[en] The technique for sequential separation of plutonium and neptunium in the bioassay samples has been standardized. This was achieved by using suitable ion-exchange resins. The resins used were Dowex 50 X 8, H⁺ form, 100-200 mesh and Dowex 1 X 8, Cl^- form, 100-200 mesh. After separations, the plutonium and neptunium fractions were electrodeposited and estimated by alpha spectrometry. The analytical procedure developed can be applied for separation of plutonium and neptunium in various environmental and biological sample matrices. (author)

[en] Radiostrontium (Sr) is a by-product of the nuclear fission of uranium and plutonium in nuclear reactors and is an important radionuclide in spent nuclear fuel and radioactive waste. Rapid bioassay methods are required for estimating Sr in urine following internal contamination. Decision regarding medical intervention, if any can be based upon the results of urinalysis. The present method used at Bioassay Laboratory, Trombay is by Solid Extraction Chromatography (SEC) technique. The Sr separated from urine sample is precipitated as SrCO₃ and analyzed gravimetrically. However, gravimetric procedure is time consuming and therefore, in the present study, feasibility of Liquid Scintillation Counting for direct detection of radiostrontium in effluent was explored. The results obtained in the present study were compared with those obtained using gravimetric method

[en] A method for separation of ¹⁴C present as carbonates in bioassay samples of occupational radiation workers was developed by us earlier. This procedure was extended for analyzing organic carbon compounds in urine samples in addition to inorganic compounds. The average radiochemical recovery for both organic and inorganic carbon compounds was found to be comparable and ranged between 90-97%. (author)

[en] Internal Dosimetry Division of Bhabha Atomic Research Centre has applied neutron activation analysis technique for estimation of Th and U in various matrices and the same technique is being used on regular basis for bioassay monitoring of occupational workers. Radiochemical separations were performed, wherever required

[en] Radiation exposure, both external and internal, can occur to radiation workers during the operation of various nuclear fuel cycle facilities and radiation facilities. The assessment of radiation doses to workers, routinely or potentially exposed to radiation, through intake of radionuclide is an integral part of the radiation protection programme. Internal dose is the radiation exposure that results from the intake of radioactive materials into the body by inhalation, ingestion, absorption through the skin or via wounds. Assessment of radiation doses arising from the intake of radioactive material by the workers is termed as internal exposure assessment. Unlike external exposure, internal exposure cannot be measured directly. Its evaluation is based on the calculation of the intake of radionuclide either from direct measurements (e.g, external monitoring of whole body or of specific organs and tissues) or indirect measurements (e.g. radioactivity in urine, faeces, breath or samples from the working environment) (ICRP Pub. 78, 1997 and NRPB-W60, 2004). Another method of internal dose assessment is based on the measurement of airborne radionuclides in the working areas of the facility and the worker's occupancy in those areas

[en] Radiation exposure, both external and internal, may occur during the operation of various nuclear fuel cycle facilities and radiation facilities. Internal dose is the radiation exposure that results from the intake of radioactive materials into the body by inhalation, ingestion, absorption through the skin or via wounds. Assessment of internal doses can be divided into two phases, namely determination of the amount of radioactive material in the human body, in body organs or wounds by direct measurements and/ or by indirect methods like excretion analysis or air monitoring followed by interpretation of the monitored data in terms of intake and/or internal dose. Determination of internal doses is a complex procedure that requires the use of biokinetic and dosimetric models which describe the behavior of the radionuclides and the deposition of their energy in the tissues. In the present talk, ICRP models will be introduced along with the need for development of similar models for Indian population. Effective and prompt medical countermeasures for decorporation of actinides in case of radiological/nuclear emergencies, i.e. mass contamination scenarios Challenges faced in development of biokinetic models for Indians are discussed. Decorporation treatment is most effective when administered shortly after contamination before the radionuclides become fixed in tissues. Current decorporation treatment options especially for actinides do not meet the challenge imposed by a mass casualty setting. An overview of factors that affect the efficacy of decorporation treatment and recent developments in this field are also covered in the present talk. (author)

[en] Cerenkov counting is a non destructive method that does not require any scintillation cocktail in contrast to conventional liquid scintillation counting. Study was carried out to assess the effect of vial material and their sizes on Cerenkov efficiencies. Cerenkov efficiency depends on beta particle energy, sample volume and nature of dispersive medium. Recommendations on sample counting geometry for optimum Cerenkov efficiencies are presented in this paper

[en] Selection of liquid scintillation cocktail in case of Liquid Scintillation Counting (LSC) is based on its performance of certain aspects like sample load capacity and compatibility, counting efficiency, quench resistance, figure of merit (FOM), sample stability over long period of time etc. In the present study, quench resistance of few scintillation cocktails used in our laboratory and sample stability of ³H standards prepared in the same cocktails have been studied. LSC cocktails used in the present study were commercially available: Aqua light, Ultima Gold, Quick Safe 400 and Optiphase HiSafe III. All the samples were prepared in duplicate in 20 mL glass vials. Measurements were carried out using Packard TriCarb 2900TR LSC. The quench resistance of scintillator cocktails was assessed by determining the volume of quenching agent (nitromethane) needed to reduce the counting efficiency by a factor of two (the quench half volume, V1/2)

[en] One of the vital features of HIDEX 300 SL Liquid Scintillation Counter (LSC) is α/β discrimination. For efficient α/β separation, various counting parameters such as Pulse length Index (PLI), Alpha Delay Time (ADT) and Alpha Tail Offset (ATO), need to be optimized HIDEX LSC analyzes the pulse after 'alpha delay time' (ADT) so that it collects mainly the delayed component (Application note: HIDEX, 2012). The purpose of this study was to standardize the calibration procedure for simultaneous determination of α/β activities in a sample using HIDEX LSC. % deviation in activities was studied as a function of PLI, ADT and ATO. Effect of sample quenching on PLI settings was also investigated

[en] ³²P is preferentially eliminated from the body in urine and is estimated by in situ precipitation of ammonium molybdophosphate (AMP) in the urine followed by gross beta counting. The amount of AMP formed depends on the amount of stable phosphorus (P) present in the urine. Hence, the present study was undertaken to estimate daily urinary excretion of P by the spectrophotometry method. P forms a colorless complex (phosphomolybdate) with molybdic acid, which on reduction produces deep-blue-colored complex called molybdenum blue. The intensity of this blue color is directly proportional to the amount of P present in the sample. 24 h urine samples collected from radiation workers were analyzed for stable P, and its range was observed to be between 0.4 and 1.4 g/day. This information was valuable in finalizing volume of the urine sample required for analysis of ³²P in bioassay sample by gross beta counting technique. (author)