A novel rapid detection approach for the analysis of radionuclides in environmental samples using graphite MALDI mass spectrometry

In nuclear technology, tri-n-butyl phosphate (TBP) diluted with a hydrocarbon diluent such as n-dodecane or NPH is the most frequently used solvent in liquid–liquid extraction for fuel reprocessing. This extraction, known as the plutonium uranium refining by extraction, is still considered as the most dominant process for the extraction of uranium and plutonium from irradiated fuels. The solubility of pure TBP in water is about 0.4 g/L at 25 °C. This is enough to create trouble during evaporation of raffinate and product solutions. Solubility data for undiluted TBP and TBP (diluted in inert hydrocarbon diluent) in various concentrations of nitric acid is not adequate in the literature. The solubility data generated in the present study provide complete information on the solubility of TBP in various nitric acid concentrations (0–15.7 M) at room temperature. The effect of heavy metal ion concentration such as uranium and various fission products on the solubility of TBP in nitric acid is also presented. The results obtained from gas chromatographic technique were compared with spectrophotometric technique by converting the organic phosphate into inorganic phosphate. The generated data is of direct relevance to reprocessing applications.     

Hexanoic acid as an alternative diluent in a GANEX process: feasibility study

Used nuclear fuel is radiotoxic for mankind and its environment for a long time. However, if it can be transmuted, the radiotoxicity as well as its heat load are reduced. Before a transmutation the actinides within the used fuel need to be separated from the fission, corrosion and activation products. This separation can be achieved by using the liquid–liquid extraction technique. One extraction process that can be used for such a separation is the Group ActiNide EXtraction (GANEX) process. One GANEX process that can successfully accomplish the separation utilizes the diluent cyclohexanone in combination with the extractant tributylphosphate (TBP) (30 % vol) and a second extractant, CyMe₄-BTBP (10 mM). However, there are some issues when using cyclohexanone as diluent. In this work an alternative diluent has therefore been tried in order to determine if it can replace cyclohexanone. The diluent used was hexanoic acid. In a system containing 10–12 mM CyMe₄-BTBP and 30 % vol TBP in hexanoic acid with the aqueous phase 4 M HNO₃, the distribution ratios for americium and curium are unfortunately low (D _Am = 1.1 ± 0.27, D _Cm = 1.6 ± 1.81). The concentration of CyMe₄-BTBP ligand, the extractant of curium and americium, could unfortunately not be increased, because of limited solubility in hexanoic acid. The distribution ratios for fission, corrosion and activation products were low for most metals; however, cadmium, palladium and molybdenum all unfortunately have distributions ratios above 1. To conclude, low americium and curium extractions indicate that hexanoic acid is not a suitable diluent which could replace cyclohexanone in a GANEX process.     

Effect of tri-n-butyl phosphate on physical properties of dodecane–nitric acid system

Extraction power of solvent depends upon the physical properties of the system. Tri-n-butyl phosphate (TBP) in dodecane is a versatile solvent used in the nuclear fuel reprocessing like PUREX process. The study of physical properties like density, viscosity, interfacial tension and solubility for TBP–nitric acid–dodecane system will be helpful in carrying out different extraction studies during PUREX process. Thus, physical properties like density, viscosity, interfacial tension and solubility have been measured for TBP–nitric acid–dodecane system using pycnometer, viscometer, pendant drop method and high performance liquid chromatography respectively. It has been observed that density and viscosity increases but interfacial tension and solubility decreases with the concentration of TBP in dodecane–nitric acid system. Physical properties of 30 % TBP–nitric acid–dodecane system have also been studied in detail. All these studies will also be useful in stripping out dissolved TBP from the nuclear waste.     

