The accounting methods currently applied for pyrochemical experiments are conventional destructive analyses such as ICP-AES, ICP-MS and у spectroscopy for concentration analyses, and weight measurements for bulk analyses. Because destructive analyses usually require a long time to prepare the liquid solution from salt or metal samples, it is important to develop a method for on-line monitoring of the concentrations of elements in molten salt. It might be difficult to measure the weights of materials in a large vessel; thus, a method of measuring bulk amounts using level probes and density equations to calculate weights from volumes must be developed. It is also important to develop sampling technology to decrease sampling errors in pyrochemistry. In this section, the development of monitoring techniques specific to pyrochemistry is summarized.

Separation of An(III) from Ln(III) elements

Separation of Am and Cm, known as the minor actinides (MA), from the fission Ln elements is a key goal of several countries seeking to substantially reduce the mass of radionuclides requiring long-term disposition in a geo­logical repository. The implementation of a partitioning and transmutation technology necessitates the removal of the Ln elements because of rela­tively large neutron capture cross sections and incompatibility with pro­posed transmutation targets. The separation of the MA from Ln is, however, a difficult chemical separation to achieve as there are minor differences in the chemical properties of these elements. Studies in recent years have shown that ligands containing soft donor atoms such as N and S offer a higher probability of being more selective for An(III) elements (Kikuchi et al., 2004, Zhu, 1995, Modolo et al., 1998b, Wei et al., 2000a, Iki et al., 1998, Morohashi et al., 2001, Mathur et al., 2001). Although the exact mechanism for this preferential complexing behavior has not been entirely elucidated, many researchers are investigating the merits of using soft donor ligands in solid-phase extraction to perform An(III) separations from Ln(III) ele­ments in high level liquid waste (HLLW) solutions subsequent to upstream removal of U and Pu.

The N-donor ligand, 2,6,-bistriazinylpyridine (BTP) has shown promise for separating An(III) from Ln(III) via differing nitrate dependencies in

solutions of moderate acidity (pH ~2). The solubility of the ligand can also be modified via the addition of alky chains. The functionalized forms 2,6-bis — (5,6-dibutyl-1,2,4-triazine-3-yl)-pyridine (n-Bu-BTP or tro-Bu-BTP) have been synthesized (Fig. 13.7) and a solid-phase extraction material devel­oped by impregnating n-Bu-BTP into a silica-co-polymer support (Wei et al., 2004a, Wei et al., 2000b, Hoshi et al., 2006).

The silica co-polymer (SiO2-P) support is prepared by polymerizing a mixture of formylstyrene and divinylbenzene monomers saturated within the pores of spherical silica particles resulting in 17.6% (w/w) co-polymer embedded in SiO2. The ко-Bu-BTP ligand (2.5g) was incorporated into the SiO2-P support (5g) via the dry impregnation method and the material tested in batch and small (20 cm3 bed) column tests (Hoshi et al., 2006). Excellent decontamination of Am, Cm, and heavy Ln(III) elements from fission products and light Ln(III) elements was achieved using a simulated HLLW solution (w/o U, Pu) containing 1 M NaNO3-0.01 M HNO3. Heavy Ln(III) elements were removed from the column with 0.3 M NaNO3-0.01 M HNO3. The strongly sorbed An(III) elements were then eluted from the column by reducing the nitrate concentration with pure water. The resulting separation factors between Am and Ln(III) are listed in Table 13.2.

The observed nitrate dependency of An(III) and Ln(III) extraction by BTP may thus be exploited by the addition of nitrate salts and is quite promising. Batch tests were also performed to examine the stability of BTP against nitric acid. The authors (Hoshi et al., 2006) determined that

13.8 Chemical structure of Cyanex-301 ligand.

extractant losses increase with increasing nitric acid concentration due to protonation and that branched Ao-Bu-BTP shows more acid stability than n-Bu-BTP. The acid effects may not be an insurmountable issue since the material is used at very low acid concentrations.

A biopolymer microcapsule containing Cyanex 301 (bis(2,4,4-trimethyl — pentyl)dithiophosphinic acid) extractant has been reported by Mimura et al., (2001). The solid phase extraction resin was prepared by mixing the extractant with 1-2.5% (w/w) sodium alginate and then adding drop-wise to 0.5 M nitrate salt solution or 1 M HCl. The beads were allowed to harden in the bath overnight and then filtered and dried. The resin was tested in a small (4.9 cm3) chromatographic column and the researchers were able to achieve a separation factor of 20 between Am and Eu in a pH range of 1-2 using freshly made resin (<1 day old).

