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Kamachi U Mudali - One of the best experts on this subject based on the ideXlab platform.
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dissolution behaviour of simulated mox nuclear fuel pellets in nitric acid medium
Progress in Nuclear Energy, 2019Co-Authors: N Desigan, N K Pandey, Kamachi U Mudali, Dasarath Maji, K Ananthasivan, J B JoshiAbstract:Abstract Data on the kinetics of dissolution of (U, Pu) mixed oxide (MOX) fuel pellets in nitric acid is an essential information required for the design of continuous dissolution system for the reProcessing of fast reactor spent fuel employing the plutonium uranium extraction (Purex) Process. In this article, the results obtained from a systematic investigation carried out on the dissolution of solid solutions of simulated (U, Ce) MOX pellets in nitric acid, with Ce as the non-radioactive surrogate of Pu, is presented. The influence of initial concentration of nitric acid, rate of mixing and temperature have been studied under typical Purex Process conditions. The apparent activation energy was found to be about 26 kJ/mol/K. The composition of the gases liberated (NOX) during dissolution was measured which is required for understanding the mechanism of the reaction.
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extraction studies of gadolinium relevant to its use as neutron poison in the Purex Process
Progress in Nuclear Energy, 2017Co-Authors: S Ganesh, N K Pandey, N Desigan, C Mallika, Kamachi U MudaliAbstract:Abstract The extraction behavior of gadolinium into tri- n -butyl phosphate (TBP) (0.37–1.1 M ) in n -dodecane ( n- DD) as diluent from different concentrations of nitric acid (1–13.5 M ) has been systematically investigated to establish the feasibility of using gadolinium as soluble neutron poison for ensuring nuclear criticality safety during the Processing of FBR spent fuel containing higher concentrations of fissile material. The effect of acidity, temperature, metal loading and gadolinium concentration on the distribution co-efficient of Gd(III) are established. The obtained results clearly indicate the feasibility to decontaminate Gd(III) under typical FBR fuel reProcessing conditions thereby enabling to qualify the final product to be pure from the point of view of nuclear poison impurity.
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modeling of pu vi distribution coefficients in 30 tbp based Purex Process
Journal of Radioanalytical and Nuclear Chemistry, 2013Co-Authors: Shekhar Kumar, Kamachi U MudaliAbstract:Hexavalent plutonium (Pu(VI)) is an important solute in the Purex (plutonium uranium extraction) Process. In 30 % TBP based Purex solvent extraction system, distribution coefficient of Pu(VI) is much lower than that of Pu(IV). This lower distribution coefficient of Pu(VI) may cause unexpected Pu loss during primary HA extraction in low acid flowsheets. An empirical model for Pu(VI) distribution coefficients in 30 % TBP and its temperature dependency has been reported in this paper. Comparison with literature data revealed a reasonably good agreement between the reported experimental and model predicted values.
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comparative studies on the determination of di n butyl phosphate in degraded solvent of Purex Process by ion chromatography and gas chromatography methods
Desalination and Water Treatment, 2012Co-Authors: P Velavendan, S Ganesh, N K Pandey, Kamachi U Mudali, R. NatarajanAbstract:Abstract This paper describes comparative studies on the determination of di-n-butyl phosphate (DBP) by ion chromatography (IC) and gas chromatography (GC) techniques in spent solvent of Purex Process used for the reProcessing of spent nuclear fuels. The ion chromatography method involves the separation of DBP from 30% TBP–NPH (tri-n-butylphosphate diluted in normal paraffin hydrocarbon) containing heavy metal ion like uranium and nitric acid by extraction of DBP into alkaline medium. DBP was subsequently eluted by ion-exchange separation in ion chromatography column and followed by suppressed conductivity detection. DBP is quantified to a lower limit of about 1 ppm with 3% RSD. However, in order to determine DBP by gas chromatography technique DBP is first quantitatively converted into its volatile and stable derivatives by using diazomethane prior to analysis by GC. Results obtained with ion chromatographic technique are compared with those of obtained by standard gas chromatographic technique. It was o...
