The actinide metals obtained in the cathode processing step are treated in an ‘injection casting’ process, with additional uranium and zirconium to adjust the composition of the recycled fuels. The uranium-zirconium alloy is vacuum melted at approximately 1560 °C and the plutonium — uranium-MA-zirconium alloy at approximately 1480 °C (Nakamura, 2009), before they are then injection-casted into a bundle of quartz moulds by ambient pressure to form metal fuel rods consisting of core and blanket fuels, respectively. Recycled metal fuel elements are fabri­cated by inserting these rods into steel cladding with sodium metal as thermal bonding.

The thermodynamic and crystallographic studies carried out on crown-ether based calix[4]arenes, in which one polyethylene chain (-CH2-CH2-O)n bridges two oxygen atoms of two opposite phenol units at the ‘narrow-rim’ (as shown in Fig. 11.4), have led to the following observations:

• p-tert-Butylcalix[4]arenes-dimethoxy-monocrown-n (n = 5 or 6) better complex large alkali cations, such as potassium and rubidium, because they can adopt a flattened partial cone conformation (Ghidini et al., 1990), which is impossible for larger alkoxy functions.

• As soon as the two remaining functions grafted onto the ‘narrow — rim’ of a calix[4]arene contain more than two carbon atoms, any of the four conformations encountered (Fig. 11.3) can be blocked in solution.

• Unlike a podant-based calix[4]arene, the selectivity of which favours sodium complexation only if it adopts the cone conformation, the pres­ence of a polyether bridge on the ‘narrow rim’ of a calix[4]arene, that has been blocked in the cone conformation, enhances its complexing and extracting properties toward alkali cations and offers additional control of the selectivity through the adjustment of the size of its coor­dinating cavity to the targeted metallic cation radius. As a result, a bridge presenting five oxygen atoms appears suitable for potassium complexation, whereas a bridge containing six O-atoms better fits the caesium cation radius (Ungaro and Pochini, 1991).

• The presence of a polyether bridge on the ‘narrow rim’ of a calix[4] arene, blocked in the 1,3-alternate conformation, strongly increases both its extraction efficiency toward caesium from acidic feeds and its selec­tivity versus other alkali cations, which is assumed to be due to a favour­able enthalpy contribution (Ungaro et al., 1994, Casnati et al., 1995, 1996, 2001, Sachleben et al., 1999, Talanov et al., 2000, 2002).

In reality, the benefit of the 1,3-alternate conformation for caesium selective extraction was first observed with the symmetrical doubly crowned calix[4]arenes (Fig. 11.5), synthesized by Vicens’ team who looked for easier manufactured bridged calix[4]arenes, obtainable in single-step syntheses avoiding the alkyl substitution of the two remaining phenol units of the calix[4]arenes-monocrown-n (Asfari et al., 1992, 1995). Like crown ethers and calix[4]arenes-monocrown-n, the calix[4]arenes — biscrown-n perfectly illustrate the benefit of matching the size of the coordinating cavity of the ligand with the ionic radius of the target cation. For instance, calix[4]arenes-biscrown-n, bearing five (n = 5) or seven (n = 7) oxygen atoms in their ether-crowns, show neither high extraction yields toward caesium, nor higher selectivity toward Cs+ (over other alkali cations) than di-(tert-butyl-benzo)-21-crown-7. As the caesium aqua complex possesses six water molecules, the six O-atoms of the ether-bridges of calix[4]arenes-biscrown-6 are consequently well pre-organized to displace the six water molecules of caesium inner co­ordination sphere.

Furthermore, outstanding Cs+/Na+ selectivity (SFCs/Na, exceeding 30 000) was obtained with calix[4]arenes-crown-6, the polyether bridges of which contain aryl rings such as benzyl or naphthyl (Fig. 11.5, Hill et al., 1994, Dozol et al., 1999). The selectivity of these ligands toward caesium is so high that they are better Cs+ sensors than any other functionalized calix[4]arenes

11.5 Examples of doubly bridged calix[4]arenes with functionalized crown ethers.

(Perez-Jimenez et al., 1998). Molecular Dynamics calculations as well as X-ray crystallographic data suggest that, provided no steric hindrance is introduced in the poly ether bridge (s), hydrophobic interactions exist between the extracted alkali cations and the n electrons of the aryl rings, hence favouring the binding of the less hydrated caesium cation as com­pared to harder alkali cations, such as sodium (Wipff and Lauterbach, 1995, Lauterbach and Wipff, 1996, Thuery et al., 1996, Lamare et al., 1997, 1998, 1999, 2001, Asfari et al., 1999, Jankowski et al., 2003).

