Depending on neutron energy, a number of personnel neutron dosimeters such as NTA film, Albedo TLD dosimeter and Nuclear Track Detectors (CR39) have been developed and reported so far. At present, CR-39 is in use as personnel neutron dosimeters for neutron monitoring of occupational workers in most of the nuclear installations due to non-availability high sensitive real time gamma insensitive personnel neutron dosimeters as per ICRP recommendations. The Superheated Emulsion Detector or bubble detector assess the magnitude of detrimental effects on health of person exposed to neutron radiation in terms of absorbed dose or equivalent dose even in presence of high gamma radiations flux in real time unlike nuclear track detectors. Bubble Technology Industries (BTI), Canada and Aphel Enterprises, USA have commercialized these type of bubble detectors for neutron and gamma measurements. The paper will elaborate the basic principle, special features, mechanism of bubble formation, national and international status including it's applications in accelerator Physics, Medical sciences, nuclear submarine and neutron /gamma dose measurements in nuclear reactors. These detectors may be most popular in due course among health physicists due to it's reusability and precise measurements for neutron measurements in reactor environment, pulse neutron measurements and personal neutron dosimetery.

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Real-time dose rate measurements along with the route followed by the radiation monitoring vehicle and the quick analysis of the data are of crucial importance during a nuclear or radiological emergency. To develop a timely response capability in different threat scenarios, such as the release of radioactive materials to the environment during any nuclear or radiological accident, Radiation Safety Systems Division, BARC has developed an advanced online radiation measurement cum vehicle tracking system for use. For the preparedness for response to any nuclear/radiological emergency scenario which may occur anywhere, the system designed is a global system for mobile (GSM) based radiation monitoring system (GRaMS) along with a global positioning system (GPS). It uses an energy compensated GM detector for radiation monitoring and is attached with commercially available GPS for online acquisition of positional coordinates with time, and GSM modem for online data transfer to a remote control centre. The equipment can be operated continuously while the vehicle is moving. The data downloaded and results plotted on GIS map helps in knowing the exact position of the vehicle along with the radiological status in terms of dose rates. This measurement information, either as raw data or results can be stored in the database. The system consumes ~250 mA including the GPS and GSM enabling thirty hours of continuous radioactivity monitoring with a 12 Ah battery source. The system has been used in road based environmental mobile radiation monitoring programme carried out at various parts of the country. With laptop support, the system maps the radiological status online onto the map of the area being surveyed, to help decision-making on countermeasures during the survey to enable the emergency managers to take appropriate decision.

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A study on the distribution and bioaccumulation of radionuclides has been undertaken in the abiotic and biotic matrices of the Nagapattinam coast (Bay of Bengal). The measurement of ambient gamma radiation levels in the coastal regions showed a rapidly changing non-uniform radiation regime ranging from 6.56 to 24.59 μR/hr. During the survey irregular and discontinuous radiation profile could be observed even in a narrow stretch of coastal area. The gross alpha and beta activities in sediments were indicative of the relative abundance of naturally occurring alpha and beta radionuclides. The mean value of gross alpha activity in the coastal area was 6.0 Bq/kg and gross beta activity in the coastal area was 16.8 Bq/kg. Gross alpha activity was non-uniformly distributed among the stations while beta activity was almost uniformly distributed. Among the several species of fishes taken for the study, Sardinella sp., accumulated higher level of ²¹⁰ Po in it's muscle (15.7 Bq/kg). The study on the distribution and bioaccumulation of alpha emitting radionuclides such as ²¹⁰ Po in the Nagapattinam, Velankanni and Karaikal coastal ecosystem, showed that the activity from this radionuclide is well within the safe limits, since there are no anthropogenic nuclear inputs.

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Dr. K.S. Parthasarathy is the recepient of IARP-Dr. A.K. Ganguly Memorial Oration Award-2010 This is the text of the Oration delivered at the 29th IARP National Conference (IARPNC-2010) on Recent Advances in Radiation Dosimetry, held during 3-5 February, 2010 at Mumbai.

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The radiation dose due to contamination can be avoided / minimized as a result of good practices, adherence to Radiation Protection Procedures and controlling by monitoring level of contamination of the personnel, in areas, on equipments and PPEs. Radioactive contamination on the materials, human body or other undesirable places is extremely harmful to the personnel at Nuclear Power Plants. Spread and cross contamination of radioactivity from the controlled areas is a very complicated problem for power reactor plant management. Protective cloths and PPEs used by workers during normal operation and maintenance as well as bieannual outage of the power plant are found contaminated by ⁶⁰ Co, ⁹⁰ Sr, ¹²⁴ Sb and ¹³⁷ Cs mainly. The contamination control and proper monitoring is a key of radiation protection to avoid superfluous local exposure which may result in local body radiation effects. A economical, simple to use and self contained new developed Laundry Monitor having 12 trays, each containing two energy compensated GM detectors with lead shielding has been installed at Kaiga Generating Station-3&4 with average monitor efficiency 0.23% (for ⁹⁰ Sr/ ⁹⁰ Yt plate source), LLD and MDA 7.97 cps and 34.65 Bq respectively. The operational experience of user-friendly Laundry Monitor provided very good results for contamination measurement and useful in controlling the spread of contamination.