A satisfactory separation of Am from Ln(III) elements with newly pre­pared Cyanex 301-SiO2-P resin has also been reported (Wei et al., 2000b). However, similar testing with 4-week old resin did not produce an accept­able separation between Am and Ln(III). These authors concluded that oxidation impurities formed in the Cyanex 301 are responsible for the poor separation and that methods for stabilizing the extractant are needed. The structure of the Cyanex 301 compound is shown in Fig. 13.8.

A solid-phase extraction resin for separating MA from Ln(III) elements has been prepared by impregnating p-tert-butylthiacalix[4]arene com­pounds in a SiO2-P support (Kikuchi et al., 2004) The structure of the ligands (CAPS, CAPS-SO2) is shown in Fig. 13.9.

Batch contacts performed with the CAPS, CAPS-SO2-SiO2-P resin did not show any measureable extraction of Am or Ln elements at pH = 2. However, raising the solution pH to 4 resulted in a separation factor of 10 for Am over Nd and Eu with the CAPS-SiO2-P resin. A separation factor of 500 for Am over Nd and Eu was achieved with CAPS-SO2-SiO2-P resin in tests at the higher pH. A proposed flow sheet for separating the MA is shown in Fig. 13.10.

The effects of gamma radiation on SiO2-P particles loaded with Cyanex 301, CAPS and CAPS-SO2 were evaluated by irradiating the resin particles to a dose of 1 MGy in a weak acid (pH = 4) at ambient temperature (Kikuchi et al., 2004). The irradiation caused a 30% degradation of the Cyanex 301 exchanger and an increased retention of Eu, thus substantially

13.9 Chemical structure of CAPS and CAPS-SO2 ligands.

Minor Ac & Ln feed solution

reducing the separation ability between Am and Eu. The irradiation resulted in only 1% degradation of CAPS-SO2-SiO2-P exchanger and the high sepa­ration factor of Am relative to Nd and Eu remained constant during and following irradiation. Irradiation of the CAPS-SiO2-P exchanger also pro­duced an increased distribution coefficient for Am. The authors suggest that the sulfur in the CAPS ligand was oxidized during irradiation to become sulfonyl groups, i. e. fully or partially transforming it to CAPS-SO2, which they postulate is the factor contributing to a higher Am selectivity. Although confirmatory analytical data are not given, the hypotheses are certainly credible and merit more detailed investigation.

The removal of substrates and other materials from solution in a fixed-film system involves complex physical and chemical processes which include: film transport, pore surface transport, adsorption reaction, and growth of biofilm and suspended biomass (Thacker et al., 1981; Weber and Chakravorti, 1974). Expressions derived from biofilm kinetics are best solved by simpli­fication by using empirical approaches. The mechanism of substrate removal by fixed-biofilm is shown in the simplified model of biofilm-on-inert-media in Fig. 15.11. The complexity of resolving mathematical equations expressed by the biofilm model lies in recognizing the critical features, namely:

(i) The moving boundary problem

The biofilm thickness (Lf) can change while substrate diffuses into the biofilm during transient-state when the rate of substrate utilization is

not constant. The moving boundary problem is analogous to the problem of heat transfer during freezing and melting of ice in ther­modynamics (the “Stefan” problem, Danckwerts, 1950), and the diffu­sion of oxygen in absorbing tissue (Crack and Gupta, 1972); and (ii) Diffusion with nonlinear reaction

At any particular time, the rate of substrate diffusion across the liquid/ biofilm interface as described by Fick’s law, is equal to the rate of substrate utilization in the biofilm governed by the nonlinear Monod kinetics.

Figure 15.11 represents removal of dissolved species y in a multispecies biofilm contain microbial cells x, where x and y are vectors of cell types and dissolved species. In the above example, an organic compound P is used as primary carbon source to support a culture containing metal reducers XE and the organics degrader Xp. In this particular system, metabolites U are produced inside the biofilm to feed the metal reducers. The metal reducers are assumed to be unable to grow on the primary supplied carbon source P.

The solution chemistry of the actinide elements has been investigated in both aqueous and selected organic solutions. The majority of the informa­tion available describes species found in aqueous solutions. Although

actinide cations can exist in a variety of oxidation states (2+ to 7+) in aqueous solution (Table 2.2), the most common for light actinides in aqueous acidic solutions are trivalent, tetravalent, pentavalent and hexavalent oxida­tion states. The stability of a particular oxidation state across the actinide series is quite variable, and for some actinides (Np, Pu) several oxidation states can coexist in the same solution. This is most evident for plutonium as there are small differences in the reduction potentials of Pu(III), Pu(IV), Pu(V), and Pu(VI) over a range of pH values (Choppin, Jensen, 2006).

