237Np, a beta-emitting nuclide with a half-life of 2.14 X 106 years, is used as target material for production of 238Pu by irradiation with thermal neutrons in the reactions

237Np(n, у) 238Np 238Pu

2.1 day

238Pu is an important alpha-emitting radioactive energy source, which has been used extensively in space missions and in cardiac pacemakers. Its advantages for these applications are its relatively high specific power of about 0.5 W/g, its rather long half-life of 89.6 years, and the absence of appreciable gamma radiation, making heavy shielding unnecessary.

7.1 Sources of Neptunium

In thermal reactors 237Np is formed in the following reactions:

235 U(n, 7) 236U(n, 7) 237U <2^- 238U(n, 2n)

0" 6,75 days

237 Np

In fast reactors, the 238U(«,2n) reaction predominates. Typical concentrations of 237Np in irradiated fuel, from Chap. 8, are

Thermal reactors: 749 g/Mg HM

Fast reactors: 249 g/Mg HM

Neptunium concentration in fuel from thermal reactors could be increased by recycling uranium containing 236 U.

7.2 Oxidation-Reduction Equilibria in Neptunium Recovery

Figure 10.8 has shown that the neptunium ions Np4+ and Npvi02 2+ have sufficiently high distribution coefficients to be extractable by 30 v/o TBP in the Purex process. On the other hand, the distribution coefficient of NpvC>2+ is of the order of 0.001, so that pentavalent neptunium is essentially inextractable. The distribution of neptunium between aqueous and organic phases in the Purex process is thus determined by oxidation-reduction equilibria among the three valences of neptunium in the presence of the oxidizing or reducing agents used in that process. A semiquantitative indication of neptunium distribution among the three valence states is afforded by the comparison of their standard oxidation-reduction potentials with those of plutonium and possible reductants and oxidants, given in Table 10.21.

Comparison of the potentials for the two neptunium couples with those for plutonium

indicate that conditions that bring plutonium into the most extractable, tetravalent state used in the conventional Purex process will oxidize neptunium above the tetravalent state and leave most of it in the inextractable pentavalent state.

Because the distribution coefficient in TBP of hexavalent neptunium is higher than tetravalent, the hexavalent form is preferred for the first extraction from fission products. The first part of Table 10.22 gives equations for the concentration ratio of hexavalent to pentavalent neptunium calculated for the three oxidants listed there, with the coefficient evaluated from exp(—38.93 ДE°).

With nitrate ion or pentavalent vanadium, high neptunium concentration ratio requires a high ratio of oxidant to reductant and is favored by high hydrogen ion concentration. High oxidant ratio would tend to convert plutonium to the less extractable hexavalent state, but this tendency is inhibited by high hydrogen ion concentration and by the complexing of tetravalent plutonium with nitrate ion. At nitric acid concentrations of 2.5 or З M, 90 percent or more of the neptunium can be extracted as Np(VI) without converting more than a few percent of plutonium to the hexavalent state.

Strong oxidants, such as tetravalent cerium, would make neptunium almost completely hexavalent but would make plutonium hexavalent also and would volatilize substantial amounts of ruthenium as Ru04. For these reasons, strong oxidants are not favored when extracting hexavalent neptunium.

Because of solution nonidealities, observed concentration ratios differ significantly from ratios calculated from the standard oxidation-reduction potentials in Table 10.22. Figure 10.30 shows equilibrium ratios

[Np(VI)] [HNQ2 ]1/2 Np [Np(V)] [H+]3/2 [N03~] U2

for oxidation of Np(V) by nitric acid observed by Gourisse [Gil, 12] at 25, 35, and 50°C and

compares them with the coefficient 0.00048 in Eq. (10.18). The kinetics of this reaction are discussed in Sec. 7.5.

The second part of Table 10.22 gives equations for the concentration ratio of tetravalent to pentavalent neptunium calculated for the three reductants listed there. In the older Purex plants the ferrous sulfamate used to reduce plutonium to inextractable Pu3+ reduced neptunium partly to inextractable Np(V) and partly to extractable Np(IV). The reductants now preferred, tetravalent uranium (possibly made electrolytically) or hydroxylamine, are sufficiently strong, in sufficient time, to make neptunium almost completely tetravalent, but the reactions are much slower than reduction of tetravalent plutonium, because of slow deoxidation of the Np02+ radical. Kinetics of these reductions are also discussed in Sec. 7.5.

In the rotary annular contactor [D2, L2, T2] shown schematically in Fig. 4.27 [T2], the organic and aqueous phases flow countercurrently by gravity in the annular space between a rotating inner cylinder and a stationary outer cylinder. Taylor-instability vortices generated in the annulus promote dispersion and interfacial area. This is one of the simplest of the mechanically agitated contactors, and it has been developed for possible application to fuel reprocessing. In laboratory extractions of uranium from nitric acid with TBP in kerosene, Davis [D2] obtained values as low as 7.5 cm for the column height equivalent to a theoretical stage. The rotor speed varied from 1200 to 2000 r/min, with annular widths of 0.1 to 0.35 cm and a stator diameter of 2.2 cm. The residence time per theoretical stage was 10 s or less.