As expected, although unusual in metallic complexation, the symmetrical arrangement of calix[4]arenes-biscrown-6 with two complexing cavities, is well adapted to the formation of both 1 : 1 (ligand : metal) and 1 : 2 com­plexes, as indicated by NMR, electro-spray ionization mass spectrometry (ESI-MS), and X-ray crystallographic studies (Arnaud-Neu et al., 1996, Allain et al., 2000).

The design of calix[4]arenes-crown-6, presenting one (or two) polyether chain(s) bridging two opposite phenol units of calix[4]arenes, blocked in the 1,3-alternate conformation, has therefore allowed both concepts (ligand pre-organization and host-guest complementarity through size fitting between substrate and receptor) to be tested in the search for caesium selective lipophilic extractants.

Intermediate level waste (ILW) contains higher amounts of radioactivity and some require shielding. It typically comprises resins, chemical sludges and metal fuel cladding, as well as contaminated materials from reactor decommissioning. It may be solidified in concrete or bitumen for disposal. Generally, short-lived waste (mainly from reactors) is buried in a shallow repository, whereas long-lived waste, for example waste from fuel repro­cessing, could be buried deep underground.

Almost every facet of chemical engineering is impacted when processing material with radioactive and fissile properties. The radioactive properties impact plant operations and maintenance where minimizing the radiation dose to workers and the environment has led to the development of process equipment and concepts unique to nuclear chemical engineering. Criticality control is required to handle material with fissile properties, invoking equip­ment construction material and geometry considerations.

The purpose of this chapter is to provide the reader with a broad over­view of nuclear chemical engineering for aqueous radioactive material separations. It is not the intention here to provide an exhaustive description of every unit operation or facet of chemical engineering employed in pro­cessing radioactive material. Rather, the more common chemical engineer­ing operations and equipment are described to exemplify the concepts and features unique to nuclear chemical engineering and on how chemical engineering is impacted by radioactive material properties. Wherever pos­sible, reference is made to operating plant for examples.

Nuclear chemical engineering for advanced aqueous radioactive material separations is a broad subject area. Different containment concepts, passive or secure cells and canyons, have evolved in Europe and the USA, respec­tively, and are discussed first, since they profoundly impact equipment and processing concepts. The special features associated with separations equip­ment are then described with reference to solid-liquid separation, ion exchange and solvent extraction unit operations. Equipment materials con­siderations not only include corrosion resistance but also the special mea­sures required to control criticality and radiation damage. Finally, future trends in the aqueous processing of radioactive material are discussed.

Requirements for safeguarding nuclear materials exist both internationally and domestically. Nations involved with nuclear activities, either at a gov­ernmental or industrial level, must establish a regulatory framework to oversee aspects of the work including worker safety, environmental protec­tion, and nuclear safeguards. In the United States, domestic requirements are established through the Nuclear Regulatory Commission and the Department of Energy. Internationally, the IAEA is responsible for nuclear materials safeguards. According to the IAEA, “The IAEA is the world’s nuclear inspectorate, with more than four decades of verification experi­ence. Inspectors work to verify that safeguarded nuclear material and activi­ties are not used for military purposes” (IAEA 2005a).

In general, the goals of international nuclear safeguards are to detect proliferation and diversion of nuclear material from the civilian nuclear fuel cycle and to provide notification of potential diversion to the international community in a timely fashion so that the consequences of the diversion can be reduced. Nuclear material is specifically defined as either special fissionable material or source material. Special fissionable material is plu­tonium-239 or uranium enriched in either uranium-235 or uranium-233. Source material is material that can be used to produce special fissionable material, which includes natural uranium, depleted uranium, and thorium.

A goal of the IAEA is to detect diversion of significant quantities of nuclear material. In general terms, a significant quantity of material is con­sidered a “threshold” amount needed to make a weapon (IAEA 1975). For example, spent light water reactor fuel contains both uranium and pluto­nium. For the reprocessing of spent nuclear fuel, plutonium is the main special fissionable material of concern. A significant quantity of plutonium is considered 8 kg total plutonium (Thomas and Longmire 2002). If dealing with highly-enriched uranium, a significant quantity is considered 25 kg of contained uranium-235 (IAEA 1981).