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flow injection analysis of hydrazine in the aqueous streams of Purex Process by liquid chromatography system coupled with uv visible detector
Journal of Analytical Sciences Methods and Instrumentation, 2012Co-Authors: P Velavendan, N K Pandey, Kamachi U Mudali, N Pandey K S Ganesh, R. NatarajanAbstract:Present study describes the development of a rapid, sensitive and selective flow injection analysis of hydrazine in the aqueous streams of Purex Process by liquid chromatography system coupled with UV-Visible detector. The method is based on the formation of yellow coloured azine complex by reaction of hydrazine with para-dimethy laminobenzaldehyde (pDMAB). The formed yellow coloured complex is stable in acidic medium and has a maximum absorption at 460 nm. The presence of uranium in hydrazine solution is not interfering in the analysis. Under optimum condition, the absorption intensity linearly increased with the concentration of hydrazine in the range from 0.05-10 mg?L–1 with a correlation coefficient of R2=0.9999 (n=7). The experimental detection limit is 0.05mgL–1. The sampling frequency is 15 samples h–1 and the relative standard deviation was 2.1% for 0.05 mg?L–1. This method is suitable for automatic and continuous analysis and successfully applied to determine the concentration of hydrazine in the aqueous stream of nuclear fuel reProcessing.
N K Pandey - One of the best experts on this subject based on the ideXlab platform.
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dissolution behaviour of simulated mox nuclear fuel pellets in nitric acid medium
Progress in Nuclear Energy, 2019Co-Authors: N Desigan, N K Pandey, Kamachi U Mudali, Dasarath Maji, K Ananthasivan, J B JoshiAbstract:Abstract Data on the kinetics of dissolution of (U, Pu) mixed oxide (MOX) fuel pellets in nitric acid is an essential information required for the design of continuous dissolution system for the reProcessing of fast reactor spent fuel employing the plutonium uranium extraction (Purex) Process. In this article, the results obtained from a systematic investigation carried out on the dissolution of solid solutions of simulated (U, Ce) MOX pellets in nitric acid, with Ce as the non-radioactive surrogate of Pu, is presented. The influence of initial concentration of nitric acid, rate of mixing and temperature have been studied under typical Purex Process conditions. The apparent activation energy was found to be about 26 kJ/mol/K. The composition of the gases liberated (NOX) during dissolution was measured which is required for understanding the mechanism of the reaction.
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extraction studies of gadolinium relevant to its use as neutron poison in the Purex Process
Progress in Nuclear Energy, 2017Co-Authors: S Ganesh, N K Pandey, N Desigan, C Mallika, Kamachi U MudaliAbstract:Abstract The extraction behavior of gadolinium into tri- n -butyl phosphate (TBP) (0.37–1.1 M ) in n -dodecane ( n- DD) as diluent from different concentrations of nitric acid (1–13.5 M ) has been systematically investigated to establish the feasibility of using gadolinium as soluble neutron poison for ensuring nuclear criticality safety during the Processing of FBR spent fuel containing higher concentrations of fissile material. The effect of acidity, temperature, metal loading and gadolinium concentration on the distribution co-efficient of Gd(III) are established. The obtained results clearly indicate the feasibility to decontaminate Gd(III) under typical FBR fuel reProcessing conditions thereby enabling to qualify the final product to be pure from the point of view of nuclear poison impurity.
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comparative studies on the determination of di n butyl phosphate in degraded solvent of Purex Process by ion chromatography and gas chromatography methods
Desalination and Water Treatment, 2012Co-Authors: P Velavendan, S Ganesh, N K Pandey, Kamachi U Mudali, R. NatarajanAbstract:Abstract This paper describes comparative studies on the determination of di-n-butyl phosphate (DBP) by ion chromatography (IC) and gas chromatography (GC) techniques in spent solvent of Purex Process used for the reProcessing of spent nuclear fuels. The ion chromatography method involves the separation of DBP from 30% TBP–NPH (tri-n-butylphosphate diluted in normal paraffin hydrocarbon) containing heavy metal ion like uranium and nitric acid by extraction of DBP into alkaline medium. DBP was subsequently eluted by ion-exchange separation in ion chromatography column and followed by suppressed conductivity detection. DBP is quantified to a lower limit of about 1 ppm with 3% RSD. However, in order to determine DBP by gas chromatography technique DBP is first quantitatively converted into its volatile and stable derivatives by using diazomethane prior to analysis by GC. Results obtained with ion chromatographic technique are compared with those of obtained by standard gas chromatographic technique. It was o...
