The problem of radiation pollution in the environment is expected to increase as industries in the world experience pressure to reduce their carbon atmospheric contributions. The dilemma is that production of goods will only increase with population growth. Most countries are considering nuclear energy as the interim solution to the growing energy crisis. This will require environmental engineers and scientists to delve deeper into research on environmentally friendly processes to counter the pollution effects to avoid leaving a pollution legacy for the future generations. Advanced microbial cultures will be sought to treat a wider variety of recalcitrant pollutants. Discussions on the possibility of genetically engineering special­ized cultures for the purpose are not new in the environmental engineering fraternity. However, the application of ideas bears a large ethical burden as it is forbidden in almost all countries in the world to introduce genetically engineered organisms into the environment. Less aggressive methods for dealing with the problem without violating ethics include in situ bioaug­mentation and molecular bioaugmentation to a certain extent.

This entails identifying indigenous species of bacteria within the vicinity of the contaminated site and determining the critical carbon sources and nutri­ents that could be supplied to encourage the growth of the target species. When the selected nutrients are introduced into the environment, either by injection into boreholes or by spreading on the ground, the target species will out-compete other species and will be able to degrade the contami­nants. The potential problem with this form of bioaugmentation is that the nutrients may be viewed as pollutants in their own right especially at the beginning of the bioaugmentation process when microbial loading is very low. The nutrients such as NO3- and SO42- have undesirable pollution effects on receiving water bodies such as eutrophication of streams receiving the base flow from the remediated areas.

The molecular bioaugmentation process utilizes genetic carriers such as transposons and plasmids to shuttle genetic information for toxic metal remediation into native species in the environment or species already adapted to the target environment. Several species of bacteria are capable of picking up and retaining circular fragments of DNA called Broad-Host — Range Plasmids which may be engineered to carry specific genes for the degradation of xenobiotic compounds and transformation of toxic metals (Weightman et al, 1984; Vincze and Bowra, 2006). The same process can be applied using genetically engineered linear DNA called transposons. Although studies have been conducted using these techniques in laboratory microcosms, the application in actual environments has not been attempted (Hill et al., 1994). In the future, it is foreseeable that these methods will find wide application for the new pollutant varieties that may be untreatable by conventional methods.

A very little understood application of bioseparation involves using micro­organisms to discriminate radioisotopes by size. So far, this application has remained conceptual due to limited understanding on the structure and function of organisms that are suspected to achieve biofractionation (Molokwane and Chirwa, 2009). In the latter study, Molokwane and Chirwa observed with a modest degree of certainty that microbial cells previously isolated from a high radiation-exposed facility accumulated C-14 while growing on a C-14/C-12 carbon matrix from powdered nuclear graphite. The experiment was conducted in a closed loop chemostat system equipped with biofilters for collection of suspended matter for analysis. The observed metabolic activity in the cells indicated that the process was possible under very low dissolved oxygen, suggesting that the microorganisms preferred inorganic forms of carbon as the primary carbon source. Bacteria that utilize inorganic carbon sources such as CO2 and HCO3- as primary carbon sources — known as autotrophic organisms — favour anaerobic conditions for growth. However, in this preliminary study, the amount of C-14 remain­ing in solution which could be required to draw a mass balance on C-14 in the system was not measured. These preliminary results on C-12/C-14 bioseparation hold promise for future HTGR nuclear reactors that produce large amounts of low radiation level waste as expired nuclear graphite. Success in the above process is also important for the decontamination and recovery of nuclear graphite from decommissioned plants for reuse in new reactors.

15.9 Sources of further information and advice

Chow B. G. and Jones G. S. (1999) Managing Wastes With and Without Plutonium Separation (Santa Monica, Calif.: RAND, 1999). The reprocess­ing LLW figure also uses data from Groupe Radioecologie Nord Cotentin, Inventaire des rejets radioactifs des installations nucleates, vol. 1, July 1999, p. 19.

International Union of Pure and Applied Chemistry (IUPAC), 1984. Makhijani A., Hu H. and Yih K. (1995). Nuclear Wastelands: A Global Guide to Nuclear Weapons Production and Its Health and Environmental Effects, Cambridge, Mass.: MIT Press, Massachusetts USA.

United States Nuclear Regulatory Commission (http://www. nrc. gov/waste).

[1] Estimating required heights and diameters for the columns. The height determines how much of the resin capacity is utilized before the con­taminant concentration in the effluent becomes unacceptable. The diam­eter mainly determines the liquid throughput for a given utilization of the resin capacity.

