Undissolved solids (UDS) and secondary oxide precipitations

Following dissolution of fuel components in nitric acid solutions, a variable fraction of the transition elements remain as insoluble metal precipitates. The amounts of each element vary according to the prior fuel burnup and dry heat pretreatments, such as the voloxidation process. Data obtained from typical experimental measurements are illustrated in Table 8.3.

The chemical nature and behavior of the fission products from oxide fuels have been studied extensively and described thoroughly by Kleycamp and others.20 The insoluble transition metals were described as metallic inclu­sions or precipitates. Other studies have shown that the Mo, Tc, and portions of the Ru and Rh in the UDS can be dissolved in 4 M KOH solution and the remaining Ru, Rh, and all Pd can be dissolved in aqua regia. These special dissolutions may be necessary if recovery of the valuable noble metals is pursued in advanced reprocessing.

After clarification of the dissolved nitric acid fuel solution by centrifuga­tion to remove the UDS, secondary oxide precipitations of some fuel com­ponents occur during the subsequent feed adjustment, solvent extraction, and waste treatment processes. The amount of precipitation can be affected by the heat treatments applied during these operations, or simply upon storage of the solutions for extended periods. The finely divided oxide pre­cipitates have been shown to cause the development of interfacial cruds during solvent-extraction operations, but operational problems have appar­ently been avoided or minimized by the use of pulsed extraction columns, which apparently are not affected seriously by the presence of the finely divided oxide particulates.

Table 8.421 shows precipitated and crystallized solid fission products that have been observed during laboratory studies of evaporation and acid reduction processes used to treat the solvent-extraction high — level waste solutions. These studies have shown that the solids contain primarily Zr, Mo, Ba, Sr, and Sn. The solids are a combination of hydrous precipitates and crystallites that are insoluble in concentrated nitric acid. Although cesium would be expected to be highly soluble, the presence of 24% in the precipitate may have been by occlusion with other hydrated precipitates.

Soluble technetium: solvent extraction using tri-butyl phosphate (TBP) as the extractant

Established reprocessing plants utilize the PUREX process (Chapter 6) to separate and remove the non-volatile and nitric-acid-soluble fission prod­ucts from the uranium and plutonium products. Advanced reprocessing may use derivatives of the process for co-processing operations to produce a plutonium or plutonium-neptunium product containing part of the uranium, thereby reducing the proliferation risk by making access to the fissile material less attractive. In those TBP-based solvent-extraction processes, where the feed stream contains both zirconium and soluble technetium fission products, the technetium will form a complex with uranium and be co-extracted.

The technetium distribution coefficient is reduced at high solvent load­ings with uranium and with higher nitric acid concentrations in the aqueous phase, as shown in Fig. 8.2.22 These data indicate that separation of the technetium from the loaded solvent can be accomplished by scrubbing with >5 M nitric acid. Alternatively, the technetium can be allowed to remain with the uranium and be co-stripped into a low-acidity aqueous product, from which the technetium can be preferentially adsorbed onto an anion — exchange resin.

8.2 Technetium distribution coefficient.

After the electrorefining operation, the dissolution residue (which mainly consists of the stainless steel cladding hull, zirconium and NFP) remain in the anode basket. The composition of the residue strongly depends on the electrorefining conditions. As the chopped U-Pu-Zr or U-Zr metal fuel dissolves from outside to inside, it leaves behind a zirconium layer that acts as a dissolution barrier (Koyama, 2002) and it is necessary to charge a higher anode potential to dissolve the zirconium and ensure complete dis­solution of the actinides. Experiments with spent U-Zr fuels have shown that an average 99.7 wt% uranium dissolution is achieved with the dissolu­tion of approximately 88 wt% zirconium, while approximately 75% of NFP was retained in the cladding hulls (Li, 2005). About one-third of the dissolved zirconium was electrochemically deposited on the solid cathode along with uranium to be cathode processed, while the rest of the dissolved zirconium fell down to the bottom of the electrorefiner (Li, 2007). The electrochemical recovery of zirconium from the bottom cadmium layer was carried out, and a zirconium-rich uranium deposit was obtained. Alternatively, a dissolution mode to leave zirconium and NFP in the anode basket is an operation to reduce the effort of recovering these materials from the bottom of the electrorefiner, although part of the actinides remain undissolved. Whichever mode is selected, the dissolution residue in the anode basket is further treated in an ‘anode processing’ step to recover residual actinide, depending on the recovery ratio required.