Safeguards goals focus on both detection and timeliness. Timeliness goals are applied to the input spent fuel and separated products, and specific timeliness goals are established based on the attractiveness level of the material of concern. For example, a timeliness goal may be based on the estimated minimum time required to convert a specific material into a form that can be used in a weapon. Conversion times can range from weeks to months and in some cases may extend to over a year (OTA 1995). A general goal is to be able to detect the diversion or loss of one significant quantity of plutonium (i. e., 8 kg) within one month. Goals for the detection of mate­rial in spent fuel are lower because the attractiveness level of that material is much lower. Therefore, the timeliness goal for detection of a significant quantity of material from spent fuel is three months.

Low-enriched uranium is also recovered from the reprocessing of spent fuel. Low-enriched uranium contains less than 20% of the uranium-235 isotope. The timeliness goal for low-enriched uranium recovered from reprocessing is that no more than 75 kg of uranium-235 is unaccounted for within a one-year period (Thomas and Longmire 2002).

The two major components of nuclear safeguards are nuclear material accountancy and nuclear material containment and surveillance. Nuclear material accountancy is the system used to determine the amount of nuclear material present in a facility and to track the changes in the quantity of material as process activities are performed in the facility. It is basically an accounting system for keeping track of amount, type, form, and location of nuclear material in a facility. Measurement systems to establish these quantities are a major part of nuclear material accountancy.

Containment and surveillance complement nuclear material accountancy. Containment and surveillance are the methods used to ensure that move­ments of nuclear material are tracked. Containment specifically deals with ensuring that once nuclear material is placed into a physical container (whether a room, vault, enclosure, or container) that its location is known or tracked. Devices to support containment and surveillance include physi­cal barriers, cameras, detectors, high radiation environments, and tamper indicating devices.

Nuclear material accountancy and nuclear material containment and surveillance are the backbone of nuclear materials safeguards for both the national and international communities. Operators of nuclear facilities, like a spent fuel reprocessing facility, establish systems for accountancy, contain­ment, and surveillance. The data generated from these systems are provided to State agencies and the IAEA.

One function of the IAEA with respect to safeguards is to examine and evaluate the data provided by nations hosting nuclear facilities. The IAEA also has a significant role in data verification. They independently collect data and information from the facilities as well as perform inspections. IAEA inspections often include independent data collection and analysis using equipment that is independently designed, operated, and maintained by the IAEA. Inspections also provide verification of facility and process designs by ensuring that the facility is performing only the declared functions.

Since the nuclear material accountancy system relies on various measure­ments, the uncertainties associated with quantitative analytical measure­ments and mass, volume, and density measurements impact the interpretation of the data. The nuclear accountancy system should be able to determine the quantity of material at the facility that is not accounted for, which is termed Material Unaccounted For (MUF). The MUF is sometimes referred to as the Inventory Difference (ID). In an ideal situation where there is perfect accountancy, when the nuclear accountancy books for a facility are closed, the MUF will assume a value of zero. However, because measure­ment uncertainties always exist, the value of MUF must be compared against the value of the measurement uncertainty. The accountancy system needs to be able to distinguish, to a reasonable degree, if the MUF values are “not significant”, in which case they are indicative of the expected measurement uncertainties, or are “significant”, in which case they are indicative of possible diversion. Obviously, high uncertainties would make it impossible to detect diversion, regardless of the MUF value.

Facilities will have detection goals to determine if the MUF values are significant. In establishing these goals, detection systems can result in two types of error. The first type of error is a false positive, where an alarm is signaled, but in fact no loss of material has occurred. False positive alarms obviously have significant operational issues in a production facility, not the least of which is loss of credibility for the system’s integrity. The second type of error is a false negative, where an alarm is not signaled, but in fact a loss of material has occurred. This situation is much more serious because it results in a case where material loss or diversion is not detected. Typical IAEA thresholds are established at a 90% detection probability for detecting diversion of a significant quantity of nuclear material and 5% as the maximum accepted value for a false alarm rate (OTA 1995).