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Gaseous effluent that is discharged through the stack of CORAL reprocessing facility is continuously monitored and the activity release estimated to comply with the regulatory requirements. Krypton-85, Iodine-131 and particulates are monitored and the activity release quantified in this facility. Release of Tritium in the gaseous effluent also needs to be monitored considering it's long half-life and the consequent environmental impact. Moreover, it has also become a regulatory requirement to monitor and quantify Tritium discharge from a reprocessing facility. In view of this, a prototype setup, adopting bubbler method, has been designed, installed and commissioned at CORAL to monitor and quantify the Tritium release during the recently concluded campaign. It was observed that Tritium was released during the chopping and dissolution processes, in measurable quantities.

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Assessment of internal exposure to the two acute exposure cases at waste management facilities have been carried out by whole body counting (in-vivo) and bio-assay (in-vitro) monitoring technique. These subjects were administered Ca-DTPA aerosol immediately after exposure. The initial whole body counting results showed internal contamination due to ¹³⁷ Cs. Urinary excretion rate of ²³⁹ Pu and ²⁴¹ Am in these cases are determined by means of standard radiochemical separation and estimation using ²³⁶ Pu and ²⁴³ Am tracers. The paper deals with the assessment of internal exposure to the radiation workers from waste management facility at Tarapur. It is observed that the assessed CED due to ¹³⁷ Cs was very low while the actinides namely, ²³⁹ Pu and ²⁴¹ Am, were the main contributors to the total CED. The CED for Case-I and Case-II was estimated as 8.54 mSv and 6.48 mSv respectively.

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Uranium mining is only one of it's kinds in the mining industry owing to associated inherent radiological hazards. Ore excavation processes may lead to release of radiologically significant materials into the surrounding environment. Such releases may lead to exposure of individuals during the course of operations near the source. The main features of radiological hazard associated in these are ²²² Rn and it's progeny, external gamma levels, long lived alpha activity in the natural uranium ore dust. The most significant internal hazards in an underground uranium mines surface facilities arises due to inhalation of short-lived decay products of radon ( ²²² Rn), which are daughter products of uranium ( ²³⁸ U). The present paper provides an estimate of inhalation dose to radiation workers engaged at the surface facilities in the vicinity of underground mines of Narwapahar. A radon gas monitor AlphaGuard PQ 2000 PRO (Genitron Instruments, Germany) was used for the measurement of outdoor atmospheric ²²² Rn concentrations in the vicinity of underground uranium mines surface facilities. Outdoor atmospheric ²²² Rn concentrations were found in the range of 10 to 87 Bq.m ^-3 with arithmetic mean of 34 Bq.m ^-3 . Average annual internal dose due to radon and it's progeny to the workers in these areas have been worked out to 0.17 mSv.y ^-1 .

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A major tool to study rates of sedimentation is ²¹⁰ Pb dating of sediment cores. In the present study, two core samples of 36cm long and 4.5 cm diameter were collected from Trombay Naval Jetty which is located at ~3 km away from BARC. ²¹⁰ Pb was estimated in the core fractions by ²¹⁰ Po which is in secular equilibrium with ²¹⁰ Pb. ²²⁶ Ra in the core fractions were estimated by High resolution 50% RE HP (Ge) Detector. The unsupported ²¹⁰ Pb was evaluated by subtracting ²²⁶ Ra from total ²¹⁰ Pb in each fraction. The log of unsupported ²¹⁰ Pb in each fraction was then plotted against depth and the slope of this line was evaluated. The sedimentation rate was obtained by dividing the ²¹⁰ Pb decay constant by the slope of the log-linear plot of unsupported ²¹⁰ Pb versus depth. The mean sedimentation rate thus calculated at Trombay Naval Jetty was 1.32 cmy ^-1 .

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In this paper, the radiological consequences of a catastrophic rupture of an HLW tank in a sub-vault of the AWTF have been studied. The postulated accident has been analysed assuming two phases, namely the accident phase and the recovery phase. Radiation dose to public through important pathways resulting from this postulated accident has been evaluated. Total effective whole body dose of 0.94 mSv has been estimated.

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For occupational workers handling protactinium-231 in nuclear facilities, there is a possibility of internal exposure to the radionuclide. Assessment of internal radiation dose due to intakes of ²³¹ Pa can be made from the activity concentration determined by the analysis of urine samples collected from the occupational workers. A study was initiated to standardize a simple and sensitive radiochemical procedure specific for the separation of Pa in bioassay samples. For this purpose, urine samples collected from members of public were spiked with Pa activity at mBq level. Pa was pre-concentrated on calcium phosphate, separated by ion-exchange technique and estimated using alpha spectrometry. The radiochemical recovery ranged from 90 % to 96 % with an average recovery of 94.1 ± 2.5%. The minimum concentration of ²³¹ Pa that can be analyzed by this method is about 0.3 mBq/L for one day counting time. This study will help in assessing the internal dose to workers handling protactinium.