Most of the batch experiments relevant to caesium selective extraction studies, involving calix[4]arenes-(mono/bis)crown-6, have been carried out in polar aromatic organic diluents, such as o-nitro-phenyl-alkylethers. Although they greatly enhance the extraction properties of calix[4]arenes- (mono/bis)crown-6 compared with alkanes conventionally used in the nuclear industry, these organic diluents are nonetheless incompatible with the implementation of counter-current process flowsheets: their viscosity and density are too high, and the kinetics of mass transfer too slow. This is why, in order to demonstrate the scientific feasibility of caesium partitioning from acidic nuclear waste streams, efforts at the CEA Cadarache (France) have focused on the substitution of o-nitro-phenyl-alkylethers by mixtures of hydrogenated tetrapropene (HTP), the diluent used in the French PUREX process, and modifiers allowing the use of calix[4]arenes — monocrown-6 without third-phase occurrence. Two reference systems pre­senting sufficiently high and selective caesium extraction from acidic nuclear waste have been optimized (Fig. 11.6):

1. The first is based on 1,3-(di-n-octyloxy)-2,4-calix[4]arene-crown-6 (0.065 mol. L-1) and tributyl phosphate (1.5 mol. L-1) as phase modifier.

System 1

Calix[4]arene-Crown-6 )o o£

1,3-[di-n-octyloxy]-2,4-calix[4]arene-cr-6 Modifier: TBP

TributylPhosphate

11.6 The two reference systems chosen at the CEA to selectively extract caesium from acidic nuclear waste streams.

2. The second is based on 1,3-(2,4-diethyl-heptylethoxy)-2,4-calix[4]arene- crown-6 (0.1 mol. L-1), more difficult to synthesize, and methyloctyl-1,2- dimethylbutanamide (1 mol. L-1) as phase modifier.

Process flowsheets have been developed for both systems and counter­current tests have been performed implementing laboratory-scale contac­tors (e. g., centrifuges, Taylor-Couette effect columns, and 15 mm diameter pulsed columns) and surrogate feeds. More than 99.9% of caesium was extracted and back-extracted in the centrifuge test, in agreement with flow­sheet modelling. Although satisfactory, the hydrodynamic behaviour of the second system (calixarene + monoamide) appeared more emulsifying in the pulsed and Taylor columns than the first system.

In the light of these promising results, two hot tests were performed in the hot cells of the Atalante facility on a genuine DIAMEX raffinate[11] (see Fig. 11.7 for the flowsheet implemented). By adding oxalic acid to the genuine feed, the extraction of molybdenum and zirconium was prevented and very high caesium recovery yields (>99.9%) were observed, thus con­firming the ability of calix[4]arenes-monocrown-6 based solvents to parti­tion caesium efficiently and selectively from acidic nuclear waste streams (Madic et al., 2002).

11.7 Flowsheet for caesium partitioning from genuine DIAMEX raffinate implemented in the Atalante facility of the CEA Marcoule (Madic et al., 2002).

The radiolytic degradation of 1,3-(di-n-octyloxy)-2,4-calix[4]arene — crown-6 was studied in the presence of nitric acid. High-performance liquid chromatography, directly coupled with ESI-MS, allowed more than 50 dis­tinct degradation products to be observed, and about 30 of them to be identified (as products issued from radical cleavage or addition, oxida­tion, and aromatic substitution) in aliphatic and aromatic diluents (Lamouroux et al., 2004). Despite the severe degradation conditions tested ([HNO3] = 3 mol. L-1 for the acidic hydrolysis and a gamma delivered dose of 106 Gy for radiolysis), 1,3-(di-n-octyloxy)-2,4-calix[4]arene-crown-6 appeared remarkably stable as illustrated by the limited losses of compound observed: 33.5%.