In the event that the IAEA receives evidence that diversion of nuclear material may have occurred, the protocol is for the IAEA Director General to deliver a report to the IAEA Board of Governors, which must then make an evaluation. If they are unable to verify that no diversion has occurred, the Board of Governors may take actions, which can include reporting to the UN General Assembly or the UN Security Council. The Security Council, the only UN body with executive powers, has authority to take action including the implementation of international sanctions.

There are presently 146 IAEA member states and 237 safeguards agree­ments in force in 163 countries. In 2007, 2122 safeguards inspections were performed. The IAEA’s detection and verification system is not perfect. To note from IAEA documentation, “safeguards can neither predict diversions ahead of time, nor physically predict them, nor be guaranteed to detect them 100 percent of the time, and they should not be expected to do so” (Thomas and Longmire 2002). Still the IAEA, working with individual nations, serves an extremely important function with respect to safeguards.

Transparency is also a critical component of safeguards and non­proliferation. Transparency consists of actions taken by a facility or nation to enhance the openness of activities to ensure other nations or organiza­tions like the IAEA that they are not performing clandestine operations. Transparency would include allowing inspectors greater access to facilities, more openness in the timing of inspections, and wider environmental sam­pling to potentially detect undeclared activities.

Once the product and waste streams, and feedstocks are defined, the next step is to determine the separations modules needed to accomplish this goal. Options are not limited to solvent extraction; hybrid processes may include ion exchange, precipitation, electrochemistry and others. The overall process is composed of a sequence of separation steps or modules linked to generate a desired set of outputs, whether products or intermediates. Although minimizing the number of steps is always desirable, process reli­ability must take precedence. Caution is needed in assessing the compatibil­ity of different steps. Often the adjustments to intermediate process streams are as critical to the success of the overall process as are the actual separa­tions themselves.

The selection of separations steps requires understanding the fundamen­tal chemistry and engineering involved. A process that looks attractive chemically may prove untenable industrially because of unforeseen chemi­cal interactions, poor process behavior, or instability under process condi­tions (temperature, pH, radiation, etc.). The selection is first done by choosing a candidate chemical system that is known to accomplish the desired separation. Design of a detailed process flowsheet that will yield the desired goal then follows. Optimization of the process flowsheet requires a sound understanding of the interactions between different process vari­ables. Such an understanding can best be gained through high fidelity models of both the chemistry and the unit operations.

Laboratory data are used to develop or refine computer models to simu­late the process. Bench-scale process flowsheets are then designed from these simulations and tested, first with simulants then, if practical, with actual spent fuel or radioactive waste. Quite often the genuine feed per­forms differently than simulants, even when the feed is relatively simple in composition. It is critical that a flowsheet be developed from small-scale tests, preferably on prototypical equipment and under prototypical condi­tions to provide data on process kinetics, hydraulics, parameter selection, and any other effects arising from process variability. The use of prototypi­cal equipment is critical as it reduces the number of variables that must be accounted for during scale-up, thus reducing process uncertainty.

Although the heart of a separations facility is the chemical processing (solvent extraction, ion exchange, etc.), precisely recovering the targeted components and stabilizing the separated constituents is not a trivial opera­tion. In fact, the footprint of the post-separations processing is often larger than that required for the separations themselves. In nuclear systems, the footprint is a major driver of the process economics because of the shielding requirements. As a result, great effort must be made to select additives
and complexants that minimize the potential to adversely affect final stabilization.

Metanitrobenzotrifluoride (F-3) is known to be used as the diluent for CCD at the operational industrial facility UE-35 ( PA “Mayak”, Ozersk, Russia), meeting all the requirements imposed in Russia on diluents being applied in radiochemistry. In the USA, there are severe restrictions placed on the use of nitroaromatic compounds; therefore, during the joint development (by RI and INL) of a double-extractant workflow (CCD + POR) for treatment of acidic wastes in Idaho, a search was conducted for some new diluents of CCD containing no nitrogroups. In this connection, preference was given to those organofluoric compounds having high chemical and radiation resis­tance, explosion-fire safety and acceptable hydrodynamic characteristics.

Amongst the investigated organofluoric compounds (ethers, esters, ketones, sulfones) particular attention was paid to the class of fluorinated sulfones. Sulfones are compounds of formula RSO2RF, where R — aryl or alkyl, Rf — polyfluoroalkyl CnF2n+i; they have many common properties with nitroaromatic compounds, because the sulfonic group is an acceptor, espe­cially in polyfluoroalkylsulfones, and also forms a conjugate system with an aromatic ring.