S. A. BRYAN, T. G. LEVITSKAIA, A. J. CASELLA, J. M. PETERSON, A. M. JOHNSEN, A. M. LINES, and E. M. THOMAS, Pacific Northwest National Laboratory, USA

Abstract: Separation processes for highly radioactive and chemically complex spent nuclear fuel require advanced safe technologies and process models based on large databases. Availability of advanced methodologies for on-line control and safeguarding of aqueous reprocessing flowsheets will accelerate implementation of the closed nuclear fuel cycle. This report reviews application of the absorption and vibrational spectroscopic techniques supplemented by physicochemical measurements for radiochemical process monitoring. In this context, our team experimentally assessed potential of Raman and spectrophotometric techniques for on-line real-time monitoring of the U(VI)/nitrate ion/nitric acid and Pu(IV)/Np(V)/Nd(III), respectively, in the solutions relevant to spent fuel reprocessing. Both techniques demonstrated robust performance in the repetitive batch measurements of each analyte in the wide concentration range using simulant and commercial dissolved spent fuel solutions. Static spectroscopic measurements served as training sets for the multivariate data analysis to obtain partial least squares predictive models, which were validated during on-line centrifugal contactor extraction tests. Achieved satisfactory prediction of the analytes concentrations in these preliminary experimentation warrants further development of the spectroscopy- based methods for radiochemical process control and safeguarding.

Key words: spectroscopic process monitoring, Raman, vis-NIR, uranium, plutonium, neptunium, nitric acid, nitrate, nuclear fuel reprocessing.

4.1 Introduction

There is a renewed interest worldwide to promote the use of nuclear power and close the nuclear fuel cycle under the Global Nuclear Energy Partnership (GNEP) Program, the Advanced Fuel Cycle Initiative (AFCI), and more recently under the Fuel Cycle Research and Development Program (FCRD). The long-term successful use of nuclear power is critically depen­dent upon adequate and safe disposal of the spent nuclear fuel, and imple­mentation of the closed nuclear fuel cycle recently regained attention in the

US community. Liquid-liquid solvent extraction is a separation technique commonly employed for the processing of the dissolved spent nuclear fuel. Availability of the closed nuclear fuel cycle requires precise solution control during processing. For example, extraction and recovery of minor actinides within an aqueous separation scheme necessitates precise control of aqueous redox conditions and acid concentrations at various places within the pro­cessing loop. Traditionally, the process performance of a given solvent extraction run is determined through sampling of the various process steams and subsequent laboratory analysis of those samples. Due to the highly radioactive nature of the process streams, this is a risky and time-consuming process. Remotely controlled on-line monitoring capabilities can help guide choice and adjustment of conditions to be used in processing of irradiated fuel and allow immediate feedback on changes made to the processing conditions. As a result, there is a renewed and urgent need for methods to provide on-line monitoring and control of the radiochemical processes cur­rently being developed and demonstrated. The instrumentation used to monitor these processes must be robust, require little or no maintenance, and be able to withstand harsh environments (e. g., high radiation fields and aggressive chemical matrices). The ability for continuous online monitoring allows the following benefits:

• accountability of the fissile materials

• control of the process flowsheet

• information on flow parameters, solution composition, and chemical speciation

• enhanced performance by eliminating the need for traditional analytical “grab samples”

• improvement of operational and criticality safety

• elimination of human error.

Sophisticated on-line monitor capabilities significantly enhance not only control over a process flowsheet but also accountancy of the inventory of a nuclear material. The increasing effectiveness of safeguards in spent fuel reprocessing plants is a great challenge to national and international communities. The International Atomic Energy Agency (IAEA) has estab­lished international safeguards standards for fissionable materials at repro­cessing plants to ensure that significant quantities of weapons-grade nuclear material are not diverted over a specified time frame. Because proliferant diversions are possible via deliberate modification of flowsheet chemistry, it is necessary to confirm proper operational performance to verify that facilities function under adequate safeguard-declared conditions. In any reprocessing facility, variability in process is expected under normal opera­tions, and currently, large-scale deviations can readily be detected and measured with certainty. Small-scale deviations in large facilities, however, are currently not well detected. Many of the methods useful for detecting deviations involve sending individual samples to a laboratory for process­ing, which can be time-consuming, thus delaying accurate appraisals of process conditions.