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flow injection analysis of hydrazine in the aqueous streams of Purex Process by liquid chromatography system coupled with uv visible detector
Journal of Analytical Sciences Methods and Instrumentation, 2012Co-Authors: P Velavendan, N K Pandey, Kamachi U Mudali, N Pandey K S Ganesh, R. NatarajanAbstract:Present study describes the development of a rapid, sensitive and selective flow injection analysis of hydrazine in the aqueous streams of Purex Process by liquid chromatography system coupled with UV-Visible detector. The method is based on the formation of yellow coloured azine complex by reaction of hydrazine with para-dimethy laminobenzaldehyde (pDMAB). The formed yellow coloured complex is stable in acidic medium and has a maximum absorption at 460 nm. The presence of uranium in hydrazine solution is not interfering in the analysis. Under optimum condition, the absorption intensity linearly increased with the concentration of hydrazine in the range from 0.05-10 mg?L–1 with a correlation coefficient of R2=0.9999 (n=7). The experimental detection limit is 0.05mgL–1. The sampling frequency is 15 samples h–1 and the relative standard deviation was 2.1% for 0.05 mg?L–1. This method is suitable for automatic and continuous analysis and successfully applied to determine the concentration of hydrazine in the aqueous stream of nuclear fuel reProcessing.
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determination of ultra traces amount of uranium in raffinates of Purex Process by laser fluorimetry
Journal of Radioanalytical and Nuclear Chemistry, 2012Co-Authors: S Ganesh, Fahmida Khan, M K Ahmed, P Velavendan, N K Pandey, Kamachi U Mudali, S K PandeyAbstract:A simple and rapid, laser fluorimetric method for the determination of uranium concentration in raffinate stream of Purex Process during reProcessing of spent nuclear fuel has been developed. It works on the principle of detection of fluorescence of uranyl complex formed by using fluorescence enhancing reagent like sodium pyrophosphate. The uranium concentration was determined in the range of 0–40 ppb and detection limit of 0.2 ppb. The optimum time discrimination is obtained when the uranyl ion is complexed with sodium pyrophosphate. Need of preconcentration step or separation of uranium from interfering elements is not an essential step.
David Woodhead - One of the best experts on this subject based on the ideXlab platform.
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Application of SAFT-VRE in the Flowsheet Simulation of an Advanced Purex Process
Industrial & Engineering Chemistry Research, 2019Co-Authors: Hongyan Chen, Megan Jobson, Andrew J Masters, Robin J Taylor, David WoodheadAbstract:SAFT-VRE is an extension of the statistical associating fluid theory for potentials of variable range (SAFT-VR) that can be used to describe the thermodynamic properties of strong-electrolyte solutions. Here the SAFT-VRE method is used in a flowsheet simulation code to calculate the densities of uranyl nitrate/nitric acid aqueous solutions and the activities of species that are needed for the calculation of the distribution coefficients of nitric acid and uranium in the nitric acid–30% tributyl phosphate (TBP) extraction system that is used in an advanced Purex Process for the reProcessing of spent nuclear fuels. Simulation results of both single-stage extraction experiments and of a multistage flowsheet test showed that the SAFT-VRE method can be used in flowsheet simulations with reasonable accuracy, thus demonstrating for the first time that thermodynamically-based mesoscale SAFT-VRE models can be linked to the macro-scale model of the Purex solvent extraction Process.