• Channeling of the liquid through the ion exchange bed. Larger diameter columns can suffer from channeling if the feed is not adequately distrib­uted across the column cross-section. Channeling can be exacerbated by fissures in resin beds caused by it swelling and shrinking.

• Feed distributor design. A well-designed feed distribution device can mitigate channeling at least to some degree in a large diameter column.

• Deciding on upward or downward flow of liquid through the bed. Up-flow velocities are limited by the need to prevent fluidization of

[2] A mixer compartment where the two liquids are mixed to form a disper­sion. Mixing is typically achieved by a mechanical impeller with the motor and gearbox mounted outside the biological shielding for ease of maintenance. Mass transfer is primarily achieved in the mixing compartment.

• A settler compartment where the two liquids are separated by gravity and exit through a system of weirs. The weir system can be designed for the separated liquids to flow to the succeeding contactor under gravity without the need for inter-stage pumping.

Mixer-settlers’ continued application in nuclear facilities is warranted on the basis of their mechanical simplicity and industrial pedigree and their important chemical engineering attributes are summarized in Table 3.2. Mixer-settlers are well-established equipment for liquid-liquid extraction unit operations for processing radioactive material and are employed in currently operating commercial UNF recycling plants (e. g. THORP, Sellafield, UK; UP3, Cap La Hague, France; and the Rokkasho Reprocessing Plant, RRP, Rokkasho Mura, Japan) mainly for purifying uranium product streams in which the fissile material content is sufficiently low that nuclear criticality can never be reached even within the non-ever-safe geometry of these contactors.

[3] Solvent extraction cycles are dedicated to virtually complete recovery of the uranium and plutonium products with a high degree of purifica­tion from the other constituents of used nuclear fuels, particularly with

[4] Dissolution of the solid used nuclear fuels to produce liquid feed to the separation (solvent extraction) operations. This portion of the process, typically termed headend operations, is an important consideration in determining and controlling the underlying chemistry of the subsequent separations processes. Obviously, the characteristics of irradiated fuel vary widely with a variety of interrelated factors, e. g. fuel type (metallic, UO2, MOx, etc.); reactor type (PWR, LWR, BWR, Fast, etc.); burnup; and enrichment.

Notes: (1) The repository addressed by the GNEP program was Yucca Mountain.

[6] Transuranic elements (of concern): Pu — plutonium, Np — neptunium, Am — americium and Cm — curium.

[7] Non-TRU: Waste containing no more than 3700 becquerels (100 nanocuries) of alpha-emitting transuranic isotopes per gram of waste and half-lives greater than 20 years.

[8] the need to decrease HLW acidity resulting in an increase of volumes of feed solutions;

• the use of complexones and salts of organic acids for stripping of actinides and lanthanides, which makes further handling of the strip product more difficult;

• insufficient mutual purification of Cs-Sr and An-Ln fractions.

Additionally, the use of this system is complicated by the fact that the syn­thesis and purification of phosphorylated PEGs is a rather sophisticated

[9] The reduction potentials of elements in LiCl-KCl eutectic salt into liquid cadmium alloy were derived when the concentrations of the elements in the salt and in liquid cadmium were the same (0.001 mole fraction).

[10] SFCs/Na = DCs/DNa, where DM is the distribution ratio of the extracted alkali cation M.

[11] A DIAMEX raffinate is the genuine highly active solution issued from the imple­mentation of a DIAMEX process, co-extracting trivalent actinides and lanthanides from a PUREX raffinate (issued from the reprocessing of a dissolver liquor by TBP solvent to recover uranium and plutonium).

[12] surrogate, spiked, or genuine PUREX raffinates, as well as well as PUREX concentrates (Serrano-Purroy et al., 2005a, b, Modolo et al., 2007a), as the feeds;

• DMDOHEMA dissolved at 0.65 M in HTP, as the solvent;

• mixer-settlers, centrifugal extractors, rotating ‘Couette-Taylor’ effect columns, and pulsed columns, as the laboratory scale contactors.

Oxalic acid and A-(2-hydroxyethyl)ethylenediamine-A, A’,A’-triacetic acid (HEDTA) were added to the feeds and the scrubbing solutions to limit the extraction of Mo and Zr on the one hand, and Pd on the other. The DIAMEX flowsheet tested in 2005 at the CEA Marcoule (France) on a genuine PUREX raffinate (31 litres, from the reprocessing of 15 kg of a 52 GWd/t UOX type spent nuclear fuel: 6.5 years of cooling), in three

[13] SANEX, for Separation of ActiNides by Extraction. The objectives of the SANEX process are to recover more than 99.9% of the An(III) in a purified product stream containing less than 5 wt.% of Ln(III).