* Maximum critically safe size for highly enriched 235 U.

* Stainless steel contains a “poison,” such as gadolinium or boron, sufficient to absorb all neutrons entering the steel.

® Estimated for a Purex TBP solvent, with a maximum density difference between phases of 0.25 g/cm3.

Source: Adapted from M. W. Davis and A. S. Jennings, “Equipment for Processing by Solvent Extraction,” in Chemical Processing of Reactor Fuels, J. F. Flagg (ed.), Academic, New York, 1961, by permission.

Despite the very low concentration of uranium in seawater, 3.34 mg/m3, the large total amount in the world’s oceans, around 4X 109 MT of uranium [Dl], has provided incentive for study of means for extracting uranium from this ubiquitous source. To produce 1 MT (1 Mg) of uranium requires processing 106/3.34X 1СГ3 =300 million m3 of seawater. This enormous volume gives rise to the principal problems in extracting uranium from seawater. These problems are (1) providing at low cost a continual supply of feed water undiluted by depleted

і Aqueous product, 231/min,_____

ЗЭдСЩТ/

Aqueous (NH4)2 S04 strippant, 23//min____________ j

Figure 5.19 Eluex solvent extraction and yellow cake precipitation at Federal-American Partners’ uranium mill.

water from which uranium has been extracted, (2) preventing fouling of equipment by contaminants dissolved or suspended in seawater, (3) minimizing energy input, (4) preventing loss of reagents through solution or entrainment in reject water or reaction with it, and (5) recovering uranium selectively in the presence of millions of times greater concentrations of other substances.

In a December 1974 report, Battelle Pacific Northwest Laboratory [B2] summarized the principal processes that had been proposed for extracting uranium from seawater and gave references to more detailed descriptions of these processes. That report concluded that the most promising process was the selective adsorption of uranium from seawater on hydrous titanium oxide (titania).

Uranium is present in seawater as the very stable uranyl tricarbonate anion, which is thought to react with titania as follows:

U02(C03),4-(®?) + TiO(OH)2(s) -*• TiOUO40) + 2HC03~{aq) + C032-(«?)

Three important advantages of this reaction are the following: (1) It proceeds nearly to completion at the hydrogen ion concentration of seawater (pH = 8), so that chemical preconditioning of the large volumes of water is not required; (2) titania has very low solubility
in seawater, so that solution losses of this relatively expensive reagent (around $3/kg Ti) should be low; (3) it has a fairly high absorptive capacity for uranium from seawater; a value of 240 mg U/kg*Ti is considered representative [B2, D1 ].

This process was first publicized by Dr. R. Spence of the United Kingdom Atomic Energy Authority (UKAEA) at the 1964 United Nations Geneva Conference on the Peaceful Uses of Atomic Energy and was described briefly in reference [D1 ]. More details of the proposed process were given by British workers in references [K2] and [D2]. These workers made preliminary civil and chemical engineering designs of a plant located on the Menai Straits in the west of England where the mean tide amplitude of 5 m and the local coast configuration is favorable to construction of the 20 km of dams and sea gates needed to provide the desired once-through tidal flow of seawater through the absorption beds. The initial conclusion of these workers was that 840 MT of uranium/year could be recovered in this plant for a total cost of from $11 to $22/lb U308, assuming 80 percent recovery of uranium put through the plant. Later British experimental data indicated that the recovery in the titania absorbers originally proposed would be only 46 percent on the first cycle and would drop to 23 percent after eight cycles, with a cost increase to $26 to $42/lb U308 [Kl]. A critique of the original design of this plant (assuming 80 percent uranium recovery) and more complete process engineering and cost estimation by Oak Ridge National Laboratory [H2] indicated that the cost in 1966 dollars of an optimized plant at the Menai Straits site producing uranium at the rate of 430 MT/year would be around $1.5 billion, and that the cost of uranium produced in it would be around $260/lb U3 08. When the reduced capacity found in reference [Kl] and cost inflation since 1966 are taken into account, it seems likely that extraction of uranium from seawater would cost of the order of $500/lb U308.

The principal steps in the process proposed by the UKAEA for recovery of uranium from seawater are shown in Fig. 5.20. The titania recovery system consists of 60 beds, 1.3 ft (0.4 m) deep, each with a flow area of 188,000 ft2 (17,500 m2), filled with hydrous titanium oxide supported on an inert carrier. The inventory of the entire system is 71 million lb (32.2 million kg) of Ti, valued at $71 million in 1966.

Ti0U04(s) + (NH4)2C03(aq) + 2NH4HC03(aq) ->• TiO(OH)2(s) + U02(C03)34- + 4NH4+

Every 2 h the ammonium carbonate feed point is changed so that the bed is moved in flow sequence toward the ammonium carbonate feed point. After six phase changes the bed is being fed with fresh ammonium carbonate solution and has been stripped of uranium.