Quantitative determination of tri-n-butyl phosphate by fluorescent X-ray spectrometry through inorganic association with an actinide

Tri-n-butyl phosphate (TBP) continues to be the most widely used solvent in nuclear fuel extraction, refining and reprocessing units for the extraction of actinides and their separation from fission products. An X-ray fluorescence spectrometric method (XRFS) for the determination of TBP content with an X-ray detectable element is presented. The method involves formation of an ion association complex of uranium with TBP-kerosene mixture in 3M nitric acid. The analytes uranium and bromine used as internal ratio elements in organic extract are excited by a primary X-ray beam from a rhodium tube. The solvent concentration is determined from the ratioed characteristic intensities of uranium and bromine. The procedure permits the determination of organic solvent in the range 0.5 to 5.0% with a relative standard deviation of 0.1%.     

New reprocessing system for spent nuclear reactor fuel using fluoride volatility method

Our proposed spent nuclear fuel reprocessing technology named FLUOREX, which is a hybrid system using fluoride volatility and solvent extraction, meets the requirements of the future thermal/fast breeder reactors (coexistence) cycle. We have been done semi-engineering and engineering scale experiments on the fluorination of uranium, purification of UF₆, pyrohydrolysis of fluorination residues, and dissolution of pyrohydrolysis samples in order to examine technical and engineering feasibilities for implementing FLUOREX. We found that uranium in spent fuels can be selectively volatilized by fluorination in the flame type reactor, and the amount of uranium volatilized is adjusted from 90% to 98% by changing the amount of F₂ supplied to the reactor. The volatilized uranium is purified using UO₂F₂ adsorber for plutonium and purification methods such as condensation and chemical traps for fission products provide a decontamination factor of over 10⁷. Most of the fluorination residues that consist of non-volatile fluorides of uranium, plutonium, and fission products are converted to oxides by pyrohydrolysis at 600-800 °C. Although some fluorides of fission products such as alkaline earth metals and lanthanides are not converted completely and fluorine is discharged into the solution, oxides of U and Pu obtained by pyrohydrolysis are dissolved into nitric acid solution because of the low solubility of lanthanide fluorides. These results support our opinion that FLUOREX has great possibilities for being a part of the future spent nuclear fuel cycle system.     

The EURO-GANEX Process: Current Status of Flowsheet Development and Process Safety Studies

A new hydrometallurgical grouped actinide extraction process has been developed to separate the transuranic actinide ions from dissolved spent fuel solution (after an initial uranium extraction cycle). This “EURO-GANEX” process is aimed towards the homogeneous recycling of plutonium and minor actinides in a future closed fuel cycle. The separation process is based on the co-extraction of actinides and lanthanides from aqueous nitric acid into an organic phase followed by selective co-stripping of actinides. A suitable organic phase has been formulated and distribution ratios determined for lanthanides, actinides and some problematic fission products under extraction and stripping conditions. The process flowsheet has been proven on surrogate feed solutions as well as with spent fast reactor fuel; excellent recoveries of the actinides and good decontamination factors from the lanthanides and other fission products were obtained. A variation on the EURO-GANEX flowsheet (the “TRU-SANEX” process) has now been designed to produce separate Pu+Np and Am+Cm products for heterogeneous recycling. Progress on underpinning process chemistry and safety studies as well as flowsheet tests are summarized.     

Insight of solvent extraction process: Reassessment of trace level determinations

Solvent extraction is hoary yet modern technique with great scope of research due to the various intriguing phenomena in the system. Tri-n-butyl phosphate (TBP) is a well known extractant which has been extensively used for separation of uranium matrix prior to elemental profiling. In this paper, one of the impurities namely Fe is being considered as it posed a challenge to the separation due to its co-extraction with TBP along with uranium. In these studies, for the first time, the existence of cation-cation inner sphere complexes between the UO₂²⁺and Fe³⁺ ions in both aqueous and organic phases have been establisted in addition to the selective separation of iron from uranium sample matrix using only TBP. The data from both spectrophotometric and thermophysical studies corroborated one another confirming the presence of cation-cation interactions (CCIs). The developed solvent extraction with only TBP showed almost no interferences on the iron extraction from matrix uranium and other co-ions like aluminum and copper. This has been the first time application of pure TBP for selective removal of iron from uranium samples. The procedure possessed excellent reproducibility and robustness.     

