The controlled formation of polymeric Pu(IV) is important to the sol-gel process for the production of spherical Pu02 particles. Ammonia is added to a nitric acid solution of Pu(IV) to precipitate Pu(OH)4, which is subsequently peptized at 50°C with dilute nitric acid to produce sols of 1 to З M Pu with a N037Pu ratio of 0.1 to 03. The sols, which remain stable over periods of a few months, are dispersed in a dehydrating organic solvent to form a gel, which is ignited to form spherical particles of PuOj [К2].

A well-designed Purex plant aims for as complete recycle of solvent as possible, to minimize costs of solvent makeup and disposal. Solvent from the uranium purification section usually contains so few contaminants or degradation products that it can be reused a number of times without cleanup. On the other hand, solvent that has processed solutions containing high activity of fission products and plutonium carries traces of these contaminants, uranium, nitric acid, dibutyl phosphate, and other radiolytic degradation products of TBP and dodecane. Uranium and plutonium should be recovered because of their value. Fission products should be removed to prevent product contamination in later cycles. Dibutyl phosphate should be removed because it forms strong complexes with tetravalent zirconium and plutonium that would impair ability of the solvent to reject zirconium and separate plutonium from uranium.

A typical solvent cleanup process for a 5 MT/day oxide fuel-reprocessing plant would be as follows.

Solvent from the first extraction cycle is transferred to a decanter of 5-m3 volume. Small amounts of water are separated and sent to waste evaporation. The solvent is then pumped to an interim storage tank of 50-m3 volume, which serves as feed tank for the wash columns.

The solvent is expected to contain approximately

Uranium: Plutonium: Fission products: Nitric acid:

It is washed successively with 0.01 M HN03, 0.2 M Na2 C03, 0.2 M NaOH, and 0.02 M HNO3. The operations are performed in mixer-settler batteries. The uranium content is reduced by a factor of 10, the plutonium content by a factor of 50, and the fission-product content by a factor of 2. Dibutyl phosphate is removed as the water-soluble sodium salt. If sodium ion in the wastes from solvent cleanup is objectionable, ammonium carbonate has been proposed as an alternative [G3].

The cleaned solvent is collected in a 20-m3 tank and, via sintered stainless steel filters, transferred to a larger storage tank for reuse. As the metal filters may become very radioactive, provision is made for back washing and remote replacement. Final solvent polishing by adsorption on anion-exchange resin has been found advantageous at Hanford [S4].

Solvent cleanup methods at the principal reprocessing plants have been summarized by Naylor [N2]. Methods used at Savannah River have been described by Orth et al. [013].

Next Page

FUEL REPROCESSING 489

This section contains examples of how limits on critical mass or dimensions can be increased from the single-parameter limits of Sec. 8.2 by limiting two or more parameters simultaneously. Many more examples than could be cited here are given in the references listed at the beginning of Sec. 8. When using multiparameter limits, the caution cited for single-parameter limits at the beginning of Sec. 8.2 is even more essential, because of the larger number of variables to be controlled.

Highly enriched uranium or plutonium. Figures 10.33, 10.34, and 10.35 show the relation between the safe diameter of a cylinder of infinite length or the safe thickness of a slab of

infinite extent and the concentration of 235 U (Fig. 10.33), 233 U (Fig. 10.34), or plutonium (>5 w/o ^Pu, Fig. 10.35) in water. In Figs. 10.33 and 10.34, curves are given for solutions of U02F2 in water up to its solubility limit and for homogeneous mixtures of uranium metal and water up to the density of the metal. The dimensions at the minima of the U02 F2 curves are lower than corresponding solution values in Table 10.25 because the figures have a greater safety factor than the table.

Figure 10.35 refers to a homogeneous mixture of Pu02 and water and is conservative for an aqueous solution of Pu(N03)4.

Curves are given for two thicknesses of water reflector. The 25-mm curves “generally provide a sufficient margin of subcriticality to compensate for water jackets around piping and for reflection by concrete 300 mm or more distant. Limits for a 300-mm-thick water reflector are appropriate when reflector conditions cannot be rigidly controlled” [T5, p. 40]. Lower limits are required when reflection is by close-fitting concrete, uranium, tungsten, beryllium, D20, or plastic.

