Degradation Studies of Cyanex 301

International audienceDespite the numerous studies found in the literature on CYANEX® 301, very few explain its degradation in depth. To the best of our knowledge none has explained the inconsistency between the “common knowledge” of “CYANEX® 301 degrades into CYANEX® 272” (dithiophosphinic acid degrading into the corresponding phosphinic acid) and the 31P spectrum obtained by NMR of the degradation compound. The 31P {1H} NMR analysis of a solution of CYANEX® 301 in prolonged contact with nitric acid shows a very complex spectrum, with resonances about 20 ppm downfield from what could have been expected.The degradation product giving those multiple resonances in a pattern that could be interpreted as a triplet of triplet is actually a dimer, where two molecules of CYANEX® 301 are linked by a disulfide bridge, corresponding to the condensation of the SH groups. The explanation of the complexity of the spectrum comes from the comparison with the spectrum obtained for the degradation of a stereoisomerically-purified CYANEX® 301. This purification led to the removal of the [R;S] and [S;R] isomers from the initial mixture, and yielded a white crystalline solid proven to comprise exclusively [R;R] and [S;S] isomers by XRD analysis. It was determined that the carbon chirality induced an asymmetry of the phosphorus atoms upon condensation, leading to a wide combination of magnetically non-equivalent P-31 nuclei, which can also exhibit coupling through the S-S bond The complete explanation of the NMR spectra was established and corroborated by elemental analysi

Summary of TRUEX Radiolysis Testing Using the INL Radiolysis Test Loop

The INL radiolysis and hydrolysis test loop has been used to evaluate the effects of hydrolytic and radiolytic degradation upon the efficacy of the TRUEX flowsheet for the recovery of trivalent actinides and lanthanides from acidic solution. Repeated irradiation and subsequent re-conditioning cycles did result in a significant decrease in the concentration of the TBP and CMPO extractants in the TRUEX solvent and a corresponding decrease in americium and europium extraction distributions. However, the build-up of solvent degradation products upon {gamma}-irradiation, had little impact upon the efficiency of the stripping section of the TRUEX flowsheet. Operation of the TRUEX flowsheet would require careful monitoring to ensure extraction distributions are maintained at acceptable levels

Characterization of radiolytically generated degradation products in the strip section of a TRUEX flowsheet

This report presents a summary of the work performed to meet the FCRD level 2 milestone M3FT-13IN0302053, “Identification of TRUEX Strip Degradation.” The INL radiolysis test loop has been used to identify radiolytically generated degradation products in the strip section of the TRUEX flowsheet. These data were used to evaluate impact of the formation of radiolytic degradation products in the strip section upon the efficacy of the TRUEX flowsheet for the recovery of trivalent actinides and lanthanides from acidic solution. The nominal composition of the TRUEX solvent used in this study is 0.2 M CMPO and 1.4 M TBP dissolved in n-dodecane and the nominal composition of the TRUEX strip solution is 1.5 M lactic acid and 0.050 M diethylenetriaminepentaacetic acid. Gamma irradiation of a mixture of TRUEX process solvent and stripping solution in the test loop does not adversely impact flowsheet performance as measured by stripping americium ratios. The observed increase in americium stripping distribution ratios with increasing absorbed dose indicates the radiolytic production of organic soluble degradation compounds

Aspects of the Fundamental Chemistry of Cesium Extraction from Acidic Media by HCCD