Gas-chromatographische Überwachung des Extraktionsmittels bei der Wiederaufarbeitung von HTR-Brennelementen

During reprocessing of HTR fuel elements by the THOREX solvent extraction process the tributylphosphate/n-alcane mixture used as extractant is subject to an intensive radiolytic and hydrolytic burden. Main degradation products are di- and monobutylphosphate which in quite low concentrations disturb the solvent extraction by retention of uranium and transfer of some fission products into the aqueous product solution. Both esters are removed from the irradiated solvent by washing with a sodium carbonate solution, after which the solvent may be recycled. For a control of the efficiency of this solvent recovery procedure as well as of the composition of the fresh and recycled solvent in the JUPITER reprocessing facility, gas chromatography will be used. This method allows the determination of di- and monobutylphosphate down to concentration levels of about 25 parts per million. An analysis requires about 100 min.     

Rapid quantification of TBP and TBP degradation product ratios by FTIR-ATR

Tri-n-butyl phosphate (TBP) is the key complexant within the plutonium and uranium reduction extraction process used to extract uranium and plutonium from used nuclear fuel. During reprocessing TBP degrades to dibutyl phosphate (DBP), butyl acid phosphate (MBP), butanol, and phosphoric acid over time. A method for rapidly monitoring TBP degradation is needed for the support of nuclear forensics. Therefore, a Fourier transform infrared spectrometry-attenuated total reflectance (FTIR-ATR) technique was developed to determine approximate peak intensity ratios of TBP and its degradation products. The technique was developed by combining variable concentrations of TBP, DBP, and MBP to simulate TBP degradation. This method is achieved by analyzing selected peak positions and peak intensity ratios of TBP and DBP at different stages of degradation. The developed technique was tested on TBP samples degraded with nitric acid. In mock degradation samples, the 1,235 cm^?1 peak position shifts to 1,220 cm^?1 as the concentration of TBP decreases and DBP increases. Peak intensity ratios of TBP positions at 1,279 and 1,020 cm^?1 relative to DBP positions at 909 and 1,003 cm^?1 demonstrate an increasing trend as the concentration of DBP increases. The same peak intensity ratios were used to analyze DBP relative to MBP whereas a decreasing trend is seen with increasing DBP concentrations. The technique developed from this study may be used as a tool to determine TBP degradation in nuclear reprocessing via a rapid FTIR-ATR measurement without gas chromatography analysis.     

Application of gas chromatography-tandem mass spectrometry to the analysis of inhibition of dimerisation of tributylphosphate under radiolysis. Identification of isomeric tributylphosphate-alkylbenzene inhibitor coupling products

Tributylphosphate (TBP), solvent used as extractant for reprocessing spent nuclear fuel, can dimerise under radiolysis. This occurs by radical radical recombination, leading to 10 isomeric dimers (TBP-TBP). These species are complexation agents and are responsible of fission product retention in the organic phase that increases the solvent degradation. In order to limit their formation two free radical inhibitors (In), isopropyl and 1,4-diisopropylbenzenes, were used. These additives reduce by about 50% the concentration of TBP-TBP dimers but this reduction is not strictly followed by TBP regeneration as mixed coupling products from TBP and inhibitor are detected. By using GC-MS-MS and selectively deuterated compounds, the identification of these different isomers (TBP-In) has been realised. From these identifications and from the analysis of the proportion of the different isomers, the major primary TBP radical generated under radiolysis was determined.     