[14] Dry impregnation — this is the most often used method wherein the extractant, or extractant diluted with an appropriate organic diluent, is

The outer and inner boundary conditions for dissolved species u and biomass x are defined by:

where kLU = DwU/Lw, is the mass transfer rate coefficient (L2T^), and u/ = dissolved species concentration at the liquid/biofilm interface (ML-3).

The system described above is an oversimplification of the actual biofilm processes in nature. However, the example serves to illustrate how complex the solution for such a simplified version could be. In the old days before fast and efficient computers, solving such problems manually was unthink­able. Lately, computer speed has increased exponentially and memory is no longer a limiting factor. Innovative tools for simulation of the PDE system and heuristic approaches for estimating parameters are now available. For example, Nkhalambayausi-Chirwa and Wang (2005) applied a custom PDE solver to solve the model equations for simultaneous Cr+6 reduction and phenol degradation in a dual species biofilm reactor. The PDE solver uti­lized the (fourth-order) Runge-Kutta method with spatial discretization using the (second-order) Crank-Nicholson and Backward Euler finite dif­ference methods for the biofilm spatial profiles. The solution of the biofilm PDE system of equations is shown as a solid line in the effluent from a biofilm reactor showing the prediction of the metal reduction process in the biofilm (Figs 15.12 and 15.13).

The parameters in the biofilm model in the example above were estimated using a heuristic procedure — Genetic Search Algorithm (GSA). The first version of this algorithm was implemented in the C programming language using subroutines adopted from Hunter (1998). The GSA uses the inverse of the mean residual sum of squares (MRSS) computed as the global vari­ance (a2) as a fitness function during parameter optimization using the principles of evolution and natural selection (KrishnaKumar, 1993). The global variance is the main objective function for GSA computed as:

G2 =—— Y((obs — y)2 15.16

n — qt!

where a = average deviation of the model from measured values, yobs = observed variables, y = simulated variables, n = number of observations, and q = degrees of freedom representing number of parameters being evaluated.

15.12 Simulation of the reduced metal Mn+ (Cr6+) in a dual species biofilm culture under a range of hydraulic loading conditions: 24 h HRT (Phase I-VI); 11.7 h HRT (Phase VII-X); 6 h HRT (Phase XI-XIV); 17.9 h HRT (Phase XV-XVIII).

15.13 Simulation of the carbon source (P) and metabolite (U) concentration in a dual species biofilm culture under a range of hydraulic loading conditions: 24 h HRT (Phase I-VI); 11.7 h HRT (Phase VII-X); 6 h HRT (Phase XI-XIV); 17.9 h HRT (Phase XV-XVIII).

In Figs 15.12 and 15.13, parameters were estimated from the data obtained from the operation of the reactor at 24 hours hydraulic retention time (HRT) (Phase I-VI). The rest of the phases (VII-XVIII) were simulated using the optimized parameters. The results show high confidence in the optimization routine as the model accurately tracked the trends in effluent concentrations for both the electron donor (P) and the electron sink (M).

The applications of mathematical models to complex biofilm process resembling natural systems are very rare. Most of the reports are based on laboratory-scale pure cultures. Black-box approaches are normally used to evaluate performance of actual systems. The above quoted example by Nkhalambayausi-Chirwa and Wang (2005) was one of the few efforts to mechanistically model a mixed culture system.

The efforts to understand the fundamental nature of biofilm systems are worthwhile since these offer unique solutions to contaminant treatment. For instance, cells growing in biofilm cultures have been observed to perform better than the same species suspended in medium (Semprini and McCarty, 1981). One reason offered for the better performance of cells in the biofilm environment is the effect of shielding from high toxicity levels. It is esti­mated in most biofilm systems that the bulk liquid concentration is much higher than the concentration in the deeper layers of the biofilm. Other complex interactions are also known to exist within the biofilm system. Cultures grown in a cooperative system where some species of microorgan­isms require a product from other species for survival (Nkhalambayausi — Chirwa and Wang, 2005). The mass transport conditions and balanced retention of substrates are required to sustain a culture for a specific purpose.

Biofilm systems have demonstrated the potential to package complex treatment systems into a small space to achieve removal of multiple sub­strates. These systems are especially useful when relatively sensitive species of bacteria are used to treat toxic waste. It is envisaged in this report that biofilm systems will in future form a significant part of the treatment and recovery regime for nuclear and radioactive waste.

In nitric acid solutions, such as PUREX raffinates (where [HNO3] > 3 mol. L-1), the 4f lanthanide metallic cations (Ln) and the 5f americium and curium metallic cations (An) predominantly show the same oxidation state, +III, and many similar physical and chemical properties (Nash, 1993, 1994, Beitz, 1994, Morss, 1994, Marcus, 1997):

• They are considered as ‘hard acids’ in Pearson’s theory (Pearson, 1963).