Mixtures with 238 U. When 238 U is mixed with 238 U, the subcritical limit for a cylinder or slab of U02F2 solution given by Fig. 10.33 may be increased by the factor given in Fig. 10.36. These factors may also be applied to uniform slurries of water and U02 provided that the 23SU enrichment is greater than 6 w/o or the particle sizes are smaller than 127 ptm. With 235 U enrichment below 6 w/o and larger particles, the factors are smaller than given in Fig. 10.36 because of reduced absorption by 238 U. The factors are conservative for aqueous solutions of uranyl nitrate because of neutron absorption by nitrogen.

When plutonium dioxide is mixed uniformly with uranium dioxide containing 0.71 w/o or less 235 U, the subcritical limits for infinite water-reflected cylinders or slabs are greater than given in Table 10.25. Table 10.27 shows the dependence of critical dimensions on w/o Pu02 in Pu02 + U02 and on plutonium isotopic composition.

Soluble neutron absorbers. The preceding limits on critical concentration or dimensions can be greatly relaxed when soluble neutron absorbers, such as boric acid or gadolinium nitrate, are

t All values are upper limits except atomic ratios, which are lower limits.

■^Plutonium isotopic composition: I—^Pu > 241 Pu. II—^Pu > 15 w/o and 241 Pu < 6 w/o. Ill—^Pu > 25 w/o and 241 Pu < 15 w/o. The small quantities of 238 Pu and 242 Pu expected in these isotopic mixtures are considered to have neglible effects on the limits.

§This concentration limit is not applicable to oxide mixtures in which the Pu02/(Pu02 + U02) ratio is less than 3 w/o because of the increased re­lative importance of 23SU in high-uranium-bearing materials.

Source: Extracted from American National Standard ANSI/ANS-8.1 2-1978, with permission of the publisher, the American Nuclear Society.

assuredly uniformly distributed in the fissile material. As one example, one atom of natural boron per atom of 235 U will keep a large volume of aqueous solution subcritical for 235 U concentrations up to 400 g/liter. As another example, Fig. 10.37 shows how the subcritical diameter of an infinite cylinder of an aqueous solution of Pu(N03)2 is increased by addition of Gd(N03)3.

Solid neutron absorbers. In the disengaging sections of pulse columns and in storage vessels for solutions, it is sometimes desirable to have larger vessels than the maximums allowed in the preceding text. By packing such equipment with borosilicate glass Raschig rings, the maximum

Table 10.28 Maximum permissible concentrations of homogeneous solutions of fissile materials in vessels of unlimited size packed with borosilicate glass Raschig rings

Maximum concentration
in vessels with
minimum glass content of

Source: American Nuclear Society, “Proposed American National Standard, Use of Borosilicate-Glass Raschig Rings as a Neutron Absorber in Solutions of Fissile Material,” Report ANS-8.5-1979, La Grange Park, 111.

concentration of fissile materials that can be contained in indefinitely large vessels without becoming critical can be increased to the values given in Table 10.28. A proposed American National Standard [A7] gives specifications on the dimensions and composition of the rings.

Plutonium polymer. At low acidity and high temperature, plutonium forms a polymer that deposits as an insoluble solid film on the walls of process equipment. Polymer deposition plugs lines, fouls surfaces, and may result in unanticipated accumulation of a critical mass of plutonium. Figure 10.38 summarizes [М3] the results of investigations of the combinations of low acidity and high temperature that must be avoided if plutonium polymer formation is to be prevented.

As an additional precaution, process equipment in which plutonium polymer might form should be soaked periodically in boiling, concentrated nitric acid. If plutonium is found in solution, the presence of a polymer deposit is indicated. Complete removal may require addition of 0.01 to 0.1 M HF to the hot HN03.

Figure 5.2 shows the nuclear reactions that occur successively as 333U decays into its stable end product 301Pb. Table 5.3 gives the half-lives of these radioactive species and their principal decay radiations. The last column of Table 5.3 gives the ratio of the number of atoms of each nuclide to the number of uranium atoms in natural uranium, again assuming that decay equilibrium has been established. At equilibrium, the activity of each of these nuclides is the same. Per megagram of natural uranium contained in the ore, the activity of 333U and each of its daughterst is

(0,007205 333U/U)(106 gU/Mg)(6,0225 X 1033/g-atom)(0.693)

(238 g nat U/g-atom)(7.1 X 10® yr)(3.154X 107 s/yr)[3.7X 1010/(Ci-s)] (5.2)

This is only one-twenty-second of the activity of 338 U and each of its daughters. This small contribution to the activity of natural uranium will be disregarded in the remainder of this chapter.

