Although crossflow filtration is not a new technique, is it nevertheless one of the more recent applications of filtration used in radioactive processing

environments. Unlike the previous two filtration methods, which produce, in a single pass, solids which are essentially free of the mother liquor, cross­flow filtration produces a concentrated slurry of the solids within the origi­nal mother liquor. Successive additions of water or other process fluid and re-filtration steps are necessary to free the solids from the mother liquor.

Crossflow filtration recycles the solid-liquid slurry at high velocity (typi­cally 4 to 5 m/sec) and moderate pressure through a bundle of several hundred parallel sintered stainless steel, or porous graphite, filter tubes, normally arranged in the manner of a shell and tube heat exchanger. The pore size can range from a few hundredths to a few tenths of a micron, depending on the particle size of the slurry. The tubes are typically 1 to 2 meters long and 5 to 6 mm internal diameter. Sintered stainless steel tubes are welded to the end plates to form the tube bundle, while graphite tubes are sealed with elastomer seals.

During each passage of the slurry through the tube bundle, some of the liquid in the feed slurry permeates the pores in the tube and flows into the “shell side” of the filter unit, which is maintained at a lower pressure than the tube side. Smaller filter tube pore sizes are found to be capable of sus­taining higher specific liquid flowrates (or flux) than larger ones. This is because they are less susceptible to blocking by the solid particles. As the slurry is recirculated through the tube bundle it becomes progressively more concentrated. When the desired concentration is reached the filtration is stopped and the concentrated slurry is mixed with water or other washing agent and the filtration cycle is then repeated. These cycles are repeated until the desired degree of separation of mother liquor from the solids is achieved.

Crossflow filtration has been developed and demonstrated for several nuclear waste processing plants that are under design and construction in the USA. At the Savannah River nuclear site in South Carolina, the Salt Waste Processing Facility will use crossflow filtration to treat high level waste from the tank farms prior to cesium-sodium separation and vitrifica­tion of the highly active constituents in the Defense Waste Processing Facility (Poirer, 2007). The high level waste is first mixed with mono-sodium titanate (MST) to sorb TRU elements in the waste, and then the MST and solids already present in the waste, are concentrated by crossflow filtration. At the Hanford nuclear reservation in Washington, the Waste Treatment Plant also will use crossflow filtration to concentrate high level waste sludge prior to vitrification (e. g. Geeting, 2006 and Peterson, 2007).

A highly successful nuclear application of crossflow filtration that has been in use since the early 1990s is the Enhanced Actinide Recovery Plant (EARP) at the Sellafield nuclear site in northwest England. In EARP, acidic, TRU-contaminated wastes arising from the “Magnox” reprocessing facility, and which also contain significant quantities of iron, are neutralized to cause the iron and TRUs to co-precipitate. The TRU-contaminated floc is sufficiently separated by crossflow filtration from the solution to facilitate discharge of the permeate to sea. In this application a modified form of filter bundle is used, with the floc inlet and outlet, and the permeate outlet positioned all at the same end of the bundle (Fig. 3.6).

Two stages of ultrafiltration are employed in EARP to optimize the pumping requirements. The de-watered floc required for immobilization and disposal is a thixotropic material with a viscosity of approximately 7 poise, which would require special pumps. Therefore, a two-stage ultrafiltra­tion is employed. Up to 90% of the water is removed from the slurry in the first stage so that a standard pump can be employed to recirculate the slurry. De-watering is completed in a second stage on a smaller slurry volume that consequently uses smaller specialized pumps than if a single stage process were employed.

Backwashing and chemical cleaning with nitric acid removes any floc that is fouling the tubes. However, the membranes are expected to have a finite

3.6 a and b Schematic representation and photograph of an EARP ultrafilter module. b Source: Nuclear Decommissioning Authority ("NDA"), copyright: Nuclear Decommissioning Authority ("NDA").

life through incomplete cleaning, mechanical damage, wear to the coating and possible blockage of the pores or the tubes themselves. Therefore, the EARP ultrafilter cartridges are specially designed for remote maintenance and replacement (Fig. 3.7), being contained in type 3 PSCs. A fixed housing is permanently built into the plant below the cell roof and thus within the biological shielding. The filter modules fit within these housings and all inlet and outlet ports and seals are incorporated into one plug unit in the module top. This is sealed to the housing by replaceable O-ring seals. When necessary, the complete module, complete with its O-ring seals, can be withdrawn from the housing into a flask positioned above the cell top, and

a replacement module installed into the housing by reversing the process. In EARP, nine modules of ultrafilters containing 800 membrane tubes are used for primary de-watering while two modules of 500 membrane tubes complete de-watering.