A series of polyfluoroalkylsulfones with different substituents was syn­thesized and their main characteristics were determined with respect to the possibility of using them as a diluent for CCD. With an allowance for the total complex of properties, phenyltrifluoromethylsulfone (FS-13) proved to be the most promising. In Table 9.7 the main properties of FS-13 are shown and compared to those of metanitrobenzotrifluoride (F-3) which is used at UE-35 PA “Mayak”.

Table 9.8 presents the extraction properties of CCD in FS-13 as com­pared to CCD diluted with F-3. It follows from the data in Tables 9.7 and 9.8 that phenyltrifluoromethylsulfone (FS-13) possesses high density, mod­erate viscosity and low solubility in nitric acid, while the solubility of CCD in FS-13 is rather high. The extraction ability of CCD in FS-13 is somewhat lower than in F-3, but quite sufficient for successful extraction

of radionuclides from HNO3 solutions. These results provide reason to investigate the possibility of using FS-13 as a diluent of the universal extract­ant containing CCD, CMPO and PEG. The extraction properties of mix­tures based on FS-13 and F-3 are shown in Table 9.9.

Taking all of the above data, it can be concluded that phenyltrifluoro — methylsulfone (FS-13) as well as metanitrobenzotrifluoride (F-3) could be

9.2 Components of universal extraction mixture (UNEX-extractant).

feasible for dilution of the universal mixture. Hence, the universal extrac­tion mixture should contain the following components (Fig. 9.2): chlori­nated cobalt dicarbolide, diphenyl-N, N-dibutylcarbamoylphosphineoxide, polyethylene glycol and phenyltrifluoromethylsulfone.

The counter current contactor is another important piece of equipment used to recover actinides from the salt in electrorefiners with separating lanthanide FPs. A high-temperature centrifugal contactor, a pyrocontactor, was developed by ANL (Laidler, 1998). The design of the pyrocontactor was based on an aqueous contactor (Leonard, 1988) to achieve intensive mixing by chemical reactions between solutes in both the molten salt and liquid metals, and to ensure the clean separation of these streams. The four — stage pyrocontactor was installed in an Ar glove box, and extraction tests were carried out using Ce, La and Y as substitutes for U, Pu(MA) and lanthanides, respectively. Although details of both the design and the exper­imental conditions were not described in the literature, stage efficiencies approaching 99% of the theoretical value were reported at rotor speeds of nearly 3200 rpm (Laidler, 1998). By contrast, CRIEPI has developed a dif­ferent type of counter current contactor that requires a rotation speed of only approximately 300 rpm with the aim of increasing long-term integrity. Because radiation damage on molten salt and liquid metal is negligible compared with that of organic solvents used in aqueous reprocessing, the

flow rate of the counter current contactor can be decreased. Using the material balance calculated for commercial throughput, the required flow rate was sufficiently slow to complete reduction reaction without strong mixing. The large density difference between molten salt (1.7 g/cm3) and liquid Cd (7.8 g/cm3) eases the phase separation problem. On the basis of this consideration, a three-stage counter current contactor was developed and installed in an Ar glove box, as shown in Fig. 10.18. The contactor was connected to four separate tanks used for salt supply, Cd supply, salt recov­ery and Cd recovery, respectively. Each stage of the contactor has a capacity of 300 ml molten salt and 300 ml liquid Cd. Molten salt and liquid Cd were supplied from the tanks at a constant rate of 10 to 50 ml/min. Extraction tests were carried out using molten salt and liquid Cd at 450 °C. Three rare earth elements, Ce, Gd and Y, were used as substitutes for U, transuranic elements and rare earth FPs, respectively. In three-stage counter current extraction test, a high recovery ratio of close to 100% with efficient separa­tion was achieved (Kinoshita, 2007, 2008).

12.1.1 Critical PWR and FR

The main restriction to introducing minor actinides into critical reactors is linked to their impact on the core’s reactivity and kinetic parameters [3-5]. This would produce:

• a drop in the fuel temperature coefficients (Doppler effect),

• an increase in reactivity effect linked to the coolant voiding,

• a reduction in delayed neutron yield.