In the Allied Chemical Company’s plant for converting uranium ore concentrates to UF6 at Metropolis, Illinois, ore concentrates are first converted to impure UF6, which is then purified by fractional distillation, through the steps shown in Fig. 5.23. The plant and processes have been described by Ruch et al. [Rl] and Sutton et al. [S5].

To minimize HF consumption and avoid formation of low-melting compounds of NaF and UF4, this plant prefers feed consisting of (NFLi)2 U207 or uranium oxide. Feed containing high concentrations of sodium or magnesium is converted to (NbLt^ U207 by reaction with hot (N114)2S04 solution, at an extra charge. (NH^UjC^ is converted to U03 by heating to 450°C. The U03 is reduced to U02 in a fluidized bed by reaction at 540 to 620° C with 1.5 times the stoichiometric amount of hydrogen, made by cracking ammonia gas. Although the reaction is exothermic, heat must be added to bring the reactants to the required temperature and to compensate for heat losses. Careful temperature control is necessary; too low a temperature leaves U03 unreduced, and too high a temperature causes sintering and loss of reactivity in the hydrofluorination step.

Conversion of U02 to UF4 by reaction with HF gas is carried out in a two-stage countercurrent fluidized-bed system as described in Sec. 9.5. This step removes as volatile
fluorides any silicon, sulfur, and boron present in the feed, and some of the molybdenum and vanadium. Off-gases are scrubbed first with water and then with aqueous KOH to remove these effluents and the excess HF used to complete conversion of U02 to UF4. The water scrub solution is neutralized with lime.

UF4 is converted to UF6 by reaction with fluorine at 425 to 535°C in an air-cooled, monel fluid-bed reactor charged with CaF2 diluent to improve heat transfer. Because of nonvolatile fluorides present in the crude UF4 feed, a small amount of CaF2 is continuously removed from the bed and processed for uranium recovery by reaction with fluorine in an ash cleanup reactor. Product gases from the primary reactor are passed through a cold trap to condense most of the UF6. Unreacted fluorine in off-gas from the cold trap is removed by reaction with UF4 in a fluorine cleanup reactor. Effluent passes through a filter, additional cold traps, and a KOH scrubber.

The UF6 condensed in the cold traps contains as possible impurities fluorides with the following normal boiling points, among others:

Because UF6 has a vapor pressure of 1 atm at 56°C, it is necessary to use two columns to purify it, one to remove low-boiling impurities and the other, high-boiling. Because the triple point of UF6 is 64°C and 1140 Torr, to prevent freeze-up it is necessary to operate both columns at a pressure over 1140 Torr with condenser cooling at a temperature over 64°C.

Figure 5.23 Allied Chemical UF6 process.

Most of the radioactive contaminants in the yellow cake fed to the plant leave in the residue from the fluorination reactor, where the nonvolatile fluorides of 226 Ra and thorium are concentrated. In addition, 234 Th and 234Pa, which form in UF6 from decay of 238 U, build up in bottoms of the UF6 distillation column (Prob. 5.4).

Uranium recovery at the Allied Chemical plant has exceeded 99.5 percent. UF6 product meets DOE specifications. The charge for converting concentrates to UF6 was around $3.50/kg uranium in 1977.

The advantage of the Allied Chemical dry process over the conventional process described in Secs. 9.2 through 9.6 is the smaller number of steps in the dry process, which makes its costs lower for certain kinds of concentrates, e. g., those not containing large amounts of Na, Ca, Mg, or Fe. Disadvantages of the process are the difficulty of handling large amounts of these impurities and its inability to produce pure U02 or pure UF4.

The grades of zirconium and zirconium-base alloys commercially available in the United States are described in ASTM Special Technical Publication 639 [S2]. The most important of these are zircaloy-2, zircaloy-4, and Zr-2.5 Nb. Table 7.4 gives the ASTM composition specifications for these three alloys and zirconium sponge.