In the final five wash phases the bed is connected in series with the four other beds that preceded it in flow sequence and washed with water in countercurrent flow sequence to recover ammonium carbonate and prepare the bed to receive seawater at the start of the next cycle.

To provide steady flow of seawater through the recovery beds during the twice-daily change of tides, an elaborate system of tidal basins, dikes, and sea gates is required, responsible for two-thirds of the plant’s high cost.

Ammonium carbonate in the eluate is removed by steam stripping and then recycled. Uranium in the steam-stripped eluate is concentrated further by conventional anion exchange.

Solvent extraction has been used commercially for recovery of thorium from minerals other than monazite, in which complexing by phosphate is not a problem. Braun et al. [B5] describe the combined extraction of thorium and uranium from nitric acid solution of uranothorianite
ore, (U, Th)02, by 33 v/o (volume percent) TBP at the Le Bouchet plant of the French Commissariat і l’Energie Atomique.

Williams [W3] reported that from 1959 to 1968 thorium was recovered as a by-product of Rio Algom’s uranium mill at Elliot Lake, Ontario. Thorium-bearing uraninite ore was dissolved in sulfuric acid. Uranium was first recovered by anion exchange. Then thorium was recovered from the acid solution by solvent extraction with “alkyl phosphoric acid,” probably di(2-ethyl — hexyl)phosphoric acid.

Crouse and Brown [C5] describe pilot-plant studies on recovery of uranium and thorium from Canadian uraninite by sulfuric acid leaching followed by solvent extraction in a two-cycle amine extraction process using trioctylamine to extract uranium and di(tridecyl)amine to extract thorium.

There are several reasons why it is useful to store or “cool” irradiated uranium fuel for several months prior to shipment for reprocessing:

Table 8.5 Actinide quantities in discharge fuel with plutonium recycle^

Core and

Core, Blankets blankets,

Radionuclide kg/yr Ci/yr kg/yr kg/yr Ci/yr

1. Decay of fission-product activity and heat generation simplifies fuel shipment, and the lower activity reduces radiation damage to the organic solvents used in fuel reprocessing.

2. The decay of 5.27-day 133Xe leaves 85Kr as the only radioactive noble gas liberated in fuel reprocessing.

Preprocessing cooling is useful for iodine decay until the 1311 activity has decayed to a level equal to the activity of 1291. The time Tc t at which these two activities become equal can be calculated by applying Eq. (8.3) for 1311 and Eq. (8.6) for 129I, with the simplification that for 129I, Ty2 > Tc. Using the yield data for 235U fission given in Table 2.9, we obtain

^ ,, , , 8.98 X 107 „ ,n n4

Tc j = 11.6 In——- ——— days (8.8)

’ Jr

where the fuel irradiation time TR is in years. Assuming a typical TR of 3 years,

Tc’ і = 200 days

This is the length of time such that further cooling produces no appreciable reduction in the iodine activity. Shorter cooling times are possible for aqueous reprocessing, because it is not necessary to reduce the 1311 activity to quite as low a level as the 1291 activity.

A common specification of the permissible activity remaining in separated and decon­taminated uranium is that the specific beta activity not exceed that of natural uranium in equilibrium with its short-lived decay products 234 Th, 234mPa, 234 Pa, and 231 Th. These activities are

/3 0.68 дСі/g

a 0.69 jLtCi/g

These specific activities correspond to 1.5 X 106 beta disintegrations^min’g uranium). This is rounded off to the specification of 106 disintegrations/(min-g uranium) as the allowable 231U activity in uranium to be recovered and recycled to isotope separation. The actual allowable 237U content must depend on the amount of material to be handled and the allowable dose rate to operating personnel. 237U activity at this level of 106 disintegrations/(min• g) would result in a radiation dose on the surface of uranium metal at the rate of 2.6 mrem/h. This is less than 9 percent of the surface dose due to gammas in normal uranium and is a safe level for direct-contact handling of uranium.

The required cooling time TCpu for 237U decay can be determined if the atoms of 237U per atom of uranium N22(Tr)/Nj at the end of the irradiation period are known^:

X (60 s/min) = 106 disintegrations/(min’g U)

_ 1 , 3.61 X 1019ЛГ27(7к)Х27

Гс’и — хГ7 Ь where Aj is the average atomic weight of the isotopic mixture of uranium in the reactor product and X27 is the decay constant of 237U.

The concentration N21(TR)/Nu depends on the 236U concentration N26(Tr)INj at the end of the irradiation. Because of its relatively short half-life, 237U will be in secular equilibrium with 236 U, and its concentration is obtained from

tThe notation for nuclides is the same as that used in Chap. 3 and is defined under Nomenclature at the end of that chapter.