CARBEX process,a new technology of reprocessing of spent nuclear fuel

Advances in the CARBEX process, a new aqueous chemical method for reprocessing of spent nuclear fuel (SNF) in carbonate media, are considered. A review of carbonate methods for SNF reprocessing is given. The CARBEX process concept is presented and experimental data for every stage of the CARBEX process: high-temperature oxidation of spent fuel composition, its oxidative dissolution in carbonate aqueous solutions, extraction refining of U(VI) and Pu(VI), solid-phase re-extraction of carbonate complexes of U(VI) and Pu(VI), and obtaining of uranium and plutonium dioxide powders for fabrication of ceramic nuclear fuel, are discussed. It was shown that the CARBEX process can be more effective and safe than the well-known industrial PUREX process.     

Comparing Branched versus Straight-chained Monoamide Extractants for Actinide Recovery

Separations of used nuclear fuel at the engineered scale have generally been completed using the Plutonium Uranium Redox Extraction (PUREX) process. The PUREX process uses tributyl phosphate (TBP) as an extractant to recover uranium and plutonium. While the TBP extractant has proven effective at recovering U and Pu at the engineered scale, TBP is potentially vulnerable to third phase formation and TBP degradation products (monobutyl and dibutyl phosphoric acids) which can complicate recovery of extracted metals from the organic phase. An alternative class of extractants, monoamides, has been considered for applications in thorium and uranium fuel cycles. When compared to TBP, monoamides tend to have higher separation factors for U or Pu from fission products, structural materials, and Th. This review summarizes the literature that explores actinide separations using monoamides by assessing the physiochemical properties between a broader library of branched and straight-chain monoamides than considered in previous reviews. An emphasis is placed on fine-tuning the selectivity of branched monoamides.     

Tri-iso-amyl phosphate (TAP): An alternative extractant to tri-butyl phosphate (TBP) for reactor fuel reprocessing

Tri-iso-amyl phosphate (TAP), an indigenously prepared extractant was utilized for reactor fuel reprocessing and compared with tri-butyl phosphate (TBP) and tri-n-hexyl phosphate (THP). The potential of these extractants was found to be in the order TAP>THP>TBP by calculating the acid uptake value (K _H). The effect of various parameters such as solvent degradation due to acid hydrolysis, radiation effect, decontamination factor and phase separation were investigated and it was found that TAP was always a better extractant in comparison to THP and TBP. In addition to this, the extraction of fission product contaminants such as ¹⁴⁴Ce, ¹³⁷Cs, ¹⁰⁶Ru, ⁹⁵Zr was almost negligible, even at very high nitric acid concentrations in the aqueous phase, indicating the potential application of TAP in actinide partitioning. Sodium carbonate solution or acidified distilled water was a good strippant for U(VI), similarly, uranium(IV) nitrate stripped Pu(IV) from the organic phase.     

Characterization of zirconium complexes of interest in spent nuclear fuel reprocessing by electrospray ionization mass spectrometry

Liquid-liquid extraction of zirconium, one of the most important fission products, was followed using electrospray ionization mass spectrometry under conditions simulating reprocessing of nuclear spent fuel. Zr(IV) can precipitate from the organic phase after extraction by dibutylphosphoric acid (HDBP), the most common degradation product of tributylphosphate (TBP) radiolysis. Different complexes were detected with electrospray used in positive or negative ion modes, according to the extraction conditions such as the ligand/metal ratio. Stoichiometry of the Zr(IV) complexes was determined by combining isotopic labeling [H(15)NO(3)] of the aqueous phase in the extraction system and tandem mass spectrometry experiments. These results were compared with the species observed using other techniques reported in the literature. The mechanisms of ionization/desorption of these complexes are proposed depending on the organic ligand character (neutral (L) such as TBP, or acidic (HL') such as HDBP), and the ionization mode used. Copyright 2000 John Wiley & Sons, Ltd.     

Tributyl phosphate degradation by immobilized cells of aCitrobacter sp.