High level waste (HLW) arises from the use of uranium fuel in a nuclear reactor and nuclear weapons processing. It contains the fission products and transuranic elements generated in the reactor core. It is highly radioactive and “hot”. In the parlance of the nuclear industry, it is regarded as the “ash” from “burning” uranium. HLW accounts for over 95% of the total radioac­tivity produced in the process of nuclear electricity generation. However, the proportion of LLW is higher in the new generation HTGRs due to emphasis on safety. This reversal of waste volume ratios has resulted in a new problem of waste volume that puts emphasis on environmental com­patibility in deference to safety.

This section focuses on the containment of plant for processing radioactive material and provides necessary background for subsequent discussion on equipment. Plants processing radioactive material require significant bio­logical shielding to protect workers and the environment from harmful doses of radiation. Phillips (2007) has provided a comprehensive review of the use of the two main containment concepts: canyons and passive secure cells (PSCs). Radioactive material processing plants using PSCs and canyons as containment have been successfully built and operated in Europe and the USA.

Canyons are used with overhead and polar cranes within the enclosure, typically shielded windows in the walls, remote manipulators and similar equipment installed. These allow operations, maintenance work and the replacement of equipment to be carried out remotely. Canyons can be constructed with no windows and rely entirely on internal cameras, or they can have shielded windows for internal viewing on at least one side and many have windows on both sides. Lighting units in the cell walls allow bulbs to be replaced from outside the canyon. Canyons are typically used to contain predominately mechanical handling equipment that requires maintenance and physical control, although in the USA they are used also for chemical plant and equipment which is connected by removable pipe­work “jumpers” to allow remote removal and replacement of equipment items. A good example of a mechanical equipment canyon is the one con­taining Advanced Gas-cooled Reactor (AGR) UNF dismantling equipment at the Sellafield nuclear facility in the UK (Fig. 3.1). This canyon allows the

remote removal of individual fuel pins from the graphite containment prior to reprocessing.

PSCs are used, on the other hand, when the cell is to be employed pre­dominately for process plant with mainly pipes, tanks, pumps, separation and filtration equipment. PSCs are located separately and linked by shielded pipe enclosures, they do not normally surround canyons. There are three types of PSC. Type 1 PSCs contain plant items with no maintainable moving parts, and the pipework, vessels and other process equipment are all welded and radiographed to nuclear standards. The use of Type 1 PSCs has led to the development of a range of liquid transfer and liquid redirection equip­ment plus measurement devices that have no moving parts and thus can be permanently installed in the cell for the facility lifetime (Phillips 2006). Type 2 PSCs contain plant items with slowly rotating or intermittently moveable parts, but all maintainable items such as motors and gearboxes are located outside the cell, with sealed through-cell-wall drives. Type 3 PSCs are a newer design, dating from the 1980s, in which all in-cell equipment items with moving parts that require maintenance are designed as removable modules.

With all PSCs, there is no expectation for cell entry, or manipulation of the equipment within it, during the life of the facility. Nevertheless access doors are usually provided and, because there are no pipework jumpers to spill small amounts of radioactive liquid when they are removed, the inter­nal surfaces of PSCs remain uncontaminated. Therefore, the radiation levels in PSCs are usually very low after the process pipework is flushed.

The Enhanced Actinide Removal Plant (EARP) at Sellafield, UK pro­vides a good example of the use of type 3 PSCs (Fig. 3.2). The housings for the equipment are permanently welded into the in-cell pipework, and the equipment modules can be withdrawn from these housings, through

Shielded roof of type 3 PSC

Access hatch

All-welded fixed in-cell housing for
removable crossflow filtration assembly

3.3 Type 3 PSC with flask in place on cell top. Source: Nuclear Decommissioning Authority ("NDA"), copyright: Nuclear Decommissioning Authority ("NDA").

removable hatches in the PSC roof, into shielded steel “flasks” (Fig. 3.3). The flasks are moved to a maintenance cell where the modules are repaired or prepared for disposal. The process is reversed to re-install the modules back into service in the PSC. Equipment modules can contain a wide range of process equipment including liquid pumps, valves, filter units, centrifugal contactors, and instruments. It is usual practice to keep a range of spare modules so that in the event of failure a replacement can be rapidly installed and plant downtime kept to a minimum.