X;7A^27(7r) _ N26 {Tr)o 26 Ф Nv Nv

where 0 is the neutron flux at the end of the irradiation. Eliminating N2i(Tr)/Nj from Eqs. (8.10) and (8.11), we obtain

The concentration N2t(TR)/Nu of 236U can be obtained by applying the equations of Chap. 3. For the FWR example of Fig. 3.31:

Ng (TR) 3.85

957

The neutron flux to which the fuel is exposed is 3.5 X 1013n/(cm2-s). The effective absorp­tion cross section^ for 236 U is 123.9 b for this reactor, and the average atomic weight of uranium in the reactor product is 238. Using the above data in Eq. (8.12), the required decay time is

TCf и = 145 days

If there were sufficient incentive to reduce the fuel-cycle inventory of plutonium, it would be possible to operate with shorter preprocessing cooling times and to take the remaining 237 U decay time after the plutonium-uranium separation. In the fast-breeder fuel cycle, where there is usually the greatest incentive to reduce fuel cycle fissile inventory and thereby to reduce the fissile doubling time, the 237U content of the recovered uranium need not be as low as 106 disintegrations/(min-g)> because the uranium is not to be recycled to isotope separation.

5.1 Americium Isotopes

Table 9.23 lists the isotopes of americium important in nuclear technology and some of their important nuclear properties.

241 Am. The Sotope 241 Am is formed by the decay of 241 Pu. It undergoes alpha decay, with a half-life of 458 years, to form 237 Np. Isotopically pure241 Am can be extracted from aged reactor- grade plutonium. Irradiation of separated 241 Am is the basis of technology to produce gram quantities of 242 Cm. This is also one route to the production of the transcurium elements. How­ever, the first neutron-capture products are 242 Am and 242m Am, which have appreciable fission cross sections and which result in considerable heat evolution as contrasted to the production of transcurium isotopes from the irradiation of plutonium rich in the isotope 242 Pu.

241 Am will grow with time in plutonium recycled for fabricating reactor fuel, as discussed in Chap. 8. The gamma radiation accompanying the decay of 241 Am will contribute external radiation and may require personnel protection.

241 Am is a source of alpha particles in neutron sources for laboratory experiments and for reactor start-up. It is alloyed with beryllium to form AmBe13, which produces high-energy neu­trons by (a, n) reactions.

242 Am. The isotope 242 Am is the 16.0-h beta-ЕС emitter formed in 83.8 percent of the neutron captures in 241 Am.

24201 Am. The isotope 24201 Am is the long-lived (152-year) isomeric state that results from 16.2 percent of the neutron captures in 241 Am. It is one of the sources of the 4n + 2 decay series in high-level wastes from reprocessing irradiated uranium or uranium-plutonium fuel.

243 Am. The isotope 243 Am is the 7950-year alpha emitter resulting from neutron capture in 242 Pu and, to a less extent, from neutron capture in 24201 Am. It is an intermediate in the production of

Reaction with 2200 m/s
neutrons

transcurium elements by the long-term irradiation of plutonium. In the U. S. process 239Pu is converted to 242Pu, 243 Am, and 244Cm by extensive irradiation in neutron fluxes of 3 to 5 X 1014 nj{cm2-s) for about 18 months, resulting in a yield of about 60 g 242 Pu, 30 g of 243 Am and 244 Cm, and 910 g of fission products per kilogram of initial 239Pu. The remaining actinides are extracted from the fission products and are then further irradiated in the high-flux isotope reactor (HFIR), which was especially constructed at the Oak Ridge National Laboratory to produce transcurium elements. The overall production is about 0.2 g of 2S2Cf per year, with a yield of about 0.1 to 03 percent of the original 239Pu [K2].

243 Am is also important as a source of 239 Pu in the high-level wastes from fuel reprocessing.

244Am. The isotope 244 Am is the 10.1-h beta emitter formed by neutron capture in 243 Am. It is an intermediate in the nuclide chain leading to 244 Cm and thence to the transcurium elements.

As an example of a nuclear fuel reprocessing plant that makes use of some of the newest design concepts, a brief description will be given of the Barnwell Nuclear Fuel Plant, built and owned by Allied-General Nuclear Services (AGNS) at Barnwell, South Carolina. Construction of this plant, with the exception of its plutonium-conversion and waste solidification facilities, had been practically completed in 1977, when work was halted by President Carter’s decision to suspend indefinitely commercial reprocessing in the United States. Allied-General is jointly owned by the Allied Chemical Corporation and the General Atomic Company, the latter jointly owned by the Gulf Oil Corporation and a U. S. affiliate of the Royal Dutch/Shell Group of Companies.

The Barnwell plant is designed to process fuel from commercial PWRs and BWRs. It will process routinely fuel having no more than 3.5 percent 235 U (or equivalent plutonium) prior to irradiation, On an annual basis, the average burnup is expected to be less than 35,000 MWd/MT and the average specific power less than 40 MW/MT. However, the design will permit the plant to process fuel containing up to 5 percent 235U before irradiation, with bumups reaching

40,0 MWd/MT at a specific power of 50 MW/MT. Such high-enrichment batches will be handled with special techniques, including higher concentrations of soluble poison in the dissolver. Fuel will be cooled a minimum of 160 days before reprocessing.