Tributyl phosphate (TBP), a plasticizer and solvent, is used in nuclear fuel reprocessing, generating TBP wastes laden with residual uranium. ACitrobacter sp. accumulated heavy metals via a phosphohydrolase(s) that precipitated metals with inorganic phosphate liberated from an organic phosphate “donor” molecule (TBP). Mutant analysis suggested that TBP hydrolysis was not attributable to a previously documented acid phosphatase (monoesterase). Purified monoesterase had little activity against phospho di- and triesters, had no requirement for Mg²⁺ or Mn²⁺, and was EDTA-resistant. Conversely, TBP cleavage by immobilized cells was enhanced by Mg²⁺, and ininhibited by Mn²⁺ and EDTA. A separate phosphotri/diesterase was implicated.     

Separation of uranium from complex nuclear reaction product mixtures

A solvent extraction procedure for rapid separation of uranium from complex nuclear reaction product mixtures is suggested. The procedure has been tested in batch experiments with tracer amounts of representative elements. It has also been tested with fission products and uranium tracer using the continuous chemical separation system SISAK at the Mainz TRIGA reactor.     

Wet-oxidation of spent organic waste tributyl phosphate/diluents

Tributyl phosphate was used in reprocessing of spent nuclear fuel inthe Purex process. The amount of uranium retained in the organic phase dependson the type of TBP/diluent. Destruction of spent TBP is of high interest inwaste management. The use of the oxidative degradation of TBP diluted withkerosene, carbon tetrachloride, benzene and toluene using potassium permanganateas oxidant was carried out to produce stable inorganic dry particle residuewhich is then immobilized in different matrices. The different factors affectingthe destruction of spent waste were investigated. The uptake and decontaminationfactor for both ^{152, 154}Eu and ¹⁸¹Hf and the analysisof the final product have been studied.     

Improved determination of tributyl phosphate degradation products (mono- and dibutyl phosphates) by ion chromatography

Tributyl phosphate (TBP) is a very important compound in the nuclear industry, particularly in the area of nuclear fuel reprocessing. This compound is used in the PUREX (plutonium and uranium refining extraction) process which consists of the extraction of uranium and plutonium from an aqueous nitric acid phase, for the purpose of recycling. But TBP may be degraded to dibutyl phosphate (DBP) and monobutyl phosphate (MBP) by dealkylation of one or two butoxy groups, respectively. We have compared and evaluated the capacity of two resins manufactured by Dionex (AS11 and AS5A) in the separation and measurement of these two degradation products. AS11 generates two interferences: nitrite/DBP and carbonate/MBP. The first one is the most serious. So, we have developed a method for oxidising nitrite ions to nitrate ions which have no trouble over the measurement. The second resin tested, AS5A, allows a very efficient separation between DBP and NO₂⁻ ions and a good separation between MBP and CO₃²⁻ in comparison with the AS11. The detection limits for the AS5A column are 0.13 μM for MBP and 0.71 μM for DBP (injection LOOP=50 μl).     

Equilibrium and kinetics of calcium–uranyl–carbonate adsorption on silica nanoparticles

Solvent extraction studies on the purification of uranium from zirconium rich sodium diuranate (SDU) feed was carried out using n-tri butyl phosphate (TBP) as extractant and n-decanol as phase modifier. The presence of Zr in SDU leached solution leads to the formation of third phase during liquid–liquid extraction of uranium which was successfully prevented by addition of n-decanol in 30% (v/v) TBP/n-dodecane mixture. A seven stage counter current extraction of SDU feed solution followed by five stage stripping were carried out using optimum concentration of phase modifier 15% n-decanol-30% TBP in n-dodecane as solvent. Based on the findings a process flow-sheet has been developed for the purification of SDU to nuclear grade ammonium diuranate.     

Dissolution of uranium dioxide in supercritical fluid carbon dioxide

Uranium dioxide can be dissolved in supercritical CO2 with a CO2-philic TBP-HNO3 complexant to form a highly soluble UO2(NO3)(2).2TBP complex; this new method of dissolving UO2 that requires no water or organic solvent may have important applications for reprocessing of spent nuclear fuels and for treatment of nuclear wastes.