The unique extraction properties of univalent polyhedral borate anions, are well known and have been extensively studied over the past three decades. This is particularly true of the hexachlorinated derivative of the chloro-protected cobalt bis(dicarbollide) anion [(8, 9, 12-Cl3-C2B9H8)2-3-Co]-, (CCD-), typically in the acid form (HCCD) and dissolved in a suitably polar diluent, such as nitrobenzene, which is known to have a high affinity for selective extraction of the Cs+ cation. Recent collaborations between Russian and USA researchers expanded the use of HCCD in the Universal Extraction (UNEX) process where Cs, Sr, actinides (An) and lanthanides (Ln) are all extracted simultaneously by incorporating a neutral extractant (specifically diphenyl-N,N-di-n-butylcarbamoylmethyl phosphine oxide, CMPO) with HCCD and PEG-400 in the organic diluent phenyl trifluoromethyl sulfone (FS-13). In recent efforts to understand the complicated and unique synergistic chemical phenomena associated with simultaneous radionuclide (Cs, Sr, An, Ln) in the UNEX process, additional insight into Cs extraction by the HCCD system has been obtained. Four data sets with 25 experimental measurements of Cs distribution ratios, DCs [Cs]org/[Cs]aq, at a variety of initial conditions (various [HCCD] and [HNO3]) have been modeled using the SXLSQI computer program developed at ORNL. The SXLSQI program was used in this analysis to help elucidate the general chemical equilibria operative in the extraction of Cs+ into an organic phase comprised of HCCD in FS-13. The experimental data is best modeled with the following (simplified) chemical equilibria and the associated equilibrium constants (T = 25°C): (1) (2) (3) Where the over bar represents species formed in the organic phase. The equilibrium constant for the primary exchange reaction (1) of log Keq = 3.07 is in excellent agreement with values reported in the literature of log K = 3.00 for the dicarbollide/ nitrobenzene system. In general, the equilibria representing the mechanism of Cs extraction by HCCD is consistent with earlier works reported in the literature, albeit derived by different experimental and modeling schemes. The details of the experimental and modeling efforts are summarized in this work

Flowsheet Testing of the Fission Product Extraction Process as Part of Advanced Aqueous Reprocessing

As part of the Advanced Fuel Cycle Initiative (AFCI), the reduction in volume and heat generation of spent nuclear fuel requiring geologic disposal is currently being addressed. The goal is to optimize utilization of the nation’s first repository and potentially reduce or eliminate the need for additional repositories. This will be achieved through separating long-lived, highly toxic elements, reducing high-level waste volumes and the toxicity of spent nuclear fuel, and reducing the heat generation of spent nuclear fuel. The Idaho National Laboratory (INL) is working closely with a team of national laboratories and other organizations to support development of these separations processes. Key to the reduction of short-term heat load in a geological repository is the separation of 137Cs and 90Sr. Removal of these highly radioactive fission products reduces the short-term (~100 yr) heat generation source of the spent nuclear fuel. Once separated, the Cs and Sr would be placed in storage until the activity has decayed to LLW levels, at which time it could potentially be disposed of as a non-transuranic (TRU) low-level waste (LLW)

TRUEX Radiolysis Testing Using the INL Radiolysis Test Loop: FY-2012 Status Report

The INL radiolysis test loop has been used to evaluate the affect of radiolytic degradation upon the efficacy of the strip section of the TRUEX flowsheet for the recovery of trivalent actinides and lanthanides from acidic solution. The nominal composition of the TRUEX solvent used in this study is 0.2 M CMPO and 1.4 M TBP dissolved in n-dodecane and the nominal composition of the TRUEX strip solution is 1.5 M lactic acid and 0.050 M diethylenetriaminepentaacetic acid. Gamma irradiation of a mixture of TRUEX process solvent and stripping solution in the test loop does not adversely impact flowsheet performance as measured by stripping americium ratios. The observed increase in americium stripping distribution ratios with increasing absorbed dose indicates the radiolytic production of organic soluble degradation compounds

ADVANCED TECHNOLOGIES FOR THE SIMULTANEOUS SEPARATION OF CESIUM AND STRONTIUM FROM SPENT NUCLEAR FUEL

Two new solvent extraction technologies have been recently developed to simultaneously separate cesium and strontium from spent nuclear fuel, following dissolution in nitric acid. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. This new strip reagent reduces product volume by a factor of 20, over the baseline process. Countercurrent flowsheet tests on simulated spent nuclear fuel feed streams have been performed with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4',4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance

Development of Technologies for the Simultaneous Separation of Cesium and Strontium from Spent Nuclear Fuel as Part of an Advanced Fuel Cycle

As part of the Advanced Fuel Cycle Initiative, two solvent extraction technologies are being developed to simultaneously separate cesium and strontium from dissolved spent nuclear fuel. The first process utilizes a solvent consisting of chlorinated cobalt dicarbollide and polyethylene glycol extractants in a phenyltrifluoromethyl sulfone diluent. Recent improvements to the process include development of a new, non-nitroaromatic diluent and development of new stripping reagents, including a regenerable strip reagent that can be recovered and recycled. Countercurrent flowsheets have been designed and tested on simulated and actual spent nuclear fuel feed streams with both cesium and strontium removal efficiencies of greater than 99 %. The second process developed to simultaneously separate cesium and strontium from spent nuclear fuel is based on two highly-specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. A solvent composition has been developed that enables both elements to be removed together and, in fact, a synergistic effect was observed with strontium distributions in the combined solvent that are much higher that in the strontium extraction (SREX) process. Initial laboratory test results of the new combined cesium and strontium extraction process indicate good extraction and stripping performance. A flowsheet for treatment of spent nuclear fuel is currently being developed

Comparison of Aromatic Dithiophoshinic and Phosphinic Acid Derivatives for Minor Actinide Extraction

A new extractant for the separation of actinide(III) and lanthanide(III), bis(otrifluoromethylphenyl) phosphinic acid (O-PA) was synthesized. The synthetic route employed mirrors one that was employed to produce the sulfur containing analog bis(otrifluoromethylphenyl) dithiophosphinic acid (S-PA). Multinuclear NMR spectroscopy was used for elementary characterization of the new O-PA derivative. This new O-PA extractant was used to perform Am(III)/Eu(III) separations and the results were directly compared to those obtained in identical separation experiments using S-PA, an extractant that is known to exhibit separation factors of ~100,000 at low pH. The separations data are presented and discussed in terms comparing the nature of the oxygen atom as a donor to that of the sulfur atom in extractants that are otherwise identical

Development of Cesium and Strontium Separation and Immobilization Technologies in Support of an Advanced Nuclear Fuel Cycle

As part of the Advanced Fuel Cycle Initiative, two solvent extraction technologies are being developed at the Idaho National Laboratory to simultaneously separate cesium and strontium from dissolved spent nuclear fuel. The chlorinated cobalt dicarbollide/polyethylene glycol (CCD/PEG) process utilizes a solvent consisting of chlorinated cobalt dicarbollide for the extraction of Cs and polyethylene glycol for the synergistic extraction of Sr in a phenyltrifluoromethyl sulfone diluent. Countercurrent flowsheets have been designed and tested on simulated and actual spent nuclear fuel feed streams with both cesium and strontium removal efficiencies of greater than 99%. The Fission Product Extraction (FPEX) process is based on two highly-specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) for the extraction of Sr and Calix[4]arene-bis-(tert-octylbenzo-crown-6) (BOBCalixC6) for the extraction of Cs. Laboratory test results of the FPEX process, using simulated feed solution spiked with radiotracers, indicate good Cs and Sr extraction and stripping performance. A preliminary solvent extraction flowsheet for the treatment of spent nuclear fuel with the FPEX process has been developed, and testing of the flowsheet with simulated spent nuclear fuel solutions is planned in the near future. Steam reforming is currently being developed for stabilization of the Cs/Sr product stream because it can produce a solid waste form while retaining the Cs and Sr in the solid, destroy the nitrates and organics present in these aqueous solutions, and convert the Cs and Sr into leach resistant aluminosilicate minerals. A bench-scale steam reforming pilot plant has been operated with several potential feed compositions and steam reformed product has been generated and analyzed