PUREX is based, by definition, on liquid-liquid solvent extraction chemis­try of the well-known extractant, tri-n-butyl phosphate (TBP), diluted to nominal 20-30% (by volume) with a normal paraffinic hydrocarbon (NPH) organic diluent. Note that the diluent is required only to maintain the physi­cal characteristics of the organic phase (primarily viscosity and density) in a workable range for use in the salient solvent extraction equipment. In its simplest representation, the process is indicated in Fig. 6.1. The primary inputs are irradiated nuclear fuel and numerous process chemicals, pre­dominately nitric acid (HNO3). Three major output streams result: 1) a rela­tively pure Pu nitrate solution, 2) a relatively pure U nitrate solution, and 3) process wastes. The U and Pu nitrate products can be further purified in additional cycles of PUREX processing and are subsequently converted to solids, typically the metal oxides. The waste is further classified into three categories; high, intermediate, and low level, as specified by the relative radioactivity content. The overarching objectives of PUREX are to produce

pure U and Pu products in very high yield and purity, while minimizing the losses to and volumes of the resulting waste streams.

Actinide chemistry in the PUREX process is relatively straightforward. TBP is a very selective extractant for the actinides in the +6 and +4 oxida­tion states; therefore the uranyl (U(VI), as UO2[3]+) and Pu+4 (Pu(IV)) cations are readily transferred to the organic phase due to the formation of neutral nitrate-TBP complexes in 2-4 M nitric acid. The distribution ratio of the actinides are exemplified in Table 6.1. The extraction mechanisms are indi­cated by the following chemical equilibria for nitric acid/nitrate media:

UO22+ + 2NO3- + 2TBP W [UO2(NO3)2 ■ 2TBP] 6.1

Pu+4 + 4NO3- + 2TBP W [Pu(NO3)4 ■ 2TBP] 6.2

The affinity of TBP is substantially lower for the +5 oxidation state, notably Np(V) as NpO2+, and virtually nil for the +3 and lower oxidation states (i. e., Pu(III), Am(III), Cm(III), Cs(I), Sr(II), etc.). The pivotal point of the process chemistry is the high thermodynamic stability of UO22+ (pre­dominate species under all process conditions) and relative ease with which the oxidation state of Pu can be controlled or adjusted, either chemically or electrochemically. Thus, by controlling the plutonium as Pu(IV) during the extraction step, both U and Pu are transferred (almost quantitatively) to the organic phase. The U-Pu partitioning is accomplished by using condi­tions conducive to the reduction of Pu(IV) to the virtually inextractable Pu(III) oxidation state, effectively back-extracting plutonium from the loaded organic. The U(VI) is subsequently recovered or stripped from the organic using dilute nitric acid. Several important nuances of PUREX process chemistry as related to TBP and the associated extraction chemistry (Eq. 6.1, 6.2) should also be noted:

• simple stoichiometric ratio of TBP to U or Pu

• excess NO3- drives the equilibrium to the right, a salting-in effect

• TBP measurably extracts nitric acid into the organic phase

• TBP is slightly (but measurably) soluble in the aqueous phase.