Flow sheet and material quantities. Figure 10.11 is a schematic flow sheet showing the principal components of the reprocessing sections of the Barnwell plant [A3]. Table 10.7 [B21, M10] gives the flow rates and concentrations of the principal plant streams for the high-enrichment, high-bumup case. In Table 10.7 stream numbers in the first column correspond with stream numbers used in Fig. 10.11. The second column gives the stream designations used in AGNS reports [A2, A3]. Table 10.8 gives the fission-product content of the most important streams of Fig. 10.11. These material quantities were kindly provided by AGNS [В21.М10].

Performance specifications. The Barnwell plant is designed to recover at least 98.5 percent of the plutonium and 98.5 percent of the uranium in the plant feed [M10].

(See footnotes on page 495.)

tData from Buckham [B21] and Murbach [Mil],

^Feed enrichment, 5 w/o 235 U; specific power, 50 MW/MT; burnup, 40,000 MWd/MT. A, aqueous; O, organic 30 v/o TBP in dodecane; S, solvent, do — decane; V, vapor; HAN, hydroxylamine nitrate; FP, fission products; [0] concentration not stated, but known to be small (assume zero); ( ), calculated by material balance.

Data from Buckham [B21] and Murbach [M10].

To convert to hourly basis, multiply by 208 kg uranium fed/h.

Specifications for uranyl nitrate product call for a total beta-gamma activity less than 200 percent that of aged natural uranium if no less than 75 percent of the activity is due to ruthenium-rhodium; otherwise, a total beta-gamma activity less than 100 percent that of aged natural uranium, which is 0.68 /гСі/g uranium (Chap. 8). Thus, uranium product betters this specification. The specified alpha activity of uranium product due to transuranium elements is less than 1500 disintegrations/min per g uranium. This is roughly equivalent to a plutonium content of 10 ppb (10~8).

Specifications for plutonium product call for less than 100 ppm of uranium, a total gamma activity less than 40 мСі/g plutonium, and a zirconium-niobium activity less than 5 дСі/g plutonium. Plutonium nitrate product 3PC, stream 55, meets the total activity specification but does not quite meet the zirconium-niobium specification with the high burnup, 40,000 MWd/MT, feed used in this process example.

The plant is designed to discharge no liquid radioactive effluents to the environment [А2].

Gaseous effluents will be processed for the maximum practicable removal of radioactivity. Annual radiation exposure with the plant operating at capacity is expected to be 4.1 mrem maximum whole-body dose at the plant boundary and 116 man-rem to the entire population within 50 mi of the plant [А2]. These figures represent only 3 and 0.12 percent, respectively, of the average exposure from natural radiation.

Process building. The process steps shown in Fig. 10.11 are carried out in the main process building, a reinforced concrete structure with overall dimensions of 105 m by 87 m by 27.4 m high. Inside are the process cells, special heavily shielded enclosures for handling highly radioactive materials, and the surrounding work aisles and support facilities. The central control room is located on the second floor and is shielded and provided with special ventilation to permit occupancy during any radiation emergency.

Equipment within the high-radiation area that may experience mechanical or electrical failure or is subject to severe corrosion can be replaced remotely; the rest of this equipment is designed for direct maintenance after cleanout and decontamination.

Decladding, dissolving, and feed preparation. The plant will use the shear-leach method of feed preparation. Fuel elements up to 6 m long and 0.3 by 0.3 m in cross section can be sheared into 2.5- to 12.5-cm lengths.

Sheared fuel is fed to a receiving basket held in one compartment of a three-compartment semicontinuous dissolver. Nitric acid is fed continuously to the compartment and leach liquor is discharged continuously from it to the feed adjustment tank. When the basket is filled with fuel, the compartment holding it is rotated to a second position in which dissolution of the remaining oxide fuel in additional acid is completed. The basket and fuel hulls are finally rotated to a third position where water washes the residual leach liquor into the feed adjustment tank. The nitric acid contains 5.6 g of gadolinium as nitrate, to prevent criticality.

In the feed-adjustment tank plutonium is brought to the tetravalent state by addition of N204, if needed, and the uranium and nitric acid molarities are brought to about 1.2 M and 2.44 M, respectively, by addition of water and nitric acid as required. Samples are taken for input material accounting.

A centrifuge removes suspended solids from the dissolver solution ahead of the HAF feed tank.

Codecontamination. HAF product, stream ЗА, is fed continuously to the codecontamination section of the solvent extraction system. This consists of the 10-stage HA centrifuge contactor, which serves as the extracting section, and the HS pulse column, which serves as the scrubbing section. In the HA contactor solvent HAX, stream 4, extracts more than 99 percent of the uranium and plutonium from the feed and less than 1 percent of the fission products. More than 99 percent of the fission products from the feed and the gadolinium leave the solvent extraction plant in high-level waste FLAW, stream 5. In the scrubbing section HS, З M nitric acid scrub HSS, stream 7, removes about 70 percent of the residual fission products from HAP, the extract stream 6 leaving the extracting section.