The application of multiple online monitoring capabilities provides a unique ability to rapidly identify unwanted/suspect deviations from normal operating conditions. The feasibility of this kind of on-line control of nuclear fuel reprocessing streams via analytical techniques was investigated as early as the 1970s (Parus 1977), and researchers have examined both the direct measurement of actinides, via spectrophotometry, and the use of physico­chemical measurements, such as temperature, density, and dielectric proper­ties, to indirectly measure actinide concentrations. These analytical methods can also be used to measure other solution components, such as NO3- or organic solvents that control actinide behavior, providing another method for actinide quantification.

Raman spectroscopy (Madic 1983,1984; Guillaume 1982) and ultraviolet — visible (UV-vis) spectroscopy (Schmieder 1972; Ertel 1976, 1985; Baumgartner 1980; Yamamoto 1988; Burck 1991; Colston 2001) are analyti­cal techniques that have been used extensively to measure concentrations of various organic and inorganic compounds, including actinides (Bryan 2007). Additionally, measurement of dielectric properties has also been proposed for on-line monitoring of fuel reprocessing systems (Yamamoto 1988).

The spectroscopic signatures of U, Pu, and Np analytes in different oxida­tion states have been extensively measured in solution and widely known (Madic 1983; Guillaume 1982; Maya 1981; Nguyentrung 1992). Even though Pu(VI) (Madic 1983) and Np(VI) (Guillaume 1982) species are Raman active, they are expected to be present in the dissolved fuel at low enough concentrations to prohibit their quantification by the Raman method. In our work, we employ Raman spectroscopy for the determination of U(VI), HNO3, and NO3- in various aqueous and organic streams. Plutonium species can be measured by visible absorption spectroscopy using multiple wave­lengths for its quantification (Ryan 1960; Cleveland 1979) and neptunium species can be monitored by vis-NIR spectroscopy (Burney 1974; and Stout 1993).

Armenta et al. reviewed the most recent literature (2000-2006) concern­ing the general (non-actinide applications) combination of flow-injection techniques and vibrational spectroscopy analysis and noted that both tech­niques have significant advantages. Flow-injection analysis is the technique in which a sample aliquot is injected into a reagent stream, where it dis­perses and the mixture flows past a detector. Flow-injection analysis offers “automated sample processing, high repeatability, adaptability to micro­miniaturization, containment of chemicals, [and] waste reduction,” while vibrational spectroscopy allows for “(i) fast monitoring of the whole spec­trum; (ii) high resolution and wide wavenumber range; (iii) many bands that can be employed for determination of each single compound; [and] (iv) simultaneous control of several compounds in the same sample.” These characteristics have allowed several commercial instruments to be applied successfully to quality and process control in the dairy, wine, gasoline, diesel, and lubricating oils industries. The authors also note that new flow-injection techniques in development have the potential to lower sample sizes and increase sensitivity (Armenta 2007).

Multiple laboratories have successfully constructed systems for remote in-line photometric measurements of actinides using spectrometers capable of scanning the entire visible and NIR range quickly enough for process measurements (Burck 1991). On a larger scale, Lascola et al. (2002) success­fully constructed a system for the on-line measurement of uranium from process tanks. Tank samples were routed through an initial vial to ensure proper mixing and were then sent through a flow-through optical cell con­nected via fiber optics to a diode-array spectrophotometer measuring in the 350-600 nm range. Partial least squares (PLS) models allowed for concen­tration measurements up to 11 g/L uranium, as well as uncertainties (2 a; concentration dependent) no greater than 0.30 g/L.

Several authors have conducted density studies of typical actinide repro­cessing solutions for use in criticality and process monitoring. Sakurai and Tachimori (1996) analyzed published density data for solutions containing plutonium(IV), uranium(VI), and nitric acid. Using regression analysis, they correlated solution density with analyte concentration that expanded the range of actinide concentrations to Pu < 173 g/L and U < 380 g/L and pro­vided standard error (0.00294 g/cm3) that improved upon the error associ­ated with previously established relationships. Kumar and Koganti (1998) extended this work to include mixed organic solutions, reporting density for UO2(NO3)2 and HNO3 in TBP/n-dodecane solutions ranging from 0 to 100% tri-butyl phosphate (TBP).

While Sakurai and Kumar have reported the most recent empirical density relations that are considered to be the most accurate empirical equations thus far, several authors have pursued theoretical calculations of actinide/nitrate solution densities in order to clarify the density functions at higher concentrations where the empirical equations begin to deviate from experimental data. Thermodynamic modeling for binary (or isopies — tic) solutions have been applied to make small but important corrections for solutions with high concentrations of the following components: UO2(NO3)2, U(NO3)4, Pu(NO3)4, Pu(NO3)3, Th(NO3)4, Am(NO3)3, and

HNO3 (Charrin 2000a, 2000b; LeClaire 2003).