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The radioactive liquid waste generated at PREFRE 3A, Kalpakkam will be transferred to the waste tank farm (WTF-P3A), for interim storage. The WTF vault and tanks are to be constructed fully underground, with only the auxiliary facilities located above ground. This paper presents the results of the shielding assessment carried out. Calculations have also been made for the decay heat source and hydrogen production rate in the waste solutions. It was concluded that, from bulk shielding point of view, the default shield thicknesses were adequate and that the HLW tanks are adequately cooled and vented.

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The indigenously developed Radiochromic film will provide visual assessment of absorbed radiation dose in cGy range. The film develops a distinctive and characteristic color upon exposure to ionizing radiation. When the active monomer polymer strip is exposed to ionizing radiation, a polymerization reaction immediately produces an intensely colored dye polymer that changes the appearance of the dosimeter. The amount of dye produced is proportional to the radiation dose. The dye polymer is cyan blue with a major absorbance peak at 676-680 nm and a minor peak at 625-630 nm. The developed film indicates progressively darker in proportion to absorbed dose in six different shades i.e. 0cGy, 10cGy, 25cGy, 50cGy, 125cGy and 350cGy. The progressively increase in darkness of same film is also observed by naked eye at 500cGy, 750cGy and 1000cGy.The dose response (10cGy to 1000cGy), dose rate response (10CGy/h- 1200cGy/h), sun light and room light effect on radiochromic film has also been studied. Sun light and room light effect on radiochromic film and color stability has also been studied. It is observed that the dosimeter is unaffected by exposures to sunlight or normal room light if radiation sensitive strip is covered by clear transparent UV protective film.

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This paper describes Health Physics experience during processing of Deeply Depleted Uranium (from PREFRE) at Uranium Metal Plant. It was observed that there is a significant difference in radiological status due to the handling of DDU oxide from Tarapur reprocessing plant. Radiation background during DDU handling was significantly higher as compared to natural uranium processing. Due to ²³² U, thoron daughter's concentration in working atmosphere was also observed which was found with in the apportionment provided for DDU handling.

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Occupational workers from Board of Radiation Isotopes and Technology (BRIT) are engaged in synthesis of various compounds of Sulphur-35 ( ³⁵ S) for application in life sciences. There is a possibility of intake of ³⁵ S by these workers during the course of their work. It is necessary to assess intake and assign internal dose to these workers following an inhalation of ³⁵ S. Estimation of intake of ³⁵ S, which is a pure beta emitter, involves analysis of urine samples collected from workers. The present study illustrates standardization of a radiochemical procedure for estimation of ³⁵ S in bioassay samples of these workers using Liquid Scintillation Spectrometer (LSS). The radiochemical recovery achieved in the present study ranged from 80.4% to 98.8 % with an average of about 91.4 %. The method, standardized, can be adopted for bioassay monitoring of workers handling ³⁵ S labeled compounds.

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Radioactive process liquid waste is generated during production of ^99m Tc gel generator at Radiopharmaceutical Laboratory, BRIT, Vashi. The wastes were transferred to Waste Management Division, BARC, in carbuoys after assessment of activity content, to ensure that the activity limit set by the regulatory authority is complied with. A study was taken up for the optimization of decay time of effluent generated during the production of Tc-Gel generator. Study on the identification of the radio nuclides present in the liquid waste, to plan for disposal is discussed in this paper. Liquid waste aliquots were counted by different techniques and were followed for decay of activity over long duration. The analysis of the samples led to identify the long lived radio nuclides like ⁶⁰ Co, ¹³⁴ Cs and ⁶⁵ Zn in liquid waste generated from ^99m Tc gel generator production facility.

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The use of extraction chromatography resin for the separation of Sr from Ca, Ba and Y in milk powder samples was studied. The effectiveness of chromatographic column in terms of it's repeated use and memory effect has been established. 39 milk powder samples were analyzed for separating Sr isotopes. The average recovery obtained with Sr-Spec resin was 79.4 with standard deviation of 5.6 and with nitrate separation method, the parameters were 68.2 and 9.4 respectively. Sr-Spec resin was found to be better than conventional nitrate method for parameters like radiochemical recovery and precision.

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High background radiation field exists in water treatment area (WTA) of rod cutting building (RCB) in Cirus due to beta, gamma contamination on it's floor. The high contamination on sides of wall and on floor is primarily due to deposition of activity generated during the regeneration of old mixed bed cartridges earlier (before year 1985) and presently due to deposition of contaminants by sump overflowing and wastes generated during maintenance/servicing of circulating pumps. RCB-WTA contribution to collective dose in present situation is up to 30% of the total collective dose of Cirus. Committee formed for developing methods and implementing measures to reduce high radiation background in RCB-WTA suggested tiling of the contaminated floor with prefabricated epoxy mixed lead shots, from ALARA point of view, after considering various options. These slabs of different thickness were tested for shielding effectiveness experimentally by using radiation source and theoretically using a code. Dose reduction factor of ~3.5 for a point source, obtained experimentally for epoxy mixed lead shots was very near to value obtained by theoretical simulation. An extended calculation for an area source using a code gives a higher dose rate reduction factor of ~28.

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