Category Archives: Advanced separation techniques for nuclear fuel reprocessing and radioactive waste treatment

The Argonne model for universal solvent extraction (AMUSE) for flowsheet design

The development of accurate predictive models is critical to the develop­ment of novel separations and to the design and refinement of existing processes. The flowsheets for each UREX+ process module were developed using the AMUSE code (Regalbuto et al., 2004). AMUSE is an updated version of the Generic TRUEX Model (GTM). The GTM was developed during the 1980s to design multistage countercurrent flowsheets for the TRUEX solvent extraction process (Vandegrift et al., 1993, 1995).

AMUSE predicts how components distribute among the aqueous and organic phases based on the compositions and characteristics of the aqueous and organic phases at given process conditions. It accomplishes this by calculating the distribution ratios, or D values, of the stream components. The predicted distribution ratios are highly accurate over a wide range of aqueous phase compositions due to the use of chemically correct models that are based on the thermodynamic activities. A countercurrent mass balance algorithm yields the final predicted elemental composition in the extraction system for both organic and aqueous phases.

The AMUSE code is composed of two major sections, SASSE and SASPE. The material balance calculations are performed in the SASSE section, which calculates multistage countercurrent flowsheets using calcu­lated distribution ratios. The SASPE section performs the distribution ratio calculations, for all of the components, based on elemental compositions and characteristics of the aqueous and organic phases. When AMUSE is used to calculate flowsheets based on the user defined input such as (1) equipment type, (2) feed compositions and locations, and (3) general flowsheet specifications (e. g., number of stages, temperature, flow rates) an iterative process among the SASSE and SASPE sections is performed as follows:

1. The SASPE section of the model will calculate distribution ratio (D) values from stage compositions generated by SASSE.

2. The SASSE section will then refine the calculated D-values based on the compositions calculated with SASPE.

3. The iterative process continues until convergence among the two modules is achieved.

4. The output report is generated that includes the compositions of all of the components in the organic and aqueous phases for each stage of the process.

AMUSE readily calculates flowsheets for PUREX, UREX, SREX, CCD-PEG and TRUEX as it can estimate the distribution coefficients for these processes internally. Flowsheets for other solvent extraction pro­cesses, including CSSX, FPEX, and TALSPEAK can be generated using appropriate user specified D-values. AMUSE is then used to perform sen­sitivity analysis on the designed process flowsheets in order to determine the effects of flow rate variation, compositional variability, other-phase car­ryover, temperature excursions, etc. The results of the sensitivity analysis establish the process operating envelope.

Flowsheet based on the UNEX process with selective recovery of the cesium fraction

If the selective recovery of the cesium fraction from HLW is required, the work flow involves the following operations:

• Combined extraction of Cs, Sr, An and REE by the UNEX-extractant (8 stages).

• Scrubbing of the extract with a solution of 0.5 M citric acid (1 stage).

• Stripping of Sr, An and REE by a solution of 0.5 M guanidine carbonate and 20 g/l DTPA (5 stages).

• Stripping of cesium by a solution containing 5 M HNO3 and 100 g/l acetamide (6 stages).

• Regeneration of the extractant by a 10 M HNO3 solution (3 stages).

• Regeneration of the extractant by a 0.1 M HNO3 solution (1 stage).

Figure 9.4 represents a work flow with selective withdrawal of cesium which has been proven under dynamic conditions at RI and Idaho National Laboratory.

Electroreduction cell

The electroreduction of oxide fuel is carried out by removing oxygen from the oxide fuel at the cathode and generating oxygen gas at the anode. Most studies are mainly focused on obtaining design parameters for the anode and cathode, to achieve high throughput and high reduction yield. In the case of a cathode to charge spent oxide fuel; however, the basket has con­tradictory design requirements. According to visual examination of incom­plete reduction products, the reaction proceeds from the outside to the inside of the oxide particle. This suggests that diffusion in the grain bound­ary or in cracks in the particle is the rate-determining step. To complete the reaction earlier, a cathode basket with smaller holes is better for charging smaller oxide particles. On the other hand, a cathode basket with larger holes is required to accelerate the circulation of molten salt to diffuse oxygen ions from the inside to the outside of the basket. With regard to the importance of particle size, KAERI has developed an engineering-scale electroreduction cell using a porous MgO crucible (18 cm ID x 42 cm H) as the cathode basket, surrounded by six platinum anode rods (3 cm OD x 30 cm H). Figure 10.25 shows an electroreduction cell of 50 cm ID, which has been used only for on-irradiated materials. When 10 kg of U3O8 powder (10-30 ^m) was charged in this electroreduction cell, more than 99% of the reduction reaction was completed in about 100 h (Jeong, 2008). As the concentration of Li2O in the molten salt decreased with time, the permeabil-


