Search Results

You are looking at 11 - 17 of 17 items for

  • Author or Editor: U. Kamachi Mudali x
  • Refine by Access: All Content x
Clear All Modify Search
Journal of Radioanalytical and Nuclear Chemistry
Authors:
Deepak Dicholkar
,
Lalit Patil
,
Vilas Gaikar
,
Shekhar Kumar
,
U. Kamachi Mudali
, and
R. Natarajan

Abstract  

The high performance liquid chromatography and gas chromatography methods were investigated for their applicability in determining micro-level concentrations of tri-n-butyl phosphate (TBP). A high performance liquid chromatograph (HPLC) equipped with refractive index detector was used in determining TBP up to 2 ppm concentration level in the aqueous nitric acid solutions. The gas chromatography incorporated with Thermionic Detector (NPD) and Flame Photometric Detector (FPD) were examined for their potential in analyzing TBP in organic phase up to sub-ppm level. The results indicated that HPLC-RI technique is well suited for direct analysis of aqueous phase. For organic phase analysis, gas chromatographic methods with the TID and FPD were suitable but performance of detectors deteriorated often due to fouling.

Restricted access
Journal of Radioanalytical and Nuclear Chemistry
Authors:
Shekhar Kumar
,
Bijendra Kumar
,
M. Sampath
,
D. Sivakumar
,
U. Kamachi Mudali
, and
R. Natarajan

Abstract  

Nuclear solvent extraction was traditionally performed with packed columns, pulse columns, mixer-settlers and centrifugal extractors. However for rapid separations at micro-flow level, micro mixer-settlers were desired and in the past, few of them were actually designed and operated in nuclear solvent extraction research. In the current era of micro-reactor and microchannel devices, there is a renewed interest for micro-mixer-settlers for costly solvents and specialty solutes where small flow-rate is not an issue. In this article, development of a simple but effective micro-mixer-settler for nuclear solvent extraction is reported. The developed unit was tested with 30% TBP/n-dodecane/nitric acid system and in both the regimes of mass transfer c → d (mass transfer from continuous phase to dispersed phase, also written as c → d) and d → c (mass transfer from dispersed phase to continuous phase, also written as d → c) nearly 100% efficiency was observed in extraction as well as stripping modes of operation.

Restricted access

Abstract  

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 × 10L 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.

Restricted access

Abstract  

A simple, inexpensive PC based potentiometric titration technique for the assay of uranium using low volumes of sample aliquot (25–100 μL) along with all reagents (total volume of solution being less than 2.5 mL) is presented. The technique involves modification of the well known Davies and Gray Method recommended for assay of uranium(VI) in nuclear materials by introducing an innovative potentiometric titration device with a mini cell developed in-house. After appropriate chemical conditioning the titration is completed within a couple of minutes with display of online titration plot showing the progress of titration. The first derivative plot generated immediately after titration provides information of end point. The main advantage of using this technique is to carry out titration with minimum volumes of sample and reagents generating minimum volume of wastes after titration. The validity of the technique was evaluated using standard certified samples. This technique was applied for assay of uranium in a typical sample collected from fuel reprocessing laboratory. Further, the present technique was deployed in investigating the optimum conditions for efficient in situ production of U(IV). The precision in the estimation of uranium is highly satisfactory (RSD less than 1.0%).

Restricted access
Journal of Radioanalytical and Nuclear Chemistry
Authors:
Satyabrata Mishra
,
Falix Lawrence
,
R. Sreenivasan
,
N. Pandey
,
C. Mallika
,
S. Koganti
, and
U. Kamachi Mudali

Abstract  

Removal of nitric acid from high level liquid wastes (HLLW) of nuclear fuel reprocessing plants is warranted for simplifying the procedure for waste fixing. Chemical denitration aims to reduce the waste volume by destroying the acidity and subsequent concentration by adding suitable reductants. Reduction of nitric acid to gaseous products is an attractive way to accomplish denitration. Nitric acid reduction with formaldehyde proceeds with the formation of CO2, NO2, NO or N2O depending on the reaction conditions and all the reaction products except water can be eliminated from the system in gaseous form. The HNO3–HCHO reaction is governed by a complex mechanism of exhibiting relatively long induction period, depending upon the temperature, concentration of reactants and nitrous acid reaction intermediate. In the present work, a homogeneous denitration process with formaldehyde which offers safety and is governed by controlled kinetics was demonstrated on a laboratory scale. The induction period before commencement of the reaction was eliminated by maintaining the reaction mixture at a pre determined temperature of 98 °C. Based on the results accrued from lab scale experiments, the equipment for pilot plant scale operation was designed, the reaction efficiency for continuous denitration was determined and the investigation of nitric acid destruction was extended to full-scale plant capacity. The role of organics in the waste in foaming up of the reaction mixture was also studied using a synthetic waste solution.

Restricted access
Journal of Radioanalytical and Nuclear Chemistry
Authors:
Shekhar Kumar
,
S. Balasubramonian
,
Pranay Sinha
,
D. Sivakumar
,
U. Kamachi Mudali
, and
R. Natarajan

Abstract  

Thermophysical properties of reversed TALSPEAK extractant (0.3 M D2EPHA/0.2 M TBP/n-dodecane) were not available in literature. Authors have experimentally measured and correlated several thermophysical properties of RT solvent like density, viscosity, refractive index, acid uptake and flash point. In this paper, results of these studies will be discussed in detail.

Restricted access

Abstract  

Recently authors demonstrated direct dissolution of g-level PHWR UO2 fuel pellet fragments and in situ extraction by TBP-HNO3 and TiAP-HNO3 solutions at atmospheric pressures. Extending the work, similar studies were performed on intact unirradiated PHWR UO2 fuel pellets (~15 g U) with varying compositions of organic solvate of tri-n-butyl phosphate (TBP). It was observed that extent of dissolution was a strong function of organic solution composition TBP·(HNO3) x (H2O) y . Complete dissolution of intact UO2 pellet in a reasonable time was observed only in case of a particular solvate composition.

Restricted access