The zircaloy series of alloys was developed by the U. S. Navy Nuclear Propulsion Program for service in the core of water-cooled nuclear reactors [R3]. Compared with pure zirconium, these alloys have greater strength and better resistance to corrosion by water or steam. Zircaloy-4 was developed later than zircaloy-2 and became the preferred material, because the nickel in zircaloy-2 promoted the absorption of hydrogen, leading to reduction in ductility.

Zr-2.5 Nb [zirconium alloyed with 2.5 w/o (weight percent) niobium] has better mechanical properties than zircaloy, but is corroded more rapidly by water containing oxygen, such as is found in boiling-water reactors. It was the material preferred in 1971 [El] for pressure tubes in Canadian pressurized-water reactors.

References [S2], [R3], and [H3] give more detailed information on zirconium alloys and their history and corrosion behavior.

The calculated elemental composition, radioactivity, and decay-heat rate for discharge fuel are shown in Table 8.7 for the uranium-fueled PWR (cf. Fig. 3.31), in Table 8.8 for the liquid-metal fast-breeder reactor (LMFBR) (cf. Fig. 3.34), and in Table 8.9 for the uranium-thorium-fueled HTGR (cf. Fig. 3.33). These quantities, expressed per unit mass of discharge fuel, are useful in the design of reprocessing operations. For the purpose of comparison, all quantities are calculated for 150 days of postirradiation cooling.

When expressed in terms of radioactivity per unit amount of energy produced, as in Table 8.1, there is little variation in the fission-product radioactivity and toxicity due to the different fuel-cycle options. However, the long-term actinide activity is considerably affected. The greater quantities of americium and curium resulting from plutonium recycle increase the amounts of all of the actinides and 226 Ra, which control the ingestion toxicity of wastes after the fission products have decayed. The resulting total ingestion toxicity for the 1000-MWe LWR operating with self-generated plutonium recycle is compared with that for uranium fueling in Fig. 8.15 [P2]. The greatest long-term ingestion toxicity results if the discharge fuel is not reprocessed, because all of the plutonium and uranium in the discharge fuel then contribute to the long-term radioactivity. The toxicity for the radioactive wastes from the uranium-plutonium fast-breeder fuel cycle is similar to that for self-generated plutonium recycle in the LWR.

The toxicity of the high-level wastes from a uranium-thorium HTGR fuel cycle is initially smaller, after the fission-product decay period of 600 years, because of the relatively small quantities of americium, curium, 239Pu, and 240Pu formed in this thorium fuel cycle. However, aifter about 100,000 years of isolation the theoretical ingestion toxicity of the wastes is governed by 226 Ra, formed by

these actinide reactions in uranium-thorium fuel result in a relatively large growth in the theoretical toxicity of the radioactive wastes after storage periods of a few hundred thousand years.

Following the long-term buildup and decay of 226 Ra, which peaks at about 200,000 years, the main contributor to the waste ingestion toxicity is 225 Ra, a daughter from the decay of 233 U:

The composition of irradiated fuel to be fed to a reprocessing plant varies widely. It depends on the composition of the fresh fuel charged to the reactor, the neutron spectrum in which the fuel is irradiated, the specific power or rate of heat generation in the fuel, the duration of irradiation, and the length of time the fuel is “cooled”—the interval between end of irradiation and start of reprocessing.

Table 8.7 gave an example of the type of irradiated fuel from a commercial nuclear power plant that will predominate in the feed to reprocessing plants. That table listed the composition of fuel resulting from irradiation of fuel initially containing 3.3 w/o (weight percent) 235U after irradiation in a pressurized-water reactor (PWR) at a specific power of 30 MW/Mg (metric ton, MT) to a burnup of 33 MWd/kg (33,000 MWd/MT), followed by cooling for 150 days. The table gave the grams of each element per megagram of fuel, the radioactivity of each element in curies, and the rate of energy production by each element in watts. Noteworthy are the large number of elements present; the intense radioactivity, totaling 4.3 million Сі/Mg; and the substantial decay power, more than 20 kW/Mg. Because almost 50 percent of the fission-

product energy production is in the form of gamma rays, spent fuel and radioactive wastes separated from it require massive shielding.