Enokida and Suzuki (1992a, 1992b) used a set of computer models to theoretically test the feasibility of using temperature profiles to determine the uranium(VI) concentrations during solvent extraction processes from nitric acid into 30% TBP/n-dodecane organic phase. Using literature values for the heat of the TBP complexation, along with mass and heat balance equations, the authors created a model that calculated both steady — and transient-state uranium(VI) concentrations when given a temperature profile and a set of flow rates and feed concentrations for a typical set of counter-current mixer-settler extractors. Their model compared favorably with other models that calculated uranium(VI) concentrations using dis­tribution coefficients, which showed that using temperature profiles at various process stages is a feasible means of uranium(VI) concentration measurement.

Yamamoto (1988) used a flow-through cell system to measure the dielec­tric properties of the 30% TBP/n-dodecane-HNO3-H2O and 30% TBP/n — dodecane-HNO3-H2O-UO2(NO3)2 systems commonly used in spent nuclear fuel reprocessing. The estimated dielectric constants were found to vary much more significantly with HNO3 than with equivalent molar additions of H2O or UO2(NO3)2. The measurements were found to be sufficient to accurately measure the HNO3 concentration in the (30% TBP/n-dodecane)- HNO3-H2O system and could be used to measure the HNO3 concentration in the (30% TBP, n-dodecane)-HNO3-H2O-UO2(NO3)2 system if the UO2(NO3)2 concentration is known.

On-line monitoring of nuclear waste streams was successfully demon­strated by combining spectroscopic measurements with physicochemical measurements (conductivity, density, and temperature) in real-time quanti­tative determination of chemical components in the waste (Bryan et al. 2005; Bryan 2008). This new on-line monitoring system, which features Raman spectroscopy combined with a Coriolis meter and a conductivity probe, was developed by our research team to provide immediate chemical data and flow parameters of high-level radioactive waste streams. This process moni­toring system was used to measure the concentration of components of high brine/high alkalinity waste solutions, such as nitrite, chromate, aluminate, phosphate, sulfate, carbonate, and hydroxide, during retrieval from Hanford waste storage tanks. The Raman bands of interest for these species are well resolved and have been easily incorporated into a chemometric model for quantitative analysis of the solution components.

By inclusion of visible/near-infrared (vis-NIR) spectrocopy, this system was modified to monitor spent fuel reprocessing streams. A fiber-optic Raman probe allows monitoring of various species encountered in both aqueous and organic phases. Raman active species include: 1) metal oxide ions, such as uranyl, neptunyl, and pertechnetate ions, 2) organics, 3) inor­ganic oxo-anions, and 4) water. The trivalent and tetravalent actinides and lanthanides in both the aqueous and organic phases are monitored by vis-NIR spectroscopy, as are pentavalent or hexavalent Np and Pu complexes that are expected to be at concentrations too low to be deter­mined by Raman spectroscopy. Process monitoring and control is feasible at various points within the fuel reprocessing streams. Notably, process monitoring/control is not specific to any single flowsheet because alternate flowsheets also contain Raman and/or UV-vis-NIR active species that can be measured spectroscopically. In addition, Coriolis (for density and flow) and conductivity instruments can be used on most process streams.

The Raman and vis-NIR spectrometers used under laboratory conditions are easily convertible to process-friendly configurations to allow remote measurements under flow conditions using different sampling capabilities, such as fiber-optic probes, dip probes, and flow-through cell geometries as have been demonstrated by our research team in the centrifugal contactor flow tests. Spectroscopic data collected during the flow test served for the validation of the chemometric predictive models developed under static batch conditions as described in Sections 4.2 and 4.3.

The long and successful history of PUREX begs the question of its utility in the future of nuclear energy. In the current climate of developing a truly “closed,” “next generation,” or “advanced” nuclear fuel cycle, it is highly probable the first incremental improvements would be initially based on the capabilities of the current industrial standard, i. e. PUREX. Hot topics regarding the concepts of future fuel cycles are multi-faceted, but could be broadly categorized as containing two objectives: proliferation resistance (i. e., “no pure Pu” separation) and environmental concerns relating to radiotoxicity and disposal of the nuclear waste. The important capabilities of PUREX are the separation of major actinides prevalent in used nuclear fuel, primarily U, Pu, and Np (as well as the long lived and environmentally mobile fission product Tc). The next generations of PUREX processing will likely involve modification in the process chemistry to recover Np for recycle with the U and Pu into future mixed oxide (MOX) fuels and pre­clude the separation of a pure Pu stream by keeping some fraction of the U and/or Np with the Pu product. Another potential use of PUREX would be the separation of the minor actinides (MA), notably Am, by making use of the accessibility of higher oxidation states, albeit this use of PUREX is conceptual and much further down the road and consequently will not be further discussed here.