Used nuclear fuels, even after dissolution, can be likened to “the cat’s breakfast” in that a large proportion of the entire periodic table resides in solution, due primarily to the complicated spectrum of elements formed by the fission process, i. e. the fission products. Due to very high selectivity of TBP, the desired decontamination of U and Pu from most of these elements is easily attained. However, there are a few fission products considered to be problematic or troublesome because their decontamination to the desired levels has proven difficult. A notable source of residual radioactive contamination of the products is from the short-lived isotopes 95Zr with a half life of ti/2 ~64 days and its radioactive decay daughter 95Nb with t/2 ~ 35 days. Consequently, residual contamination with 95Nb is always concomitant with 95Zr contamination. In modern reprocessing plants, 95Zr/95Nb contami­nation is controlled or alleviated by cooling the used nuclear fuel for >3 years prior to reprocessing. Another obstinate source of residual radioac­tive contamination of the U and Pu products is from 106Ru and its 106Rh daughter, both with a t/ ~ 1 year. Impractical cooling times of >10 years would be required to obtain the same benefit from cooling as observed for Zr. Consequently, provisions for the acceptable decontamination of the U and Pu products from 106Ru/106Rh must occur in the flowsheet. Iodine and Tc are problematic for environmental reasons; furthermore, Zr interferes with Tc decontamination and Pu partitioning. Iodine, on the hand, is largely reduced to unimportant levels with the typical >3 year cooling period and controlling the fuel dissolution chemistry, and will subsequently not be further discussed. Last, but not least, larger quantities of Np are formed in higher burn up LWR fuels associated with modern reactor core manage­ment techniques. Coupled with environmental concerns, Np chemistry in PUREX has been extensively studied in recent years. The chemistries of these elements are comprehensively reviewed in the literature (Shultz 1984, Benedict 1981, Sood 1996).

The inner working of a PUREX “black box” as represented in Fig. 6.1 is obviously a rather complicated series of interrelated process steps that function to achieve the overall objectives of the process. The process steps can be characterized in general terms to include: [4] 2

regard to fission products. Recovery and purification is achieved by going through a succession of liquid-liquid extraction, scrub, back — extraction, and solvent cleanup cycles. Note that the concomitant high performance requirements of purity and recovery in nuclear reprocess­ing are uncommon to typical hydrometallurgical applications; in con­ventional metal recovery industries, the aim is to promote recovery efficiency, at the expense of purity, or vice versa.

3. The organic phase is recycled in the process and solvent cleaning and regeneration are important aspects of the process flowsheets.

4. Due to the solubility of TBP in the aqueous phase, all flowsheets include a “diluent wash” step to back extract or recover TBP from the aqueous effluents. This is a particularly important aspect if a stream is to be concentrated via evaporation (e. g., with nitric acid recovery) to mitigate potential safety concerns and precipitate formation.

5. Treatment of the radioactive waste effluents ultimately results in the solidification or vitrification of the process wastes for final storage and, ultimately, disposal. Additional steps in the waste treatment cycle may include evaporation, process chemical recovery (notably HNO3) for recycle into the process, and compositional adjustments.

6. The final uranium and plutonium products are typically oxides. A con­version process is included to recover U and Pu from aqueous nitrate media as the solid metal oxides. Typical steps included in the conversion process are precipitation, usually as U peroxides and Pu oxalates (which often facilitates further decontamination), with subsequent roasting or calcination to the solid metal oxides.

All of the flowsheets developed for the UREX+ demonstrations, conducted at Argonne National Laboratory, were designed to be operated using 2 cm countercurrent centrifugal contactors as the separation equipment. The flow rates were adjusted to the equipment design maximum. To ease the demonstration the solvent was not recycled, with the exception of the CCD-PEG and FPEX process tests where the supply of solvent was limited. In an actual plant application, a solvent wash section would be added to

each process module, before recycling the solvent to the front end of the process module.

A flowsheet with selective stripping of Cs and Sr has been developed for partitioning HLW with long-lived (Cs-137 and Sr-90) and ultralong-lived (An) radionuclides into fractions. The flowsheet involves the following operations:

• Combined extraction of Cs, Sr, An and REE by a UNEX-extractant (11 stages).

• Scrubbing of the extract with a solution of 1 M HNO3 containing 0.03 M citric acid (1 stage).

• Combined stripping of Cs and Sr by a solution containing 1 M guanidine nitrate, 0.1 M HNO3 and 0.03 M citric acid (6 stages).

• Stripping of An and REE by a solution containing 1 M guanidine car­bonate, 0.2 M acetohydroxamic acid and 0.03 M DTPA (3 stages).