The nominal design period for storing fuel after irradiation before reprocessing is 150 days.’*’ This allows the radioactivity and decay power to decrease to the levels of of Table 8.7, high though they be. Other reasons for storing fuel this long are to permit all gaseous fission products except tritium, 8SKr, and 1591 to decay to inconsequential levels, to reduce the activity of 8-day 1311 to manageable levels, to complete decay of 2.35-day 239Np to 239Pu, and to complete decay of 6.75-day 237U. Some reprocessing plants require longer cooling periods; the proposed plant for oxide fuel at Windscale, England, requires that fuel be cooled for a year [B17] before reprocessing.

Jackson and Walser [Jl] have given a detailed description of the operations conducted in 1970 by the Atlantic Richfield Hanford Company to separate and decontaminate 233 U and its associated thorium from aluminum-clad thorium dioxide irradiated to low burnup in the Hanford reactors.

Dissolution. The aluminum cladding was dissolved in a solution of mixed sodium nitrate and sodium hydroxide, and the undissolved uranium, thorium, and fission-product oxides were separated by filtration and centrifugation. The oxides, together with some adherent sodium hydroxide, sodium nitrate, and sodium aluminate, were dissolved in a mixture of boiling 13 M nitric acid, aluminum nitrate, and potassium fluoride. From 16 to 48 h were required for dissolution. After dissolution was complete, the composition of the solution was as given in the first column of Table 10.17.

Feed adjustment. Dissolver solution was concentrated by evaporation till the boiling tempera­ture reached 135°C and the density 2.35 g/liter. This increased the thorium concentration to

3. M. One difficulty with this operation was volatilization of some of the ruthenium as Ru04. The solution was next made 0.20 M acid-deficient by steam stripping at constant volume. After cooling to 70°C, water was added to bring the solution to the desired composition of 1.5 M thorium and 0.10 M nitric acid-deficient for solvent extraction feed.

Feed composition was as given in the second column of Table 10.17.

Solvent extraction. The aqueous feed solution of nitrates was separated into decontaminated uranium product, decontaminated thorium product, and high-level fission-product waste by solvent extraction at 30°C with 30 v/o TBP in normal paraffin hydrocarbon diluent (лСю — иСі4), with controlled amounts of nitric acid used as salting agent. The flow sheet, described in detail by Jackson and Walser [Jl], used a codecontamination and partition cycle to separate feed into partially decontaminated uranium, partially decontaminated thorium, and high-level waste, as described in the following section. The uranium was purified by two additional cycles of solvent extraction followed by cation exchange. The thorium was purified by one additional cycle of solvent extraction. Overall separation performance and fission — product decontamination in the final group of runs, batches 2-1 to 2-44, is summarized in Table 10.18. A total of 285.5 short tons of thorium containing 452.6 kg of uranium was processed.

Codecontamination and partition cycle. Because the codecontamination and partition cycle is the critical step in the acid Thorex process, it will be described in more detail. In this cycle, shown in Fig. 10.21, most of the fission products were separated from the uranium and thorium, which were then separated from each other. The four solvent extraction units, HA, 1BX, IBS, and 1C, were pulse columns with dimensions given in the figure.

Feed HAF to the decontamination column HA had been adjusted to —0.1 M HN03 acid-deficient and 1.5 M Th(N03)4. It contained 0.61 g uranium/liter and the amounts of KF, A1(N03)3, NaN03, and KN03 shown in the figure. The KF had been added to catalyze

dissolution of Th02. The A1(N03)3 and NaN03 came from a residue of the Na3A103 produced by reaction of cladding with NaOH. The A1(N03)3 was left in the dissolver to complex fluoride ion and reduce corrosion of stainless steel.