10.25 Engineering-scale electroreduction cell installed in hot cell of KAERI.

ity of oxygen ions through the MgO crucible was found to be very slow, suggesting that the holes in the MgO crucible were too small. However, INL employed a percolated stainless-steel basket to charge crushed spent oxide fuel particles with 0.45-2.8 mm diameter (Herrmann, 2005) but no decrease in Li2O concentration was observed, and it took 36 h to complete the reduction reaction of 41 to 50 g of spent fuel. INL continued reduction experiments using voloxidized spent oxide fuel powder of less than 45 ^m diameter in small baskets made of stainless-steel mesh, with sintered stain­less-steel frit (Herrmann, 2007). With either mesh or sintered frit stainless — steel baskets, a higher current can be passed for the powder form than for crushed particles. As an alternative, CRIEPI has proposed processing porous chunks from the voloxidized fuel powder for charging in a stainless — steel basket with a large opening. According to experiments, using roughly sintered UO2 pellets with a porosity of approximately 30%, the time to complete the reduction reaction was less than 10 h and was unaffected by an increase in the amount of fuel pellets from 10 g to 100 g (Sakamura, 2008). Although the procedure is more complex, it is one possible method for realizing practical throughput.

With regard to anode design for generating oxygen gas, developments have been mainly focussed on the durable anode material, which will be described in the next section. An effective evacuation of the oxygen gas from the vicinity of anode is another key issue for anode design.

Basic methodology of solid-phase extraction

First, let us provide a more descriptive definition of solid-phase extraction for the purposes of this discussion. Solid-phase extraction refers to a liquid, semi-liquid or potentially solid complexing agent dispersed more or less homogeneously within an inert, solid medium. A solid-phase extraction material comprises three major components: a solid support or substrate (typically inert polymer), a stationary extractant phase, and a mobile fluid phase (e. g., conditioning, feed, wash, or strip solution). The three phases of a solid-phase extraction resin are depicted in Fig. 13.1.

Various materials have been used for the solid support, e. g. inorganic materials such as silica gel, organic polymers such as polystyrene-divinyl — benzene copolymers and polymethacrylate resins, or resins made of both organic and inorganic materials. The stationary phase is typically an organic extractant, many of which have been well characterized for their use in liquid-liquid extraction. Similar to liquid-liquid extraction, the target metals are converted from the hydrated ionic form to a neutral organophilic metal complex within the stationary phase. The solid-phase extraction material is made in the form of small beads. Although different in terms of functional­ity, the beads are similar in size and shape to those used in conventional ion exchange practice and can be used in a fixed bed or column arrange­ment. The solid-phase extraction resin has many of the operational advan­tages of ion exchange processes and similar methods for modeling and engineering scale-up may also be used as discussed in a later section. The solid-phase extraction resin may, however, provide certain benefits over conventional ion exchange resins that are attractive to the separations sci­entist or chemical engineer. Potential advantages of solid-phase extraction are summarized as follows:


13.1 Three phases of solid-phase extraction resin.

• Selectivity: recent and continuing advances in the synthesis of organic extractants provide numerous options for highly selective An and Ln separations, most especially from dilute streams of high ionic strength.

• Solvent loading: the utilization of macroporous resins can yield higher specific mass loadings of the complexing molecule relative to functional­ized ion exchange resins.

• Cost: simple immobilization of complexing agents or extractants within an inert polymer may prove to be a less costly preparation route than the complex synthesis mechanisms of covalently linking a specific func­tional group to the backbone of the resin.

There are, however, some potential disadvantages of the solid-phase extrac­tion approach which are discussed in a later section.