Although many safeguards discussions focus on reprocessing facilities, the scope of facilities under IAEA safeguards is much broader. Uranium enrichment plants are also a major focus. Other types of facilities subject to some degree of IAEA safeguards include conversion plants, reconver­sion plants, fuel fabrication plants, nuclear power reactors, interim fuel storage facilities, and waste repositories. Based on 2002 data, there were 239 reactors, 80 fuel storage facilities, 41 fuel fabrication facilities, and 10 enrichment plants under IAEA safeguards. In contrast, there are only 6 reprocessing plants under some level of IAEA safeguards (IAEA 2003).

With the anticipated growth of nuclear power, there is a potential that reprocessing may expand. This has justifiably resulted in increased concerns about sustainability and environmental issues associated with spent fuel management. Nuclear reactors are unloading more than 10 500 MTHM of spent fuel per year, and there are estimated to be 340 000 MTHM in storage in 2010. With the capacity for commercial reprocessing at only 5 550 MTHM per year (IAEA 2008), more plants may need to be deployed to manage the existing inventories.

Reprocessing plants do have challenges with respect to nuclear materials safeguards. Verification requirements for reprocessing facilities impose what are probably the most complex design aspects of nuclear facilities. Verification activities for a reprocessing plant require more than 750 man- years compared to 6 to 12 man-years for a light water reactor (IAEA 2005b). Many of the challenges are associated with the nature and scale of large chemical plants that have potential high-throughputs of nuclear mate­rials. As noted earlier, a goal of safeguards is the detection of a significant quantity of nuclear material in a timely manner. This goal is independent of the amount of material processed through the facility.

A typical commercial reprocessing facility may process on the order of 800 MTHM of light water reactor fuel per year. This spent fuel contains approxi­mately one percent plutonium. Therefore, the annual plutonium throughout is 8 000 kg. This equates to more than 660 kg of plutonium processed on a monthly basis, or 83 significant quantities. At these high throughput rates, detection of the loss of a significant quantity of plutonium (i. e., 8 kg) can be extremely challenging based solely on the uncertainties associated with veri­fication measurements. Continual improvements are being made in repro­cessing technologies, facility designs, and measurement techniques in order to further enhance the ability to safeguard reprocessing facilities.

In addition to the safeguard issues associated with large quantities of special nuclear material, reprocessing facilities also have a couple of other unique challenges. Reprocessing facilities subject to IAEA safeguards pres­ently deploy aqueous separations technologies. For example, variations on the PUREX technology are routinely used. From the PUREX process, rela­tively pure separated plutonium is recovered. In the process portion of the plant, the fissile material is sometimes hard to assay accurately because the material is in the form of a solution as it is being processed.

IAEA safeguarded facilities have ranged from the first commercial reprocessing plant, EUROCHEMIC, built in Belgium during the 1960s to the Rokkasho Reprocessing Plant expected to come on line in Japan in 2009. As reprocessing technologies have advanced with respect to efficien­cies and reduction of environmental impacts, steps have also been taken to further improve nuclear materials safeguards. Advancements in the separations technologies have also focused on non-proliferation aspects of reprocessing.

Safeguards remain a key factor to reduce the risk of proliferation from reprocessing facilities, as well as other nuclear facilities. With respect to reprocessing, technology development has also focused on areas outside of safeguards to potentially reduce the risk of proliferation. Proliferation barriers are divided into two groups, intrinsic and extrinsic. Intrinsic barriers are those that are inherently part of the system, while extrinsic barriers are more institutional or administrative in nature.

Much of the technology development associated with reprocessing has focused recently on increasing intrinsic barriers. Examples include develop­ment of new process flowsheets that do not result in the separation and recovery of purified plutonium. This work includes development of flow­sheets that result in the recovery of a plutonium-uranium product instead of a purified plutonium product. A flowsheet of this type is deployed in Rokkasho. Another type of intrinsic feature is complete group recovery of all actinides (i. e., uranium, plutonium, neptunium, americium, and curium). Recent developments in both aqueous and dry separation technologies are focused on these efforts. Other intrinsic features include co-location of facilities to minimize transportation of special nuclear material. Co-location of fuel fabrication facilities has long been integrated with deployment of some dry process technologies, but it is also now being considered for aqueous systems (Vinoche et al. 2005). Development of fuel types that are harder to process would also be considered an intrinsic feature. Extrinsic measures include activities such as the application of robust nuclear safe­guards. Both extrinsic and intrinsic barriers are critical measures for non-proliferation.

