Efficacy and Safety of Meropenem\u2013Vaborbactam Versus Best Available Therapy for the Treatment of Carbapenem-Resistant Enterobacteriaceae Infections in Patients Without Prior Antimicrobial Failure: A Post Hoc Analysis

open5siIntroduction: Infections due to Klebsiella pneumoniae carbapenemase (KPC)-producing Enterobacteriaceae are associated with increased morbidity and high mortality. Meropenem–vaborbactam (MV) is a novel β-lactam/β-lactamase inhibitor combination active against KPC-producing Enterobacteriaceae. The aim of this post hoc analysis of the TANGO-II randomized controlled trial was to assess the efficacy of MV versus best available therapy (BAT) in the subgroup of patients without prior antimicrobial failure. Methods: The primary outcome measure was clinical cure at the test of cure (TOC). Secondary outcome measures included (1) clinical cure at the end of therapy (EOT), (2) microbiological cure at TOC, (3) microbiological cure at EOT, and (4) 28-day all-cause mortality. Results: First-line MV was associated with a 42.9% absolute increase in clinical cure rate at TOC (95% confidence intervals [CI] 13.7–72.1) in comparison with first-line BAT. A 49.3% absolute increase in clinical cure rate at EOT (95% CI 20.8–77.7), a 42.6% absolute increase in microbiological cure rate at EOT (95% CI 13.4–71.8), and a 36.2% absolute increase in microbiologic cure rate at TOC (95% CI 5.9–66.6) were also observed, in addition to a 29.0% absolute reduction in mortality (95% CI − 54.3 to − 3.7). Overall, fewer adverse events were observed in the MV group than in the BAT group. Conclusion: MV was superior to BAT in the subgroup of patients with serious carbapenem-resistant Enterobacteriaceae (CRE) infections and no prior antimicrobial failure, with very high rates of clinical success, and was well tolerated. Post approval and real-world studies remain essential to clearly define the most appropriate population for early, empirical MV coverage, in accordance with antimicrobial stewardship principles. Funding: The Medicines Company.openBassetti M.; Giacobbe D.R.; Patel N.; Tillotson G.; Massey J.Bassetti, M.; Giacobbe, D. R.; Patel, N.; Tillotson, G.; Massey, J

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

Processing Irradiated Beryllium For Disposal

The purpose of this research was to develop a process for decontaminating irradiated beryllium that will allow it to be disposed of through normal radwaste channels. Thus, the primary objectives of this ongoing study are to remove the transuranic (TRU) isotopes to less than 100 nCi/g and remove {sup 60}Co, and {sup 137}Cs, to levels that will allow the beryllium to be contact handled. One possible approach that appears to have the most promise is aqueous dissolution and separation of the isotopes by selected solvent extraction followed by precipitation, resulting in a granular form for the beryllium that may be fixed to prevent it from becoming respirable and therefore hazardous. Beryllium metal was dissolved in nitric and fluorboric acids. Isotopes of {sup 241}Am, {sup 239}Pu, {sup 85}Sr, and {sup 137}Cs were then added to make a surrogate beryllium waste solution. A series of batch contacts was performed with the spiked simulant using chlorinated cobalt dicarbollide (CCD) and polyethylene glycol diluted with sulfone to extract the isotopes of Cs and Sr. Another series of batch contacts was performed using a combination of octyl (phenyl)-N,N-diisobutylcarbamoylmethylphosphine oxide (CMPO) in tributyl phosphate (TBP) diluted with dodecane for extracting the isotopes of Pu and Am. The results indicate that greater than 99.9% removal can be achieved for each isotope with only three contact stages

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

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

Actinide partitioning studies using dihexyl-N,N-diethycarbamolymehtyl phosphonate and dissolved zirconium calcine

A baseline flowsheet capable of partitioning the transuranic (TRU) elements from dissolved zirconium calcines has been developed. The goal of the TRU partitioning process is to remove the TRUs from solutions of dissolved zirconium calcines to below the 10 CFR 61.55 Class A waste limit of 10 nCi/g. Extraction, scrub, strip, and wash distribution coefficients for several elements, including the actinides, were measured in the laboratory by performing equal volume batch contacts. A solvent containing diheyl-N, N- diethylcarbamoylmethyl phosphonate (CMP), tributylphosphate (TBP), and a branched chain hydrocarbon as the diluent were used to develop this process. A non-radioactive zirconium pilot-plant calcine was spiked with the TRUs, U, Tc, or a radioactive isotope of zirconium to simulate the behavior of these elements in actual dissolved zirconium calcine feed. Distribution coefficient data obtained from laboratory testing were used to recommend: (1) solvent composition, (2) scrub solutions capable of selectively removing extracted zirconium while minimizing actinide recycle, (3) optimized strip solutions which quantitatively recover extracted actinides, and (4) feed adjustments necessary for flowsheet efficiency. Laboratory distribution coefficients were used in conjunction with the Generic TRUEX Model (GTM) to develop and recommend a flowsheet for testing in the 5.5-cm Centrifugal Contractor Mockup. GTM results indicate that the recommended flowsheet should remove the actinides from dissolved zirconium calcine feed to below the Class A waste limit of 10 nCi/g. Less than 0.01 wt% of the extracted zirconium will report to the high- activity waste (HAW) fraction using the 0.05 M H{sub 2}C{sub 2}O{sub 4} in 3.0 M HNO{sub 3} scrub, and greater than 99% of the extracted actinides are recovered with 0.001 M HEDPA

Separation of Minor Actinides from Lanthanides by Dithiophosphinic Acid Extractants

The selective extraction of the minor actinides (Am(III) and Cm(III)) from the lanthanides is an important part of advanced reprocessing of spent nuclear fuel. This separation would allow the Am/Cm to be fabricated into targets and recycled to a reactor and the lanthanides to be dispositioned. This separation is difficult to accomplish due to the similarities in the chemical properties of the trivalent actinides and lanthanides. Research efforts at the Idaho National Laboratory have identified an innovative synthetic pathway yielding new regiospecific dithiophosphinic acid (DPAH) extractants. The synthesis provides DPAH derivatives that can address the issues concerning minor actinide separation and extractant stability. For this work, two new symmetric DPAH extractants have been prepared. The use of these extractants for the separation of minor actinides from lanthanides will be discussed

Quantitative structure of an acetate dye molecule analogue at the TiO2- acetic acid interface

The positions of atoms in and around acetate molecules at the rutile TiO2(110) interface with 0.1 M acetic acid have been determined with a precision of ±0.05 Å. Acetate is used as a surrogate for the carboxylate groups typically employed to anchor monocarboxylate dye molecules to TiO2 in dye-sensitised solar cells (DSSC). Structural analysis reveals small domains of ordered (2 x 1) acetate molecules, with substrate atoms closer to their bulk terminated positions compared to the clean UHV surface. Acetate is found in a bidentate bridge position, binding through both oxygen atoms to two five-fold titanium atoms such that the molecular plane is along the [001] azimuth. Density functional theory calculations provide adsorption geometries in excellent agreement with experiment. The availability of these structural data will improve the accuracy of charge transport models for DSSC

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