a 5 = superior; 4 = good; 3 = average; 2 = below average; 1 = poor. b Process flexibility includes such factors as the range of O/A flow ratio, the turndown in flowrate, and the ease with which the location of feed and product streams can be changed.

c Considered an advantage when process chemistry and kinetics requires long residence time.

d Considered an advantage when solvent degradation is a concern.

The process details discussed above indicate that there are several reasons to improve the current PUREX process. The distribution of neptunium between product streams that require additional decontamination, espe­cially the uranium product, is one of the major concerns. Currently, Pu is separated in high purity, converted to an oxide and subsequently back blended with U to prepare MOX. Intuitively, separating a combined U and Pu product (preferably in the appropriate and controllable ratios for MOX fuels) would substantially simplify the overall process, decrease the number of unit operations, and alleviate proliferation concerns. Also, there is inter-

est in introducing centrifugal contactors in advanced separation cycles due to the benefits of short residence time, small footprint, lower holdup volumes, and decreasing the risks from criticality issues. This technology obviously limits the chemical reactions to those with fast kinetics putting further constraints on the future use of PUREX processes.

Much work has been carried out in the United Kingdom and in France to improve PUREX into an “Advanced PUREX” or COEXTM process, respectively. This work has aimed at bleeding U (and possibly Np) into the Pu product to eliminate the pure Pu stream and to reduce the number of solvent extraction cycles. Furthermore, by controlling the neptunium oxida­tion state and providing a single route for the neptunium, the uranium purification cycle could conceivably be eliminated, simplifying the process and effectively decreasing the amount of waste (Taylor 1997). Using similar means to achieve an improved PUREX process, the path considered in the US is slightly different due to other political goals (Laidler 2001). The pos­sibility to dispose of uranium as low-level waste (Vandegrift 2004) without the need for uranium purification cycle(s) would require thorough decon­tamination of the uranium from plutonium and neptunium. This route has been dubbed the “UREX” process since the goal is URanium EXtraction and the process is very similar to the PUREX process but eliminates plu­tonium extraction such that the Pu would follow the remaining transuranic actinides.

162 Advanced separation techniques for nuclear fuel reprocessing

Semi-volatile species, which include the elements I/Br, Ru, Rh, Tc, Cs, Rb, Mo, and Se/Te, can be released under certain processing conditions. Early removal of the semi-volatiles simplifies the off-gas treatment compared with recovery from wet NOx-laden off-gas generated from the downstream dissolution process. Iodine is difficult to fully remove from liquid dissolver solutions and can be problematic as a corrosive. Also, the early removal of specific semi-volatiles can simplify the downstream separations processes; for example, the recovery of technetium by means of dry volatilization will eliminate the need to co-extract it with uranium and separate the technetium under conditions that may complicate the uranium recovery. Moreover, the recovery of molybdenum by means of dry volatilization will eliminate the major source of precipitates (e. g., zirconium molybdates) that can foul the evaporation and solvent-extraction equipment.11

In the basic voloxidation process (oxidation in air at 480°C), less than about 0.2% of each of the semi-volatiles, 106Ru, 125Sb, and 134-137Cs, is evolved. Higher temperatures and vacuum operation increase the fraction evolved.

Collaborative tests by Idaho National Laboratory (INL), the Korean Atomic Energy Research Institute, and Oak Ridge National Laboratory (ORNL) showed that nearly 100% of the semi-volatiles can be separated under a variety of conditions using air or oxygen as the oxidant atmosphere, temperatures up to 1250°C, and vacuum operation. The removal effective­ness was highly dependent on the processing conditions. In general, initial oxidation at relatively low temperature (around 500°C) to generate the fine fuel powder with increased surface area is beneficial. Increasing tempera­tures enhanced the amounts of the semi-volatiles released in the following approximate order: Ru + Rh, Tc, Cs, Te, and Mo.1213

Enhanced oxidation, at selected temperatures and with alternative reagents such as ozone, NO2, and steam has been tested at ORNL to com­plete the removal of volatiles (all xenon, krypton, carbon, and iodine) and to remove semi-volatiles (molybdenum, technetium, and others). At low temperature (200-300°C), the U3O8 powder obtained by conventional voloxidation may be further oxidized with ozone (O3) to produce a finer UO3 powder. Also, NO2 can be used and will oxidize UO2 or U3O8 to UO3. The fine UO3 powder may be heated to higher temperatures, in a secondary step if necessary, to remove those species released by diffusion-controlled processes. Although higher temperatures cause UO3 to revert to U3O8, the benefit of the finer particles will remain. Additionally, the process can be cycled between UO3 and U3O8. Further fracture of the particles will enhance the release of volatile species and can be accomplished by cycling between higher and lower oxide species.