The study using SRBs showed that SRB cells could achieve up to 68% removal of strontium from the medium (Ngwenya and Chirwa, 2011). Most of the solid phase Sr2+ species were easily desorbed from biomass using MgCl2 (Fig. 15.8). Metal species in the desorbed fraction gave an indication of the amount of Sr2+ that is bound on the biomass surface by relatively weak electrostatic interactions which are easily released by the ion-exchange process (Dahl et al., 2008). The elevated concentrations of Sr2+ in the desorbed fraction may be due to the release from the complexing agents on the microbial cell surface.

These experiments showed that the SRB cells were excellent cation exchangers. The significance of these findings was that Sr2+ and other diva­lent cationic fission products could be extracted from water using bacteria under natural biological conditions. Separation of bacteria from water is relatively easy and cost effective. The bacteria could then be treated with an eluant to reverse the process thereby recovering the metals. The bacteria could then be returned into the biosorption reactors.

Interactions at interfaces significant to the nuclear fuel cycle

Interactions that occur at phase boundaries (in particular, liquid-liquid interfaces in solvent extraction, solid-liquid interfaces in used fuel dissolu­tion, in the environment, in electrometallurgy and in the cleanup of wastes at the former nuclear weapons complex) by definition control the rates of materials transfer which in turn often governs the efficiency of the separa­tion process. In solid-liquid separation systems, molecular motions on the solid side of the interface are generally limited while the fluid phase motions are quite dynamic. In liquid-liquid interfaces, both sides of the interface are

in a state of dynamic movement. At the interface, solute and solvent mol­ecules reorder their relative structures to facilitate phase transfer. In solvent extraction, surface active molecules arrange themselves to assist in the transfer of polar species into the less polar regime represented by the organic phase. The organization of these molecules at the interface controls the rate of mass transfer. Such systems have been investigated, but the necessity of probing a dynamic interfacial zone of with dimensions of only a few molecular diameters hinders a complete understanding. Computational modeling studies have attempted to create a rational framework for advanc­ing understanding of interfacial interactions, but progress is slow.

Austentic stainless steels

Austenitic stainless steels are used extensively as the main material of con­struction for process vessels and pipework. Much effort has been expended in defining the operating limits of these steels. After many thousands of hours of laboratory testing and decades of plant experience, it has been possible to define realistic and predictable corrosion rates before and after welding, cold working and hot working. The main advantages of austenitic stainless steels are their inherently high resistance to corrosion in oxidizing media such as nitric acid and the relative ease with which they can be decon­taminated. They also have excellent impact resistance down to sub-zero temperatures, they are readily available and are easy to fabricate and weld.

There are two main types of austenitic stainless steel used throughout a reprocessing plant (Baldev, 2006). The first is a special grade developed by British Nuclear Fuels Limited and the steel companies which has been given the name NAG (nitric acid grade) 18/10L (Merrill, 1990). NAG 18/10L is basically a 304L type stainless steel with reduced carbon levels and close control on residual elements to give improved corrosion resistance. Careful controls are also placed on the manufacturing and testing processes for this material and it can only be manufactured by a limited number of companies which have to meet stringent technical and quality assurance requirements. It is used on all high integrity nitric acid applications. The 304L grade is used for less arduous duties such as cladding, water jackets and low or non­active pipework and vessels.

Uranium stripping

The final step in the partitioning cycle is the uranium stripping operation (Fig. 6.2, U strip contactor) to recover the U from the organic phase prior to solvent cleanup and recycle. At this point in the process, uranium is the predominant metal present in the organic phase. Because of the strong affinity of TBP for the uranyl ion, U back extraction is somewhat more difficult than for many of the other extracted metals. The uranium stripping operation is universal and relatively straightforward in PUREX processing: the metal laden organic is contacted with a low-acidity (0.01 M HNO3) aqueous phase, at elevated temperature (50 °C). The aqueous-to-organic phase ratio (O/A) is slightly higher than unity to help maintain a large concentration profile between phases.

Advanced reprocessing for fission product separation and extraction

E. D. COLLINS, G. D. DEL CUL and B. A. MOYER, Oak Ridge National Laboratory, USA

Abstract: The United States inventory of used nuclear fuels contains approximately 2 to 5 wt % fission products, depending on the extent of fuel burnup during irradiation, with the greater amounts produced in higher burnup fuels. For reprocessing of used nuclear fuels, fission products are more often divided into categories according to their chemical and radiological properties. Advanced reprocessing includes further separations processes to enable capture and disposal of the volatile fission product elements in improved solid waste forms, as well as additional separations processes being developed (1) to enable recovery and recycle of the remaining minor transuranium element actinides, neptunium, americium, and curium, and (2) to segregate the lanthanide fission products and the intermediate-lived heat-generating radionuclides, 137Cs/137mBa and 90Sr/90Y.