• The 4f and 5f orbitals have a rather small radial extension and are more or less protected by the saturated lower electron orbitals, respectively

the 5s2-5p6 for the lanthanides and the 6s2-6p6 for the actinides. Thus, the nf electrons scarcely interact with electrons of neighbouring ligands and their electronic properties are only slightly affected by their environments.

• The ionic radius shortens along the 4f and 5f series as the atomic number increases. Thus, it is easy to predict the higher electrostatic reactivity (formalized by the ionic potential closely linked to the charge density) of an element of higher atomic number, Z, compared to that of an element of lower atomic number in the periodic table.

• Since Ln(III) and An(III) have the same positive charge (+3), their discrimination through solvent extraction involving ‘hard bases’ (e. g., ligands bearing oxygen donor atoms in their structures) will mainly be due to geometric and/or steric hindrance reasons: the better the fitting of a metallic cation radius with the cavity size of the complexing/extract — ing agent or its coordinating site, the better the discrimination. However, the separation of the two series of trivalent elements will not be com­plete because of the similarities in the ionic radii among 4f and 5f elements.

• Ln(III) and An(III) are highly hydrated in aqueous media: 8 to 9 water molecules can be numbered in their inner-coordination spheres, as com­pared to 4 to 5 only in the case of penta- and hexavalent actinides. It is, however, admitted, although difficult to demonstrate by a structural proof, that an outer-coordination sphere of water molecules exists and interacts with the water molecules present in the inner-coordination spheres of the metallic cations through hydrogen bonds.

• As for other metallic cations, hydration of the 4f and 5f trivalent ele­ments is of capital importance in their extraction mechanisms, since they can be partly or completely dehydrated while being extracted in organic solvents.

• The coordination numbers in complexes of trivalent lanthanides and actinides vary from 6 to 12, depending on the bonding chemical system involved.

However, a slight chemical behaviour difference does exist between the two series of trivalent elements: the 4f orbitals of the lanthanides are slightly more localized around their nuclei than the 5f orbitals of the actinides, which can consequently interact more easily with their electronic environ­ments than the corresponding lanthanides (Beitz, 1994, Morss, 1994). Unlike trivalent lanthanides, trivalent actinides create stronger chemical bonds with ligands bearing ‘softer’ donor atoms than oxygen, such as for instance sulphur or nitrogen (Musikas et al., 1983, Musikas, 1984). The drawback of hydrophilic and/or lipophilic compounds containing sulphur and/or nitro­gen atoms is their usually strong affinity for protons in acidic media.

Although more rational (considering the inventory of elements present), the direct and selective extraction of An(III) from PUREX raffinates has been the most challenging of the unresolved research topics radiochemists have addressed for the past 50 years throughout the world. This is why, except for specific single-step processes, such as SETFICS (Nakahara et al., 2007) or DIAMEX-SANEX (Madic et al., 2002) processes which will not be covered by this chapter, most of the strategies adopted to selectively recover trivalent minor actinides from PUREX raffinates show the same two-step process approach:

1. The co-extraction of the An(III) together with the Ln(III) in a front — head process, such as the TRUEX, DIAMEX, or TODGA processes, which make use of oxygen donor extractants (‘hard’ bases), such as carbamoyl phosphonate/phosphine oxide or diamide compounds, and specific scrubbings with hydrophilic masking agents to achieve the sep­aration of An(III) and Ln(III) from the rest of the fission products (FP).

2. The partition of An(III) from Ln(III) in a second cycle process, either through the selective stripping of the An(III) thanks to a hydrophilic highly selective ligand, or through the selective extraction of the An(III) thanks to a lipophilic highly selective extractant (both types of com­pounds bearing ‘soft base’ electron-donor atoms, the use of which is made possible by the lower acidity of the feeds coming from the front- head processes than that of the PUREX raffinates).