As noted, a key component of safeguards is nuclear material accountancy, which is the system for tracking the quantity and movement of nuclear material in a facility. Because of the challenges associated with reprocessing plants, the specific physical areas over which measurements are performed are critical. These areas are called the material balance areas (MBAs) and they are incorporated into the physical design of the facility. An MBA is defined as an area inside or outside a facility over which the quantity of nuclear material transferred into or out of the facility can be determined. Additionally, the inventory within the MBA must be able to be determined (IAEA 2001). It is the movements of nuclear materials across the boundary surrounding the MBA that are reported to oversight groups like the IAEA.

An example of MBAs for a reprocessing plant consists of three distinct MBAs (OTA 1995, Thomas and Longmire 2002). The first MBA is associ­ated with the receipt and storage of spent nuclear fuel to be processed. Fuel chopping and dissolution would be included in this MBA. The second MBA is associated with the reprocessing operations. The final MBA is associated with storage for the final products. Within MBAs, there are key measure­ment points or the locations where nuclear materials are in a form that is amenable to quantitative measurement. In general, these points need to occur near the inputs and outputs to the MBAs, but they are not limited to those locations.

Measurements to establish the inventory within a given MBA or process unit operation can involve a number of different techniques and physical data are needed. Depending on the form of the material measured, critical measurements may include mass, volume, flow rate, density, temperature, and chemical composition.

Analytical samples for safeguards measurements have some unique requirements. First, the sampled material must be homogeneous and, most importantly, the sample must be representative of the bulk material being considered. Next, the sample integrity must be maintained by chain-of — custody tracking to ensure that potential diversion is not masked by sample tampering or sample switching. Samples for safeguards purposes may be taken by the facility operator, but samples are also taken independently by the IAEA for verification. Those samples must be independently analyzed by the IAEA and may require shipment to off-site IAEA laboratories. For a large reprocessing facility like Rokkasho, the IAEA established an on-site laboratory (Duhamel et al. 2005).

Concentration measurement techniques are generally divided into destructive and non-destructive analysis. Destructive analysis employs tech­niques that result in the destruction of the sample. Non-destructive analysis employs techniques that do not produce a significant physical or chemical change in the sample. Destructive analyses are used to quantify inventories fairly precisely. Non-destructive analyses are used predominately for veri­fication (DeMuth et al. 2007).

Samples in a reprocessing plant consist of in-process solutions, final prod­ucts, and waste streams. Examples of waste streams include cladding hulls after dissolution of spent fuel and undissolved solids.

The analytical technique employed for a sample depends on the sample form, characteristics to be measured, and the required uncertainty. For safeguards purposes, there are requirements to know both elemental com­position (uranium and plutonium) and isotopic composition (uranium-235, uranium-238, plutonium-239, etc.).

Mass spectrometry is one key class of analytical techniques for these types of measurements. It is a destructive analysis. Mass spectrometry works by first ionizing the elements or compounds and then measuring their mass to charge ratios. Thermal ionization mass spectrometry (TIMS) is a critical technique for measurement of plutonium and uranium isotopics. The instrument is capable of high precision and accuracy. It can be used for many materials containing plutonium and uranium, and is the primary analytical method used for solutions of dissolved spent fuel. Isotopic dilu­tion mass spectrometry (IDMS) is also employed as a technique for increased precision and accuracy. This technique involves dissolution of a known amount of a given isotope from the elements of concern in the sample. When analyses of spiked and unspiked samples are compared, sig­nificant improvement can be achieved in the precision and accuracy. Other destructive analysis techniques include X-ray fluorescence, potentiometric titration, controlled potential coulometry, ignition gravimetry, K-edge absorption densitometry, alpha spectrometry, and gamma ray spectrometry (OTA 1995, IAEA 2003).

Gamma ray spectrometry can also be used as a non-destructive technique. This technique takes advantage of the fact that most nuclear materials emit gamma rays of specific energies. Neutron detection can also be employed. Neutrons are emitted from spontaneous fission, induced fission, and reac­tions from the emission of alpha particles (alpha-neutron reactions).