In the extracting section of the HA column, uranium and thorium in the aqueous feed were transferred to the organic phase by HAS solvent, 30 v/o TBP in dodecane, flowing at a ratio of 940 volumes to 100 volumes of aqueous feed HAF plus 130 volumes of aqueous scrub HAS. The solvent reduced concentrations in high-level waste HAW to 0.29 g thorium/liter and <0.0003 g uranium/liter. These represented losses to waste of 0.2 percent thorium and <0.12 percent uranium. To keep these losses this low, it was necessary to raise the distribution coefficients of thorium and uranium by increasing the HN03 concentration in the column above the acid-deficient condition of the feed. HN03 addition was divided between 130 volumes of 1.0 M acid added at the top at HAS and 18 volumes of 13 M acid added near the bottom at HAS. The HAS scrub added at the top provided enough nitrate ion to drive uranium into the organic phase, without greatly increasing the distribution coefficients of fission products. Thorium distribution coefficients above the HAF feed point were also high enough to drive thorium there into the organic phase. Below the feed point, however, in the extracting section, as the Th(N03)4 concentration decreased, it was necessary to increase its distribution coefficient. This was done by adding the FLAX stream, 13 M in HN03, near the bottom of the extracting column.

By splitting the HN03 addition in this way, with high HN03 concentrations only near the bottom of the column where the Th(N03)4 concentration was low, formation of a second organic phase was avoided. Conditions for second-phase formation are shown in Fig. 10.27. This split feed of HN03 also reduced extraction of fission products.

In the scrubbing section of the HA column, fission products were scrubbed from the organic phase leaving the extraction section by the HAS scrub stream. It contained 0.01 M H3P04 to complex protactinium and zirconium-niobium and reduce their extraction. It also contained 0.01 M ferrous sulfamate to reduce plutonium and chromium corrosion product to inextractable species.

The HA column also processed 223 volumes of recycle solvent HAO containing low concentrations of thorium and uranium.

In the 1BX column thorium in solvent HAP leaving the FLA column was returned to the aqueous phase by stripping with 0.2 M FLN03 thorium strippant 1BX. This reduced the thorium/uranium ratio in the solvent 1BU leaving the column to about 1:5.5.

Uranium in the 1BU solvent stream was transferred to the aqueous phase in the uranium-stripping column 1C by back-extraction with 0.01 M HN03 and left this section of the plant as crude uranium product 1CU. Thorium was removed from this crude uranium in the second and third uranium cycles (not shown) and was returned to 1BXF feed in the 2BW and 3BW streams.

Uranium in the partially stripped thorium stream 1BXT leaving the 1BX column was extracted from the thorium in the IBS column by additional 30 v/o TBP in dodecane containing 0.01 M HN03. The aqueous stream leaving 1BX was the crude thorium product 1BT. The organic stream HAO leaving IBS contained some uranium and thorium and was recycled to the HA column.

Decontamination factors in this codecontamination and partition cycle were as follows:

Two additional uranium purification cycles (not shown) removed the thorium and small amounts of fission products remaining in the crude uranium stream ICU and returned the thorium to the first cycle in streams 2BW and 3BW.

The term non-high-level waste includes low — and medium-level wastes (LLW and MLW) and covers a very large range of wastes [Bl, С4]. Whereas the generation of HLW is determined by the quantity of fission products and transuranium elements inevitably generated in nuclear fission, that of non-high-level waste is rather dependent on specific process design and performance. It should be minimized in terms of volume as well as activity by appropriate process design. Recycling of waste streams in a reprocessing plant to reduce the volume of tritium waste is one example of how this can be done. A crucial point is the non-high-level alpha waste. The environmental benefit of the reprocessing fuel-cycle option depends largely on the minimization of this waste stream. Only if this minimization can be achieved will the long-term environmental impact of the fuel cycle be limited to a very small volume of solidified HLW with most of the plutonium eliminated by recycling.

The goal of non-high-level waste treatment is primarily volume reduction. This, however, does not hold for alpha-bearing waste, often called TRU waste. t Effective immobilization may

+ For transuranium.

generally be required for alpha waste. In contrast to HLW, recovery of actinides from alpha-bearing non-high-level waste may be beneficial. Because of the low fission-product concentration, actinide recovery will appreciably reduce the actual ingestion hazard of non-high — level waste. Moreover, it will be simpler from a technical point of view. As most of the actinide contamination in this waste will be plutonium, there may be even a certain economic compensation. Non-high-level waste may be classified into three categories:

1. Process waste (aqueous solutions, slurries, ion-exchange resins, organic liquids)

2. General trash (combustible and noncombustible trash)

3. Discarded equipment (contaminated or activated items)

There are three basic steps in treating these wastes, which may be applied in sequence or individually:

1. Volume reduction

2. Actinide recovery

3. Immobilization and packaging