• Regeneration of the extractant by a 3 M HNO3 solution (5 stages).

This flowsheet (Fig. 9.5) was also tested at pilot plants with centrifugal contactors at RI and the Idaho National Laboratory.

The selection of compatible materials depends strongly on the process environment. As described in the first section of this chapter, most of the materials used for the pyroprocess are in moderate corrosion condi­tions during normal operation. Carbon steels and stainless steels can be used without corrosion problems. For example, long-term durability of steel materials has been tested with LiCl-KCl salt and liquid Cd with actinides and FP at approximately 500 °C. On the other hand, stainless- steels are not suitable for the crucible containing liquid Cd because nickel dissolves in them. No detectable damage has been reported for a mild steel crucible in a laboratory-scale electrorefiner (Battles, 1993) or for Mo a steel crucible in INL’s engineering-scale electrorefiner. The integrity of Mo steel was also confirmed in CRIEPI’s experiments with an electrorefiner and with a counter current contactor. The materials suitable for the more severe conditions in the pyroprocess are described in this section.

There are numerous variants of producing solid-phase extraction material, but the most widely used techniques are based on the physical adsorption of the extractant into the pores of the support material. The preferred method for incorporating the chosen extractant into the support is often dependent upon the type of extractant-metal complex and is therefore determined by the intended separation process. Preparation routes reported in the literature during the last forty years primarily involve adsorbing the extractant into the lattice of a polymer substrate — known as impregnation methods — or adding the extractant to a mixture of monomers during the bead polymerization process, i. e. the well-known Levextrel resins which incorporate the extractant in polystyrene-divinlybenzene during the copo­lymerization step (Kauczor et al., 1978, Poinescu et al., 1985, Ionue et al., 1987, Yoshizuka et al., 1990). Other techniques such as adsorbing the extract­ant in silica gel, silica and organic polymer mixtures or adding the extractant to a dissolved polymer and then precipitating the mixture have also been reported. Excellent reviews of the classical preparation techniques have been presented by Warshawsy (1981) and Cortina et al., (1997) and the reader is directed to these works for detailed descriptions of the synthetic methods. Variants of the classical impregnation methods have received the most recent attention in terms of developing solid-phase materials with extractant systems relevant to An and Ln separations. A brief review of the impregnation methods is therefore germane to this work in order to famil­iarize the reader with this production scheme.

Recalling the assumption that a solid-phase extraction resin represents a complexing agent dispersed homogeneously within a solid polymeric medium, the impregnated extractant should behave as in liquid-liquid extraction, but maintain a strong affinity for the solid polymer matrix. In order to approximate these criteria, Warshawsky (1981) presents the fol­lowing requirements for the extractant, polymeric support, and the impreg­nation process:

• The extractant should be a liquid or retained in the liquid state by the addition of an appropriate diluent.

• The extractant should have a very minimal solubility in the aqueous solution containing the solute to be extracted.

• The polymeric support should be fully expanded and remain so during the impregnation process — macroporous polymers exhibit minimum volume variations during impregnation and are therefore preferred.

• The impregnation process should not have a deleterious effect on the properties of the extractant or polymer.

Previous reviews by Warshawsy (1981) and Cortina et al. (1997) have classified the methods for producing extractant impregnated resins as follows: [14]

contacted with the polymer in batch mode for a period of time necessary to obtain maximum saturation of the extractant within the polymer pores. The diluent is then removed via vacuum evaporation resulting in a two-component, polymer-extractant material.

• Wet impregnation — an extractant or extractant and diluent is adsorbed into the polymer as described above, but the water-immiscible organic diluent is not removed by evaporation.

• Modifier addition method — a modifier is added to the extractant-diluent and the mixture is adsorbed into the polymer as described in the above methods. The diluent is then removed by evaporation, leaving a poly­mer-extractant-modifier resin. The chosen modifier is typically more polar than the extractant and is added to enhance water penetration into the porous network of the polymer.

Of course, slight modifications to these methods may be used depending on the desired application of the solid-phase extraction material. The dry method is preferred for impregnating the more hydrophilic extractants such as amines, ketones, and esters and is the most used technique for making solid-phase extractants applicable to An or Ln separations. The preparation of a small batch of resin using the dry impregnation method is shown in Fig. 13.2.