The novel item in this part of the plant is the HA centrifugal contactor, named the Robatel, after its inventor. It was developed and built by the French firm Saint-Gobain Techniques Nouvelles. It consists of 10 centrifugal contacting stages, each similar in principle to the single stage described in Sec. 7.4 of Chap. 4, stacked vertically on a single rotating shaft. More information on this device has been given by Bebbington [B7] and Tarnero and Dollfus [Т2].

Partitioning. Plutonium nitrate and uranyl nitrate in the extract HSP stream 8 leaving the HS column are separated in the IB electrocell and pulse column. To this end, plutonium is reduced from the organic-soluble tetravalent state to the organic-insoluble trivalent state in the electrocell and returned to the aqueous phase IBP, stream 28, by the strip solution 1BX, stream 10.

The novel feature of the partition section is the electrocell, an electrolytic pulse column patented by Gray, Schneider, Cermak, and Ayers [G13] of AGNS. One concept of the electrocell column is shown schematically in vertical section in Fig. 10.12. A porous, electrolytically conducting alundum tube D separates the outer annular anolyte compartment G from the inner cylindrical catholyte compartment C. Across the inner compartment are placed at regular intervals perforated plates F which enable this compartment to be operated as a pulse column.

The outer compartment G is provided with a cylindrical anode E made of corrosion — resistant metal wire mesh. The anode compartment is filled with aqueous nitric acid, around З M.

The inner compartment C is provided with a set of disk-shaped cathodes В made of similar mesh. These disks are spaced between the perforated plates. They are mounted on a cathode center post A, which also supports the perforated plates through insulators. The inner cathode compartment C contains a dispersion of the organic phase flowing up through a down-flowing continuous aqueous phase, as in a conventional pulse column.

The organic phase is a solution of uranyl nitrate and tetravalent plutonium nitrate in 30 v/o TBP. The aqueous phase is a solution of uranyl nitrate and tetravalent and trivalent

plutonium nitrates in nitric acid. When an electric current flows from anode to cathode, the following net reactions take place:

In the outer, anode compartment, OH" -> |H20 + 502 + e~

In the inner, cathode compartment, Pu4+ + e~ -* Pu3+

As the organic phase rises through the inner compartment, tetravalent plutonium is extracted from it by the counterflowing aqueous phase. There the tetravalent plutonium is reduced electrolytically to the organic-insoluble trivalent form. In this way, the organic phase leaving the top of the electrocell column becomes stripped of plutonium, and the aqueous phase IBP, stream 28, leaving the bottom of the column carries practically all the plutonium in the feed. This aqueous effluent contains less than 2 percent of the uranium in the feed because of its relatively high, 2.6 M, nitric acid content and the high, 7:1, organic-to-aqueous flow ratio.

In the IB column remaining traces of plutonium are stripped from the solvent by a strippant 1BX, stream 10, containing hydrazine as holding reductant. A decontamination factor of 200 for removal of plutonium from uranium is anticipated for the IB columns. The big advantage of this partitioning system is that it adds no nonvolatile materials such as ferrous sulfamate to the system.

The uranium in the 1BU extract stream 11 is returned to the aqueous phase by 0.07 M HNO3 strippant 1CX, stream 12, in the 1C column.

Uranium purification. Uranium is purified by a second solvent extraction cycle and by silica gel adsorption. To this end, the uranium-bearing aqueous stream 14, 1CU, leaving the 1C column is concentrated by evaporation, reacidified, and passed through the 2D column. There uranium is extracted by 30 v/o TBP in solvent 2DX, stream 18. Extract 2DU, stream 21, leaving this column is scrubbed with dilute nitric acid 2DS, stream 19, containing hydroxylamine and hydrazine. The scrub stream is intended to free the uranium of traces of plutonium and fission products, which leave column 2D in the aqueous raffinate 2DW, stream 20.

Uranium in the extract 2DU, stream 21, is returned to the aqueous 2E Prod, stream 24, by stripping with 0.01 M HN03, stream 22. This stream is concentrated by evaporation and passed through silica gel, which removes most of the remaining fission products from the uranyl nitrate product stream 27.

Plutonium purification. Plutonium in the aqueous IBP stream 28 leaving the electrocell is purified by two additional cycles of solvent extraction. This plutonium is oxidized to the tetravalent state and reacidified by addition of N204, stream 29. In the 2A column extraction with solvent 2AX, stream 32, and scrubbing with 1 M HNO3 2AS, stream 33, reduces the fission-product content of the extract 2AP, stream 35, to 1 percent that of the feed 2AF, stream 31. Plutonium and traces of uranium in the extract 2AP, stream 35, are returned to the aqueous phase in column 2B by stripping with 0.35 M HNO3 2BX, stream 36. In streams containing plutonium at 35°C, the nitric acid concentration must not drop below 0.35 M; otherwise, insoluble plutonium polymers will form. These are inextractable by TBP and deposit in equipment, plug lines, and represent a criticality hazard. Conditions under which plutonium polymer forms are detailed at the end of this chapter.