For instance, the impacts of these factors as a function of minor actinides content loading are given for the case of a large fast reactor in Fig. 12.3. The minor actinides content limit is around 3% of total heavy nuclides for large sodium cooled reactors. The gas cooled fast reactor allows a higher

loading in MA than liquid metal fast reactor due to the disappearance void effect constraint: 5% of MA seems acceptable even for a large core.

The void effect is also actually the most restricting criterion for MA loading in PWR from the point of view of its impact on physics and safety. Without the coolant, the neutron spectrum moves to higher energy and the minor actinides contributions of the thermal and epithermal resonances vanish. The minor actinide content must then be limited to about 1% of total heavy nuclides.

For critical reactors, some conclusions can be proposed:

• The admissible quantities of minor actinides in the core have to be kept low (about 1% for PWR type and 3% to 5% for FR type).

• The maximum MA fission rates are about 5 to 10% in PWRs and from 15 to 30% in FRs. In both cases, a multi recycling is necessary in order to reach satisfactory overall performance.

• In PWRs, curium recycling is to be avoided as it produces by itself and by producing californium, an intense source of neutrons. In FRs, curium recycling also produces upper elements, but they stabilize at a far lower level than in the PWRs and therefore do not pose any specific new problem.

Fast spectrum reactors can operate as breeders or burners, or in a self­sufficient breakeven mode. Breeders incorporate external blankets, both axial and radial. When reflectors replace blankets, FRs become net burners of fissile material. If an appropriate amount of blanket is incorporated, then a self-sufficient mode can be maintained. In whichever mode they operate, FR discharged fuel contains a large fraction of fissile inventory, and hence recycling is mandatory. As a matter of fact, resource utilization improve­ment is the primary rationale for fast reactors and recycling is required to achieve that goal.

Historically, only uranium and plutonium have been recovered from LWR spent fuels, and only plutonium recycling has taken place or been envisioned for the fast reactor fuel cycle. However, we have seen that minor actinides can be also recycled along with plutonium in fast reactors. Minor actinides and even-mass isotopes of plutonium may not be attractive as fuel for thermal reactors because they have unfavourable ratios of fission to capture, as demonstrated in the previous chapter. These same materials, as well as odd-mass isotopes of plutonium, are fissionable in fast spectrum, where we have seen that the fission to capture ratio is much more favour­able. Furthermore, the high content of recycled plutonium may require remote fabrication and hence minor actinides can be more easily incorpo­rated into continuous recycling.

For economic and political reasons and because of proliferation concerns in some countries, fast reactors and their fuel cycle development pro­grammes have been curtailed since the 1990s, except in China, France, India, Japan and the Russian Federation. Fast reactor concepts for actinide trans­mutation have been of interest in recent international initiatives such as the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) or Generation IV International Forum (GIF). Apart from these developments, the value of preserving the large technology base developed in Japan, France, Germany, the Russian Federation, the United Kingdom and the USA, as well as information developed in other countries, has been a subject of essential scientific interest.

Double layer system

Discussed here are PRBs specifically designed to utilize microorganisms in the treatment processes. A typical design comprises a double-layer with an aeration zone followed by the bioremediation zone. One such system was evaluated against the removal of methyl-tert-butyl-ether (MTBE) con­taminated groundwater (Fig. 15.6) (Liu et al., 2006). The aeration in this case was achieved chemically by the oxidation of calcium peroxide (CaO2) to release oxygen into the medium. Other growth nutrients were added

to encourage the growth of MTBE degrading organisms in the second layer.

Notably, inorganic salts such as potassium dihydrogen phosphate (KH2PO4) and ammonium sulphate ((NH4)2SO4) can act as buffers against pH changes caused by the oxidation of CaO2 into carbonates (CO32-). Thus, nutrients added in the second layer must include the phosphate buffer for the proper functioning of the barrier.

Another documented application is the treatment of petrochemical pol­lutants (i. e., benzene, toluene, ethylbenzene, xylene and polyaromatic hydrocarbons), heavy metals (i. e., lead, arsenic, etc.), and cyanide in the system designed by Doherty et al. (2006) using a modified ash system. The BPRB system was implemented at an abandoned gas manufacturing plant after 150 years of operation.

Specific application of the biological permeable reactive barrier (BPRB) system for the removal of Cr(VI) in groundwater has not been attempted. This has been both due to the unavailability of microorganisms capable of growing under nutrient deficient conditions and lack of information on the fate of the reduced chromium species in the barrier.