It should also be noted that the class of inert, macroporous or macrore­ticular polymers have become the most widely used substrates in the con­temporary synthesis of solid-phase extraction media for separating the An

and Ln elements (Horwitz et al., 2006). These polymeric macroporous resins have a rigid three-dimensional structure that provides minimum solvent swelling during the impregnation process. The inert macroporous resins are also capable of adsorbing large amounts of the extractant due to a high specific surface area (500-900 m2/g) and typical porosity fraction ranging from 0.4 to 0.6 (Van Hecke et al., 2006). The macroreticular polymers are typified by a continuous gel phase and a continuous pore phase and are manufactured with varying degrees of hydrophobicity. Thus the specific polymer can be matched with the properties of a given extractant to opti­mize the hydrophobic/hydrophilic balance. This is important in the solid — phase extraction media since the goal is to minimize losses of the extractant to the aqueous phase while maintaining enough hydrophilicity to maximize mass transfer of the solute into the resin pores and between phases. Examples of the macroreticular polymers frequently used are the non-ionic Amberlite XAD and Amberchrom CG series resins.

Most studies support the assumption that immobilization of the extract­ant on the internal surface of the macroporous, inert resins is a combina­tion of adsorption via van der Waals forces (Hommel et al., 1983, Bobozka et al., 1985, Cote et al., 1987, Handley et al., 1991, Cortina et al., 1994a, Cortina et al., 1993, Villaescusa et al., 1992) and potentially physical trap­ping of the ligands within the pores of the resin beads (Handley et al., 1991). Impregnation of the extractant within the stationary phase is con­sidered to be a combination of pore filling and surface adsorption. The extractant first fills the pore space beginning with the smallest pores and moving up to pore sizes of approximately 10 nm. Surface adsorption then becomes the dominant retention mode in the larger pores (Guan et al., 1990). Although it is not yet possible to perform a precise a priori predic­tion of the adsorptive and retention properties of a given ligand-polymer system, the hydrophobic or non-polar sections of a ligand molecule are generally attracted to hydrophobic polymers and the hydrophilic or polar molecules to hydrophilic surfaces. For example, the non-ionic, hydrophilic XAD-2 and XAD-4 resins (Amberlite) are often used with many of the ligands important to An and Ln separations because the non-polar vinyl and styrene groups of the polymeric matrix serve to anchor the extractants through their alkyl chains and/or aromatic rings, while the functional groups in the ligand remain capable of forming the desired metal complex (Fig. 13.3). This has been shown by various studies (Bobozka et al., 1985, Cote et al., 1987, Cortina et al., 1994a, Cortina et al., 1993) wherein the authors concluded that this weak interaction between the extractant mol­ecules and the polymer support is the primary mode of ligand retention, but does not adversely affect the complexing characteristics of the extractant.

13.2

Depiction of ligand arrangement on polymer substrate.

The fate of strontium species in the microbial culture was determined by scanning electron micrographs (SEM) of SRB biomass previously exposed to medium containing Sr2+. Precipitates on the cells surfaces were analyzed by an energy dispersive x-ray (EDX) spectrometer using a Cd109 radioiso­tope source and a Si(Li) semiconductor detector of resolution 170 eV for

1300

1000

5.9 keV Mn Ka x-ray. The analysis revealed whitish crystallization on the surface of the cells. The composition of the precipitate as observed by the SEM was indeterminate (Figs 15.9 and 15.10). Speciation analysis of bacte- ria-free controls revealed that the solutions were undersaturated with respect to insoluble Sr species. This was consistent with the observation of adsorption as the dominant mechanism for metal removal from solution.