Technology development and deployment activities are continuing to occur to further improve measurement techniques to enhance further nuclear materials safeguards. For large reprocessing plants this work is beneficial to meeting detection goals in a timely manner. Much of this work has focused on a concept known as near-real-time accountancy. Inventories at the facilities are performed monthly and annually. The annual inventories are extensive and typically require the facility to stop process operations. The monthly inventories may be less precise because the facility is not necessarily shutdown, implying that the measurements are less precise for in process inventories. To improve detection of diversion, unattended instru­ments may be positioned in the facility at key points to provide measure­ments of nuclear materials. These instruments typically deploy non-destructive techniques like neutron coincidence counting (OTA 1995). Rokkasho is making extensive use of unattended sampling and measurement devices. Approximately three quarters of the data collection is performed unat­tended (IAEA 2005b). Data from process monitoring can also be used to support safeguards to verify that the facility is being used in the manner for which it was designed.

As mentioned earlier, a key aspect of safeguards is containment and surveillance, which complements nuclear material accountancy. This tech­nology is also used extensively for safeguards by IAEA. Containment and surveillance technologies are divided into two categories: optical surveil­lance and sealing systems. Optical surveillance works very effectively for storage areas to observe movement and transfers of materials. Fuel storage pools are good candidates for deployment of cameras. Seals in general are applied to containers to help indicate if nuclear material has been removed or added to a container.

For the case of LWR recycling, a number of incentives exist to separate and recycle the actinides, and to separate and manage fission products. The main incentive for LWR spent fuel separations under the Global Nuclear Energy Partnership (GNEP) program was to extend the US geologic repository capacity by: 1) recycling of transuranic elements for ultimate destruction and recovery of energy content, and 2) reducing the volume of waste, long­term radiotoxicity and heat load in a geologic repository (GNEP, 2007). To achieve this goal, a number of separations product and waste streams were identified. Table 7.1 lists these product and waste streams along with the separation incentive.

Table 7.1 GNEP LWR SNF recycle product/waste streams

Product/waste Specification stream

(1) In all cases, iodine is removed as an off-gas from the dissolution process.

(2) Processes are designed for the generation of no liquid high-level wastes.

U: Uranium (contributor to dose rate, and the mass and volume of high-level waste). Tc: Technetium (long-lived fission product, minor contributor to long-term dose). Cs/Sr: Cesium and strontium (primary short-term heat generators, affect waste form loading and repository drift loading).

TRU: Transuranic elements: Pu — plutonium, Np — neptunium, Am — americium, Cm — curium (primary long-term dose rate contributors).

Ln: Lanthanide fission products.

FP: Fission products other than cesium, strontium, technetium, iodine, and the lanthanides.

Table 7.2 shows the processing options for LWR SNF based on the tar­geted distribution of the various product streams given in Table 7.1. Each of the process outputs will require further processing to a solid form. This form will have an associated set of requirements that must be met for acceptability. In the case of the product, such as a mixed plutonium-uranium stream, the material must be suitable for use as a reactor fuel. Though some further chemical modification may be possible as part of the solidification processing, it is unlikely that significant chemical rework will be a compo­nent of fuel fabrication. Waste forms will likely have looser tolerances than recycled fuel, but the solidified waste precursor must eventually yield a waste form that is acceptable to the ultimate disposal site.

Composition of the universal extraction mixture may be varied depending on the characteristics of HLW being treated, in the range of 0.06-0.12 M CCD, 0.005-0.03 M CMPO, 0.075-0.25 M PEG-400 in diluent FS-13. Such composition provides the balance needed for effective recovery of Cs, Sr, An and REE with acceptable physico-chemical and hydrodynamic proper­ties of the extractant.

Figure 9.3 shows typical dependence of the distribution coefficients for metals on the acidity of the aqueous phase.

In addition to radionuclides, the UNEX extractant recovers some stable elements: analogs of strontium, barium and lead are completely recovered; potassium and molybdenum are partly extracted; co-extraction

of zirconium and iron considerably decreases in the presence of fluoride — ions; other metals (HLW components) are practically not extracted at all.

The high extraction ability of the UNEX extractant makes the subse­quent stripping of radionuclides more difficult. Cesium and strontium can be stripped by HNO3 containing additions such as amines, amides, alcohols. These additions facilitate the transfer of a proton into the organic phase and the displacement of Cs and Sr cations from it.