In column ЗА of the third plutonium cycle the fission-product content of the plutonium is reduced another factor of 100 by another extraction with TBP and scrubbing with 1 M HN03. The 3B column returns plutonium in the 3AP extract stream 43 to the aqueous phase by stripping with 1 M HNO3 3BX, stream 48, containing hydrazine and hydroxylamine to reduce plutonium to the trivalent state. Uranium in the organic feed to 3B remains in the organic effluent 3BW, stream 45. Scrubbing with solvent 3BX, stream 46, reduces the uranium content of the aqueous plutonium product 3BP, stream 47, to less than 0.01 percent.

The dilute aqueous solution of plutonium nitrate, stream 47, is acidified with nitric acid and washed with dodecane diluent in the 3PS column to remove traces of TBP. After concentration by evaporation in the 3P concentrator, this becomes the concentrated plutonium nitrate product solution 56 of the solvent extraction portion of the plant.

Recycle streams. Uranium and plutonium remaining in solvent leaving columns 2B (stream 37) and 3B (stream 45) are recovered by recycling these as stream 9 (POR) to the IB electrocell. Plutonium and uranium in aqueous streams leaving the 2D column (stream 20), the ЗА column (42), the 2A column (34) and the 3P concentrator (54) are combined as stream 1SF (57). Plutonium in stream 57 is made tetravalent by N204 (61). This uranium and plutonium are extracted with 30 v/o TBP in the IS column. The extract ISP, stream 60, is returned to the third stage from the top of the HA centrifugal contactor.

Pulse columns. Dimensions of the pulse columns of the Barnwell plant are given in Table 10.9.

Solvent recovery. To prevent cross-contamination of products and to allow for the greater degradation of solvent by high concentrations of fission products and plutonium, two independent solvent recovery systems are provided. Solvent recovery system 1 processes solvent 1CW, stream 13, which has been used in the high-activity codecontamination, partitioning, and plutonium purification cycles. System 2 processes the low-activity solvent 2EW, stream 23, which has been used only for final uranium decontamination. Solvent in both systems is processed before recycle by a sodium carbonate wash, filtration and a nitric acid wash. System 1 also uses a second sodium carbonate wash.

Aqueous wastes. High-level aqueous wastes HAW (stream 5) and 1AW (stream 59) are concentrated by evaporation. The condensate, containing recovered nitric acid, is recycled. Processes for converting the concentrated, unneutralized liquid wastes to solid suitable for long-term storage have not been finalized.

Inside Separating

diameter height

Nitric acid recovery. Aqueous nitric acid is recovered from the high-level waste evaporator, from recycle streams 15 and 25, and from nitrogen oxides in dissolver off-gases as described in Secs. 4.11 and 4.13.

Effluent treatment. All gaseous effluents are processed for maximum practicable recovery of contained radioactivity. Most of this is in the fission-product gases vented from the shear and in the dissolver off-gases. These pass through a dust screen, condenser, mercuric nitrate iodine scrubber, and NO* absorber. Gases leaving the NO* absorber are combined with vessel off-gases from the solvent extraction system and pass through a second iodine scrubber, a silver zeolite iodine absorber, and a high-efficiency particle filter, before discharge with ventilating air through a 100-m stack.

Because of the extensive recycle of liquids, the net water feed is only 2 m3/h. Excess water is vaporized into stack effluent. No liquid radioactive wastes are to be released to the environment.

The principal radionuclides to be discharged from the plant as originally designed are gaseous tritium and 85 Kr. The 100-m stack provides adequate dilution and dispersion. Equipment for removing 8SKr will be added when fuel irradiated after 1982 is to be reprocessed.

The main sources of radioactive waste are fuel reprocessing plants. More than 99 percent of the total radioactivity generated by nuclear technology appears eventually in wastes from re­processing plants, most of it in HLW. In a nuclear economy doing without reprocessing, spent fuel itself would be high-level radioactive waste.

Liquid HLW is the concentrated aqueous raffinate from the liquid-liquid extraction process. It contains practically all fission products, neptunium and transplutonium elements as well as

0. 5 to 1.0 percent of the uranium and plutonium fed to the extraction process. It represents a very small fraction of the total radioactive waste volume produced in nuclear installations.

Solid high-level wastes are the cladding hulls of spent fuel elements from the chop-leach

114C being released from reprocessing plants now operating and to be released from those under construction or in the planning may become a waste in the future. Because of its long half­life, it will accumulate in the atmosphere and, in the long run, contribute significantly to the total radiation exposure from fuel-cycle operations. It may therefore become necessary to recover 14 C from dissolver off-gas and to treat it as a waste. There is, however, no urgent need to develop the required technology. Only if high-temperature gas-cooled reactor fuel were to be reprocessed would 14 C recovery be necessary.

head end of the reprocessing plant and some undissolved solids as sludges from feed clarification. Their radiation characteristics are similar to liquid HLW but on a lower concentration level.