For example, in the AIROX1415 and DUPIC161718 processes, the fuel is first subjected to oxidation in air to obtain U3O8 powder. The resulting powder is then subjected to repeated cycles of hydrogen reduction to UO2 (usually at temperatures between 600°C and 800°C) and then re-oxidization of the UO2 to U3O8. Alternatively, a cycle between U3O8 and UO3 can be carried out using O3 or NO2 to generate UO3, then heated above 450°C to decompose the UO3 to U3O8. This is a lower temperature cycle that avoids any potential complication due to alternating hydrogen and oxygen atmospheres.

10.1.1 The pyrometallurgical process for metal fuels

There are different versions of the pyroprocess depending on the material to be treated. Figure 10.1 shows a flowchart of the pyroprocess for repro­cessing a metal fuel, i. e. U-Pu-10wt%Zr alloy for the core fuel and U-10wt%Zr alloy for the blanket fuel of a fast neutron reactor. In this section, the principles of each step of the pyroprocess will be described.

Disassembly and chopping

In the ‘disassemble and chopping’ process, hexagonal spent fuel assemblies are disassembled, and fuel elements with stainless steel cladding are chopped into small segments. The technologies developed for oxide fuel assemblies can be applied in this process.

Electrorefining

The next step is the electrorefining process, schematically shown in Fig. 10.2, where the chopped spent metal fuels are dissolved in an anode basket,

10.1 Process flowsheet of pyroprocess for metal fuel reprocessing. (Revised by the author according to the latest experimental results.)

whilst actinide metals are recovered at a solid cathode, or a liquid Cd cathode. The main reaction schemes of electrorefining are described as follows:

Anode — Uin spent fuel ^ U + 3e 10.1

Puin spent fuel ^ Pu3+ + 3e — 10.2

Cathode (solid): U3+ + 3e — ^ U 10.3

Cathode(liquid Cd): U3+ + 3e — ^ U in Cd 10.4

Pu3+ + 3e — + 6Cd ^ PuCd6 10.5

This process utilizes the oxidation-reduction potentials of the relevant elements shown in Table 10.1. The constituents of the spent fuel that have lower standard potentials than zirconium, i. e. actinides and less noble FP (LFP) such as the alkali metal FP, the alkaline-earth metal FP and lantha­nide FP, are electrochemically dissolved at the anode. Conversely, elements having higher standard potentials, i. e. zirconium, iron (cladding), cadmium, and noble FP (NFP) such as ruthenium, rhodium, palladium, technetium and molybdenum, remain undissolved in the anode basket. At the solid cathode, uranium is preferentially reduced and collected since it is the most easily reduced element among the dissolved materials. Other actinides, such as plutonium, neptunium, americium and curium are recovered with uranium in the liquid cadmium cathode because the reduction potentials of these elements are very close to that of uranium, as shown in the right — hand column of Table 10.1, due to their strong affinity with cadmium, i. e.

they form intermetallic compounds such as PuCd6. Consequently, this gives a proliferation-resistant nature to the process, since it is almost impossible to separate out pure plutonium. Typical cathode products obtained in a laboratory-scale electrorefiner (Koyama, 2002) are shown in Fig. 10.3, where (a) is a uranium deposit on the solid cathode and (b) is a plutonium — uranium-MA deposit in the cadmium cathode. As shown in this figure, the actual cathode deposits contain adhering salt and/or alloying cadmium in addition to actinide metals, though the actinide cations are reduced into metal form at the cathodes.

Ester, ketone, and amide functions have been introduced on the ‘narrow rim’ of calix[n]arenes (Fig. 11.4) in order to mimic natural substrates (Chang

11.4 Examples of functionalized calix[n]arenes at the narrow rim.

and Cho, 1986, 1987, Arnaud-Neu et al., 1989). The trends observed while investigating the complexing properties of these ligands can be summed up as follows, although the experimental conditions often differ:

• The blocked cone conformation and the four ‘hard’ donor functions (bearing oxygen atoms) pointing in the same direction confer the sub­stituted calix[4]arenes interesting complexation and extraction proper­ties toward sodium cation, in the following order: esters < ketones < amides (Arduini et al., 1988, Schwing et al., 1989, Arnaud-Neu et al., 1991, 1995, Gradny et al., 1996). The stability constants of the 1 : 1 (metal : ligand) complexes measured in solution are greater than those observed with crown ethers, almost challenging those of cryptands, thus enabling the use of these substituted calix[4]arenes as active agents for chemical sensors (Forster et al., 1991, Brunink et al., 1991, Ganjali et al., 2006).