Key words: separation/extraction techniques, voloxidation, transition metal behavior, lanthanide recovery, cesium-strontium isolation.

7.2 Introduction

The United States inventory of used nuclear fuels contains approximately 2 to 5 wt % fission products, depending on the extent of fuel burnup during irradiation, with the greater amounts produced in the higher burnup fuels. The fission products are produced predominantly in two mass fractions, typically centered around a peak of mass 90-100 in the light fraction and 135-145 in the heavy fraction. For reprocessing of used nuclear fuels, the fission products are more often divided into categories according to their chemical and radiological properties, as illustrated in Table 8.1.1,2

The properties shown in Table 8.1 represent used fuels from light water reactors (predominantly thermal-neutron-driven fission of 235U). The fission product elements — tin, antimony, and tellurium — are included here but would become more important when produced in higher concen­tration from fast neutron-driven fissions. The radioisotopes of significance shown in Table 8.1 were confined to long-lived and intermediate-lived isotopes that are most significant in reprocessing used fuels that have been stored for five years or more after reactor discharge. The significant

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Transition metals (35%)















































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147 (2.62у)

154 (8.8y), 155 (4.71 у)

Подпись: 93 (1.5E6y) 93 (3.5E3y) 99 (2.1E5y) 106 (1.02y) 107 (6.5E7y) 110m (0.68y) 113 (9E15y) 115 (5E14y) 119m (0.8y) 125 (2.76y) 106 (with Ru-106)

125m (with Sb-125)

Re-enrichment and recycle


MOX fuel recycle


Compaction or grouting



Geologic disposal



8.1 Used fuel component separations in current plants.

radionuclide-decay isotope pairs that are in secular equilibrium are shown in Table 8.1 because, in effect, they double the radioactivity emitted by the parent radioisotope.

In established reprocessing plants (Fig. 8.1), the volatile fission products, xenon and krypton, are vented to the environment, while the iodine is trapped from the off-gas and then released to the sea as liquid waste. Carbon-14 is scrubbed from the off-gas and released to the sea as low-level liquid waste or converted to insoluble barium carbonate and encapsulated in a solid waste form, such as cement. After dissolution of the fuel compo­nents, the cladding hulls and hardware are removed by screening and put into a solid waste form (compacted metal or grout) for subsequent geologic disposal. The dissolved fuel solution still contains finely divided undissolved solids (UDS), composed of variable portions of the transition metal ele­ments. The UDS are removed by centrifugation and disposed in vitrified high-level waste, along with the soluble fission products that are separated from the uranium and plutonium products by means of solvent extraction processes.

Advanced reprocessing will include further separations processes to enable capture and disposal of the volatile fission product elements in improved solid waste forms. Also, additional separations processes are being developed (1) to enable recovery and recycle of the remaining minor transuranium element actinides, neptunium, americium, and curium, and (2) to segregate the lanthanide fission products and the intermediate-lived heat-generating radionuclides, primarily 137Cs/137mBa and 90Sr/90Y. Details of the advanced fission-product recovery processes are described below. The actinide recovery processes are described in other chapters.

General flowsheet for treatment of the UNEX process end products


Following on from the above discussions, Figure 9.9 presents a basic flow­sheet proposed for the management of HLW treatment end products by UNEX-extractant, as applied to sodium-bearing waste of INL. Hence, the

9.9 General basic diagram for management of end products of HLW treatment by UNEX process.

implementation of the UNEX-process together with subsequent treatment of its end products enables HLW volumes to be reduced by a factor of 200.

The effectiveness of the UNEX process was confirmed by a feasibility study conducted in 2000 by a US engineering company, with participation of RPA KRI. It was shown that, in the case of vitrification of Idaho acidic HLW without radionuclide recovery, 1 m3 HLW would produce 324 kg of glass. Further, with the vitrification of the UNEX-strip product, the amount of glass produced would decrease to 14 kg, i. e. by a factor of 23.