To strip REE and An, it is necessary to use solutions of rather efficient complexones like oxyetylidenediphosphonic acid. Uranium is stripped by carbonate solutions only. In the presence of complexones, ammonium car­bonate also strips Sr, REE and An. Essential differences in the behavior of radionuclides at the stripping stage (stripping from the UNEX-extractant) allow partitioning of radionuclides to be made. If required, sufficiently pure fractions of Cs, Sr, U, REE and An can be obtained.

To produce concentrates of all radionuclides in the HLW treatment process, a common strip agent should be applied. Since uranium is stripped by carbonate solutions, the possibility of using complexone solutions with carbonate was investigated. It was found that a solution containing 1.0 M guanidine carbonate and 0.05 M DTPA effectively stripped Cs, Sr, An and REE.

In the current INL program to treat irradiated EBR-II fuels, electrorefiner salt is directly converted to glass-bonded sodalite without the condensation of LFP by a zeolite column, nor with the separation of actinides by counter­current extraction. According to INL (Priebe, 2008), the major steps for forming ceramic waste from the electrorefiner salt are:

1) to blend zeolite and salt at approximately 500 °C to form salt-loaded zeolite;

2) to mix the salt-loaded zeolite with glass; and

3) to heat the mixture to form glass-bonded zeolite.

Frozen salt blocks are crushed and ground to a particle size of 100-200 pm then mixed with dried zeolite-4A powder (with a particle size of 45-250 pm) in the rotating V-mixer shown in Fig. 10.20 at 525 °C for about 18 h. During the mixing, the electrorefiner salt is absorbed into the zeolite structure to form salt-loaded zeolite. The V-mixer is then allowed to cool to ambient temperature, and samples are taken to confirm that the content of free chloride is in an acceptable range. In the next step, a glass frit with a particle size of 45-250 pm is added to the salt-loaded zeolite and mixed at the ambient temperature to ensure homogeneity. The fraction of glass in the

mixture is about 25%. Finally up to 400 kg of the salt-loaded zeolite and glass mixture is loaded into a production furnace with internal dimensions of 68 cm diameter and 3 m height, as shown in Fig. 10.21. The furnace tem­perature is increased to around 925 °C which is maintained for approxi­mately 72 h to allow complete consolidation. The furnace is then cooled to ambient temperature, to allow the monolithic, glass-like CWF (as shown in Fig. 10.21) to be removed. The furnace is currently being subjected to quali-

10.22 Metal waste furnace (left) and processed metal waste (right).

fication tests out of the cell in preparation for hot-cell operation (Goff, 2009).

Systems specifically dedicated to MA transmutation have been proposed and largely studied in some countries [6]: they comprise a subcritical reactor coupled to an external neutron source supplied by a proton accelerator. Such dedicated ADS (accelerator driven system), operating in a sub critical mode, offers:

• a safety guard against accidental reactivity increase which allows con­sidering large MA loads whose reactivity feedback (Doppler effect) and proportion of delayed neutrons would make them prohibitive with criti­cal cores,

• an acceptance of a greater variation in the fuel’s isotopic composition, resulting from transmutation during the cycle, as the system is no longer constrained by self-criticality during the entire time the fuel is in the reactor.

The complexity of these systems means that they cannot be considered as electricity-generating reactors. Their development involves the design of highly technical elements, such as a reliable power accelerator which has to supply an intense high energy proton beam for long periods of time without beam interruption, a spallation target which has to produce high energy neutrons under the effect of the accelerator’s proton using the lead bismuth eutectic (LBE) and a window separating the accelerator void from the LBE, and a reactor core operating in sub critical and fast spectrum mode. For the first time ever in 2002-2006, the main components were successfully assem­bled for studies at zero power in the French CEA Cadarache Masurca reactor. The results of this programme have allowed development of vali­dated sub-criticality level measurement techniques. Nevertheless, several technological constraints still remain and must be overcome before judging the viability of a powerful ADS; studies are ongoing on these subjects as part of the Belgian MYRRHA Demo ADS project and the European EUROTRANS project dealing with ADS designs (both experimental and industrial), neutronics, fuels, LBE technology and specific nuclear data.