In high temperature gas-cooled reactors (HTGR), also known as fast reac­tors, graphite is utilized as the moderator of the nuclear reaction. The graphite is either used as part of the structural materials for the reactor

Steps SF ILW LLW Tailings Comments Mining and milling 65,000 In terms of radiation doses and numbers of people affected, uranium mining has been one of the most hazardous steps in the nuclear fuel chain, disproportionately impacting indigenous communities. Conversion — — 32-112 — Besides airborne and waterborne uranium, hazards include chemicals such as hydrofluoric acid, nitric acid, and fluorine gas. Enrichment 3-40 Typically buried at dump sites with a high risk of leaching radionuclides into the groundwater. Waste is contaminated with polychlorinated biphenyls (PCBs), chlorine, ammonia, nitrates, zinc and arsenic. Fuel fabrication " " 3-9 " Because fuel fabrication does not involve the production of liquid waste, its effects are mainly restricted to workers and are on the same order as for workers in the reprocessing sector. Reprocessing and vitrification not applicable not applicable not applicable not applicable Wastes from reprocessing, together with spent fuel, contain more radioactivity than any other waste in the fuel cycle. Phenolic and chlorinated compounds are produced in large amounts due to the use of decontamination reagents such as CCI4 together with phenolic tar (Gad Allah, 2008). Reactor operations — 22-33 86-130 — Boiling water reactors have considerable emissions of radioactive noble gases. Spent fuel storage and encapsulation 2 0.2 Considerable quantities of "low-level" waste are created due to fission products leaking into the spent fuel pools from cracks in the fuel cladding (Choi eta/., 1997). Spent fuel final disposal 26 — — — Insufficient treatment can cause continued exposure to environment and local population. Decommissioning — 9 333 — Most of the radioactivity from reactor decommissioning waste is in a relatively small volume of intensely radioactive material. Totals 26 33-44 457-624 65,000

Steps SF HFW IFW FEW Tailings Comments Mining and milling 50,060 Mill tailings account for over 95% of the total volume of the radioactive waste from MOX-OT processing cycle. This does not include mine wastes. Many tailings sites all over the world remain unremediated and/or neglected and pollute ground and surface water with radioactive and non-radioactive toxic substances. Conversion — — — 25-86 — Besides airborne and waterborne uranium, hazards include chemicals such as hydrofluoric acid, nitric acid, and fluorine gas. Enrichment 3-25 Typically buried at dump sites with a high risk of leaching radionuclides into the groundwater. Waste is contaminated with polychlorinated biphenyls (PCBs), chlorine, ammonia, nitrates, zinc and arsenic. Fuel fabrication 13 7.4-12.5 Because fuel fabrication does not involve the production of liquid waste, its effects are mainly restricted to workers and are on the same order as for workers in the reprocessing sector. Reprocessing and vitrification 2-4 17-39 8016-8037 As in FUE-OT system, wastes from reprocessing, together with spent fuel, contain result in the highest risk. The waste is high inorganic content. There is a particularly high risk of further contamination through accidents of storage facilities at the reprocessing plant. Reactor operations — — 22-33 86-130 — Boiling water reactors have considerable emissions of radioactive noble gases. Spent fuel storage and encapsulation 0.3 0.03 As in the FUE-OT system, large quantities of "low-level" waste are created due to fission products leaking into the spent fuel pools from cracks in the fuel cladding. Fisson products are trapped in resins in filters, which then become "low-level" waste in the United States and intermediate level waste in Europe. Spent fuel final disposal 26 — — — — Insufficient treatment can cause continued exposure to environment and local population. Decommissioning — — 10.1 315 — Most of the radioactivity from reactor decommissioning waste is in a relatively small volume of intensely radioactive material. Totals 26 2-4 62-95 8452-8615 50,060

core vessel or as fuel containment elements in the form of pebbles. The graphite used from natural sources contains non-carbon impurities within the carbon matrix. Among these impurities are oxygen and nitrogen from entrapped air, cobalt, chromium, calcium, iron, and sulfur (Khripunov et al, 2006). Upon exposure to high neutron flux, most of the impregnated impuri­ties are expected to transmute to unstable radioactive forms. For example, experimental exposure of graphite in nuclear reactors have shown that the stable forms of oxygen, nitrogen, and C-12 are converted to radiocarbon-14 (C-14) as shown in Table 15.3.

The radioactive fission products are created within the fuel grains and migrate through grain boundaries and then through microscopic cracks in the graphic matrix (Fig. 15.2). Most of the fission products are entrained in the matrix — a small proportion escapes through the outer layers into the gas phase. The challenge of reprocessing involves the separation of the metallic radionuclides from the graphite matrix and reducing the amount of C-14. Impurities in the fuel itself include: (1) metallic fission products (Mo, Tc, Ru, Rh, and Pd) which occur in the grain boundaries as immiscible micron to nanometre-sized metallic precipitates (e-particles); (2) fission products that occur as oxide precipitates of Rb, Cs, Ba, and Zr, and (3) fission products that form solid complexes with the UO2 fuel matrix, such as Sr, Zr, Nb, and the rare earth elements (Kleykamp, 1985; Shoesmith 2000; Buck et al., 2004; Bruno and Ewing, 2006).