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simulation of neptunium extraction in an advanced Purex Process model improvement
Solvent Extraction and Ion Exchange, 2017Co-Authors: Hongyan Chen, Megan Jobson, Andrew J Masters, Robin J Taylor, David Woodhead, Colin Boxall, Scott EdwardsAbstract:ABSTRACTRouting neptunium to a single product in spent nuclear fuel reProcessing is a significant challenge. In this work, we have further improved the simulation of neptunium extraction in an advanced Purex flowsheet by applying a revised model of the Np(V)–Np(VI) redox reaction kinetics, a new nitric acid radiolysis model, and by evaluating various models for the nitrous acid distribution coefficient. The Np disproportionation reaction is shown to have a negligible effect. The models are validated against published ‘cold test’ experimental results; the ‘hot test’ simulation suggests that high neptunium radiolysis could help to achieve high recoveries using this flowsheet.
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development and validation of a flowsheet simulation model for neptunium extraction in an advanced Purex Process
Solvent Extraction and Ion Exchange, 2016Co-Authors: Hongyan Chen, Megan Jobson, Robin J Taylor, David Woodhead, Andrew J MastersAbstract:ABSTRACTAn Advanced Purex Process has been developed for separation and recycling of neptunium from spent nuclear fuel. This work presents a new flowsheet simulation model for the extraction of neptunium using centrifugal contactors, where mass transfer is modeled using two-film theory and a linear driving force. Distribution coefficients and neptunium redox reactions are modeled using published models. Mass transfer between the organic and aqueous phases in the phase separation zone is shown to have a negligible effect. The model is applied to a previously tested flowsheet and its predictions are shown to be in good agreement with the experimental results.
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recent developments in the Purex Process for nuclear fuel reProcessing complexant based stripping for uranium plutonium separation
Chimia, 2005Co-Authors: Eddie J Birkett, Michael Carrott, Danny O Fox, Christopher J Jones, Chris Maher, Cecile V Roube, Robin Taylor, David WoodheadAbstract:In order to recycle potentially valuable uranium and plutonium, the Purex Process has been successfully used to reProcess spent nuclear fuel for several decades now at industrial scales. The Process has developed over this period to treat higher burnup fuels, oxide as well as metal fuels within fewer solvent extraction cycles with reduced waste arisings. Within the context of advanced fuel cycle scenarios, there has been renewed international interest recently in separation technologies for recovering actinides from spent fuel. Aqueous fuel Processing research and development has included further enhancement of the Purex Process as well as the development of minor actinide partitioning technologies that use new extractants. The use of single cycle Purex solvent extraction flowsheets and centrifugal contactors are key objectives in the development of such advanced Purex Processes in future closed fuel cycles. These advances lead to intensified Processes, reducing the costs of plants and the volumes of wastes arising. By adopting other flowsheet changes, such as reduced fission product decontamination factors, U/Pu co-Processing and Pu/Np co-stripping, further improvements can be made addressing issues such as proliferation resistance and minor actinide burning, without adverse effects on the products. One interesting development is the demonstration that simple hydroxamic acid complexants can very effectively separate U from Np and Pu in such advanced Purex flowsheets.
Shinichi Aose - One of the best experts on this subject based on the ideXlab platform.
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conceptual design study on advanced aqueous reProcessing system for fast reactor fuel cycle
Journal of Nuclear Science and Technology, 2004Co-Authors: Takeshi Takata, Yoshikazu Koma, Koji Sato, Masayoshi Kamiya, Atsuhiro Shibata, Kazunori Nomura, Hideki Ogino, Tomozo Koyama, Shinichi AoseAbstract:The design study of an aqueous reProcessing system has been progressed for the feasibility study on commercialized fast reactor cycle systems in Japan. A simplified Purex Process, with the addition of a uranium crystallization step and a minor actinide (MA) recovery Process was proposed as the NEXT Process. For the simplified Purex Process and the SETFICS/TRUEX Process for MA recovery, small-scale hot tests were conducted. The test results showed the decontamination factor (DF) of 105 for the U, Pu and Np product was a reasonable target for the simplified Purex Process. Concerning to the Am and Cm product, the DF of rare earth elements (REs) and the other fission products (FPs) were over 10 and over 102, respectively. For economical competitiveness, an aqueous reProcessing plant with the capacity of 200 tHM/yr cleared the target cost, while the plant with the capacity of 50tHM/yr did not clear it. For efficient utilization of resources, the calculated U and TRU recovery of both plants was over 99wt%. Usin...