Refabrication plants for plutonium-recycle fuel and liquid-metal fast-breeder reactor (LMFBR) fuel to be combined with reprocessing plants will be the principal sources of alpha wastes. These may be liquids (sludges) or solids, the latter being combustible or not. Combustion is an effective way to reduce the volume. The beta and gamma activity concentration of alpha waste is by orders of magnitude lower than that of corresponding HLW. Therefore only light or no shielding and no cooling are required. However, the total alph^ activity is within the same order of magnitude as that of HLW, causing a long-term biological hazard potential similar to that of HLW. This is reflected in similar critieria for conditioning and disposal techniques.

Among the MLW and LLW streams from reprocessing plants are non-alpha wastes. The term non-alpha waste includes all nongaseous wastes that are not high-level wastes and whose radioactivity is due mainly to beta and gamma emitters. Usually, their alpha radio­toxicity is on the same order as that of a relatively rich uranium ore. These waste streams are generated by various operations, including decontamination of equipment. Their biological hazard potential is much smaller than that of HLW and alpha waste, and lasts for much shorter periods of time.

Tritium as well as radioactive krypton and carbon dioxide are in today’s technology released to the atmosphere. Increasing fuel-cycle activities and enforced environmental protec­tion standards, however, have resulted in the requirement for recovery and safe storage of krypton and in some countries of tritium. The design of a German 1400 MT/year reprocessing plant provides for the recovery of at least 95 percent of the krypton, which is to be recovered by cryogenic distillation and will be delivered to the waste management section pressurized in steel bottles. Iodine will be fixed on solid absorbers and tritium will be collected as tritiated water in tanks for intermediate storage. At the AGNS^ plant, provision is presently made only for iodine removal (Chap. 10).

Two more LLW streams from sources other than reprocessing should be mentioned that are significant because of their large volumes. Nuclear power plants produce large volumes of non-alpha waste whose biological hazard potential is small. Techniques for treatment and disposal of this waste type will easily meet the safety requirements but will have to be optimized in terms of economics. Uranium mills generate large amounts of ore tailings with relatively high concentrations of alpha emitters, particularly radium. This is basically a naturally occurring material. However, it is moved from an underground ore deposit to the surface and therefore creates an additional health hazard. Attention given to mill tailings is surprisingly modest compared to that given waste from reprocessing.

In this chapter, the primary emphasis will be on HLW. Non-high-level alpha waste, tritium, 1291, and 85 Kr will be treated to some extent. Volumes and radioactivity concentrations of these wastes to be expected from a 1400 MT/year reprocessing plant according to a German design are given in Table 11.1 [D2].

1.5 Uses

As the uranium atom density is higher in uranium metal than in any uranium compound, metal is the preferred form of uranium for applications where the highest nuclear reactivity or highest density is wanted. For example, uranium metal was used by Fermi to create the world’s first nuclear chain reaction with the limited amount of uranium then available. Uranium metal is still the fuel form preferred for nuclear reasons in reactors fueled with natural uranium and moderated by graphite, such as the Magnox nuclear power reactors used extensively in Great Britain. Because of its high density, uranium metal is used for compact shielding of x-rays or gamma rays and as counterweights in machinery. In such applications, uranium depleted in 235 U is preferred because nuclear reactivity is not needed.

1.6 Phases of Uranium

The phases of metallic uranium and their transition temperatures are listed in Table 5.5.

The process for producing uranium metal by reaction of UF4 and magnesium developed at Iowa State College in the United States was described by Wilhelm [W3] at the first Geneva Conference. In this process, the additional heat needed to melt the products was provided by preheating the reactor and its charge of UF4 and magnesium according to a schedule designed to heat the reactor and its entire contents by several hundred degrees before any part reached a high enough temperature to set off the reaction. The reactor used to contain the high pressure produced momentarily when the reaction mixture reached the melting point of MgF2 consisted of a steel tube lined with an insulating layer of specially purified dolomitic lime. Lime is one of the few oxides that does not react with and contaminate molten uranium.

The largest reactors used at Iowa State College were 35 cm in diameter and 120 cm high, made of standard-wall seamless steel pipe. They were closed with a welded steel plate at the bottom and a cover bolted to a flange at the top. The liner was made of electrically fused dolomitic lime which was finely powdered and packed into the steel tube to form a dense layer

2.5 cm thick on the bottom and 1.25 cm thick on the sides.

A mixture of finely ground UF4 and a slight excess of granular, oxide-free magnesium was packed tightly in the reactor to within 8 cm of the top. The charge was covered with a tightly fitting graphite plug, and the space above the graphite was filled with powdered lime. Closure was by a steel cap bolted to the reactor flange.

The reactor was placed in a preheating furnace, whose optimum temperature was in the range 550 to 700°C. With the 35-cm reactor, a furnace temperature of 56S°C initiated the reaction after about 3.5 h. The reaction was over in about 1 min, and the uranium remained molten for about 10 min. The reactor was then cooled, first by air and then by water, opened, and the contents removed. An average of 108 kg of massive metal was obtained per batch, in

98.3 percent yield based on UF4 charged.

The process used in France at Malvesi [B5] is generally similar, except that the reactor is lined with MgF3. A preheating period of 10 h is used to produce a 210-kg ingot, and a 20-h period for 450 kg.