For the small part that precipitated, EDX analysis showed that about 65% was Sr2+ compounds (Fig. 15.10). In spite of the positive identification of Sr in the precipitate around the cells, the EDX scans clearly show that amount of precipitated Sr around the cells was indeed very low to insignifi-

cant. The low levels of precipitates as well as Sr peaks in the control sample after EDX analysis confirms that the observed Sr2+ in the precipitate is produced by the interaction with the bacteria. The low levels in both samples suggest that the production of Sr precipitates is transient in nature, that the accumulation of the precipitate is limited by the sorption rate to reaction sites on the cell. The findings on the fate of Sr in the SRB consortium culture are in agreement with other reports where functional groups on the cell surface of bacteria facilitated bulk metal binding from aqueous solution (Sherbet, 1978). However, further studies still need to be conducted to clarify the observed findings.

At the beginning to the 21st century, the emergence of rapid modernization in the economies of China, India, Russia and Brazil, together representing a bit more than 30% of the global population, has increased demands for new supplies of energy. In addition, it is recognized that geometric growth of the human population on the planet could result in a 40% increase in the global population by 2050. Further, either depletion of fossil carbon resources or alteration of the global climate due to CO2 emissions will also bring additional pressures for new power sources that do not emit substan­tial quantities of greenhouse gases. These factors taken together would appear to be primary drivers for a projected expansion of the effective utilization of existing energy sources.

The single pass nuclear fuel cycle, practiced by about two-thirds of the nuclear power generating economies globally, is quite wasteful of this resource, extracting only a few percent of the energy potential of uranium (and for the moment almost none of the energy potential of thorium). At present usage levels and patterns, it is estimated there is approximately a 50 year supply of uranium available in economically recoverable mineral resources. With the application of enhanced recovery systems for uranium mineral resources, this supply could be extended (with current usage pat­terns) to about 250 years. If the nuclear component of global energy produc­tion grows (as many are projecting), these estimates are far too optimistic. It will become increasingly important to more efficiently utilize this resource through the recycle of plutonium; first to light water reactors and ultimately to a fleet of fast spectrum reactors that are capable of “burning” actinides that do not readily undergo fission in light water reactors. This will require recycling and fabrication of fuels containing a larger proportion of heavier actinide isotopes. Another advantage gained by more extensive recycle of actinides for power production will be the potential for eliminating (or at least reducing the amount of) long lived radiotoxic actinides like Am, thus improving the viability of any geologic repositories that are constructed.

To further extend the potential of this resource, it makes sense to consider the potential (an effort for the moment being led by India) of the thorium — uranium breeder reactor cycle. As thorium is three to four times more abundant than uranium, the viability of fission powered electricity is further extended by taking this step.

There are certain types of evaporators which are too demanding for aus­tenitic stainless steel. For these processes titanium and, more recently, zir­conium are used (Baldev, 2006). These materials have an extremely high resistance to nitric acid corrosion (zirconium is virtually immune to corro­sion in pure nitric acid). However, serious problems can arise in the pres­ence of impurities such as fluorides, and great care is needed to ensure correct process chemistry. These materials are very expensive and can be difficult to fabricate without the necessary know-how, i. e. only companies with the relevant expertise are capable of manufacturing high integrity plant and equipment in these materials. Zirconium, in the form of specially designed rings formed from strip, is used in packed columns to supply a large surface area on which condensation can occur. The older packing material, stainless steel, contributed significant quantities of iron to the process fluids even though their corrosion rate was very low and this iron, added to the process liquor, could enhance the corrosion of stainless steel vessels and pipework. Zirconium has virtually zero corrosion even in nitric acid vapor.

The separated U and Pu streams emanating from the partitioning cycle still contain some fission product impurities. In order to complete decontamina­tion of products, complementary purification cycles are required based on the same principles of the extraction chemistry described above. These two independent purification cycles are represented in Fig. 6.3. The primary contaminants in the Pu stream are typically Ru, Rh, and in some cases Np (depending on the operation and design of the early partitioning step), and the Pu purification cycle allows additional decontamination factors to be achieved with respect to these fission products. The uranium purification cycle is used primarily to enhance the decontamination from Np; it has been reported that 75% of the Np initially in the dissolved feed ends up in the uranium product from partitioning that goes to uranium purification. Of that, 99.8% of this Np is recovered in the raffinate of the uranium purifica­tion cycle. A Np decontamination factor of DF > 150 has been reported for the uranium purification cycle (Baron 1993).