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conceptual design study on advanced aqueous reProcessing system for fast reactor fuel cycle
International Conference on Nuclear Engineering, 2004Co-Authors: Takeshi Takata, Yoshikazu Koma, Koji Sato, Masayoshi Kamiya, Atsuhiro Shibata, Kazunori Nomura, Hideki Ogino, Tomozo Koyama, Shinichi AoseAbstract:The design study of an aqueous reProcessing system has been progressed for the feasibility study on commercialized fast reactor cycle systems in Japan. A simplified Purex Process, with the addition of a uranium crystallization step and a minor actinide (MA) recovery Process was proposed as the NEXT Process. For the simplified Purex Process and the SETFICS/TRUEX Process for MA recovery, small-scale hot tests were conducted. The test results showed the decontamination factor (DF) of 10 5 for the U, Pu and Np product was a reasonable target for the simplified Purex Process. Concerning to the Am and Cm product, the DF of rare earth elements (REs) and the other fission products (FPs) were over 10 and over 10 2 , respectively. For economical competitiveness, an aqueous reProcessing plant with the capacity of 200tHM/yr cleared the target cost, while the plant with the capacity of 50 tHM/yr did not clear it. For efficient utilization of resources, the calculated U and TRU recovery of both plants was over 99 wt%. Using salt-free reagents and repeating concentration of waste solutions, the environmental impact was reduced. Pu was not isolated through the whole Processes in the NEXT Process in order to keep the proliferation resistance.
R. Natarajan - One of the best experts on this subject based on the ideXlab platform.
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reProcessing of spent fast reactor nuclear fuels
Reprocessing and Recycling of Spent Nuclear Fuel, 2015Co-Authors: R. NatarajanAbstract:Abstract Fast reactor (FR) fuel reProcessing, due to the higher concentration of plutonium and fission products, throws up enormous challenges in terms of designing and operating facilities safely. Though it is possible to reProcess FR spent fuel using nonaqueous Processes, it will be advantageous to deploy the time-tested Purex Process, if it can be adapted, as vast experience is available in the operation of such facilities. Various challenges in adapting the Purex Process for the high plutonium concentration bearing FR spent fuels are described. These are challenges that have been identified and addressed in the operation of the CORAL facility that has been successfully operating in India over the past 10 years. Substantial experience has been accumulated with reProcessing spent fuel from the fast breeder test reactor in this facility. An account of the current status of FR fuel reProcessing in India and other countries is given in this chapter.
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comparative studies on the determination of di n butyl phosphate in degraded solvent of Purex Process by ion chromatography and gas chromatography methods
Desalination and Water Treatment, 2012Co-Authors: P Velavendan, S Ganesh, N K Pandey, Kamachi U Mudali, R. NatarajanAbstract:Abstract This paper describes comparative studies on the determination of di-n-butyl phosphate (DBP) by ion chromatography (IC) and gas chromatography (GC) techniques in spent solvent of Purex Process used for the reProcessing of spent nuclear fuels. The ion chromatography method involves the separation of DBP from 30% TBP–NPH (tri-n-butylphosphate diluted in normal paraffin hydrocarbon) containing heavy metal ion like uranium and nitric acid by extraction of DBP into alkaline medium. DBP was subsequently eluted by ion-exchange separation in ion chromatography column and followed by suppressed conductivity detection. DBP is quantified to a lower limit of about 1 ppm with 3% RSD. However, in order to determine DBP by gas chromatography technique DBP is first quantitatively converted into its volatile and stable derivatives by using diazomethane prior to analysis by GC. Results obtained with ion chromatographic technique are compared with those of obtained by standard gas chromatographic technique. It was o...