• The ester derivatives of calix[6]arene extract potassium better than sodium and show a selectivity plateau for voluminous alkali cations. The amide derivatives of calix[6]arene extract the alkaline-earth elements better than the alkali elements, with a preference for calcium and strontium.

• Calix[8]arene derivatives are the least efficient extractants of the series.

E. M. N. CHIRWA, University of Pretoria, South Africa

Abstract: Due to the impact of the rapidly growing demand for energy worldwide, as well as concerns over global warming, there has been a resurgence of interest in nuclear energy in the developed world. However, further deployment of this otherwise cleaner source of energy in other lesser-developed regions is hindered by concerns over accumulation of radioactive waste from nuclear reactor operation and fuel processing. This chapter discusses emerging biological technologies for treating radioactive waste, focusing on biological reduction and recovery processes that may in future improve the viability of this energy source. The processes discussed include biological reduction of uranium (VI), biosorption of fission products and isotopic biofractionation processes. These technologies offer a possibly cost — effective and environmentally friendly alternative to physical-chemical processes currently used for treating radioactive waste in the nuclear — power industry.

Key words: nuclear waste minimization, uranium (VI) reduction, Cr(VI) reduction, radioisotope bioseparation, cationic species biosorption, fission product recovery.

15.1 Introduction

Due to the impact of the rapidly growing demand for energy, as well as concerns over global warming, there has been a resurgence of interest in nuclear energy in the developed world. Nuclear energy is one of the few economically viable base-load electricity generation technologies, which avoids the production of about eight percent of the present level of CO2 emissions in the energy sector (Mourogov et al, 2002). Currently, there are about 438 nuclear power plants in operation in 31 countries around the world providing about 14 percent of the world’s primary energy needs (IAEA, 2009). The world’s nuclear generating capacity currently stands at about 372 GWe, with the United States of America and France as the major producers, 27 and 17 percent, respectively (Fig. 15.1).

Electricity consumption in developing countries, such as South Africa, has been steadily increasing since the 1980s and, currently, the electricity

0 5 10 15 20 25 30 35

15.1 Global nuclear power generating capacity (%) per country (IAEA, 2009).

consumption/capita is estimated at about 5039.7 kWh. It is predicted that by the year 2025 electricity demand will exceed supply (Musango et al., 2009). To accommodate future expansion in the domestic and industrial electricity consumption, there is need to develop safer, more efficient, and environmentally friendly technologies to replace the current fossil-fuel based power generation technologies.

In the developed world, most nuclear power plants in operation today have reached or are nearing their design life. Most of these power plants were constructed in the 1960s and 1970s. These need to be replaced by new, environmentally sustainable power generation technologies with improved safety features. An example of these new generation reactor systems is the Pebble Bed Reactor (PBR) technology, a Generation IV reactor technol­ogy that utilizes graphite as the neutron moderator. In this latter system, the reactor core is cooled by an inert gas such as helium instead of water (Koster et al., 2003). Because the reactor can be allowed to operate at higher temperatures than the conventional water cooled reactors, the efficiency of the system is greatly enhanced. The drawback is that impurities in the con­tainment material (graphite) are difficult to treat due to the inert nature of the graphite. This results in the accumulation of large volumes of low radia­tion level waste beyond the capacity of designated waste storage areas.

Potential radioactive pollution to the environment does not only concern nuclear power plants. Other activities such as radioisotope manufacturing
and biomedical research also release large amounts of potentially harmful radioisotopes. Most of the radioactive pollutants from the latter activities are organic in nature and are amenable to biological degradation (Cerniglia et al, 1984; Bouwer and Zehnder, 1993). However, due to the toxic nature of the waste stream, an additional effort is required to isolate specialized bacteria that are resistant to the toxic effects of the released compounds and that are capable of breaking the complex structures of the organic compounds (Tikilili and Chirwa, 2009).

The following section provides a concise review of the waste compounds originating from nuclear power generation and other radioisotope releasing activities and how these could be treated for beneficial use. Biochemical processes that have yielded positive results are presented as part of the review and their impact on the future of power generation is evaluated.