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flow injection analysis of hydrazine in the aqueous streams of Purex Process by liquid chromatography system coupled with uv visible detector
Journal of Analytical Sciences Methods and Instrumentation, 2012Co-Authors: P Velavendan, N K Pandey, Kamachi U Mudali, N Pandey K S Ganesh, R. NatarajanAbstract:Present study describes the development of a rapid, sensitive and selective flow injection analysis of hydrazine in the aqueous streams of Purex Process by liquid chromatography system coupled with UV-Visible detector. The method is based on the formation of yellow coloured azine complex by reaction of hydrazine with para-dimethy laminobenzaldehyde (pDMAB). The formed yellow coloured complex is stable in acidic medium and has a maximum absorption at 460 nm. The presence of uranium in hydrazine solution is not interfering in the analysis. Under optimum condition, the absorption intensity linearly increased with the concentration of hydrazine in the range from 0.05-10 mg?L–1 with a correlation coefficient of R2=0.9999 (n=7). The experimental detection limit is 0.05mgL–1. The sampling frequency is 15 samples h–1 and the relative standard deviation was 2.1% for 0.05 mg?L–1. This method is suitable for automatic and continuous analysis and successfully applied to determine the concentration of hydrazine in the aqueous stream of nuclear fuel reProcessing.
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spectrophotometric determination of dissolved tri n butyl phosphate in aqueous streams of Purex Process
Journal of Radioanalytical and Nuclear Chemistry, 2012Co-Authors: S Ganesh, M K Ahmed, P Velavendan, N K Pandey, Kamachi U Mudali, R. NatarajanAbstract:A spectrophotometric method is developed for the determination of dissolved tri-n butyl phosphate (TBP) in aqueous streams of Purex Process used in nuclear fuel reProcessing. The method is based on the formation of phosphomolybdate with added ammonium molybdate followed by reduction with hydrazine sulphate in acid medium. Orthophosphate and molybdate ions combine in acidic solution to give molybdophosphoric (phosphomolybdic) acid, which upon selective reduction (with hydrazinium sulphate) produces a blue colour, due to molybdenum blue. The intensity of blue colour is proportional to the amount of phosphate. If the acidity at the time of reduction is 0.5 M in sulphuric acid and hydrazinium sulphate is the reductant, the resulting blue complex exhibits maximum absorption at 810–840 nm. The system obeys Lambert–Beer’s law at 830 nm in the concentration range of 0.1–1.0 μg/ml of phosphate. Molar Absorptivity was determined to be 3.1 × 104 L mol−1 cm−1 at 830 nm. The results obtained are reproducible with standard deviation of 1 % and relative error less than 2 % and are in good agreement with those obtained by ion chromatographic technique.
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Process modeling of in situ electrochemical partitioning of uranium and plutonium in Purex Process benchmark results with uranium reduction
Desalination and Water Treatment, 2012Co-Authors: V Reshmi, S Ganesh, N K Pandey, Kamachi U Mudali, Mushtaq Ahmed, R Sivasubramanian, R. NatarajanAbstract:Abstract In-situ reduction of plutonium and uranium for the separation of U/Pu is suitable for plutonium-rich fuels such as FBR fuels. The mathematical basis for a computer program PUSEP (Plutonium Uranium Solvent Extraction Program) for the analysis of partitioning cycle of Purex Process involving in-situ electrochemical reduction of uranium and plutonium is described in the present investigation. Model equations have been developed on the basis of the idealized model for mixer settlers incorporating distribution coefficients and redox reactions of the species involved and solved numerically to obtain concentration profiles of components. The validity of the model equations and associated computer program is tested by carrying out experiments in a proto type 20-stage electrolytic ejector mixer-settler operating without diaphragm for the electro reduction of uranium. The stage-wise experimental concentration profiles of U(VI), U(IV) and nitric acid were obtained and compared with the theoretical predictio...
