Shift in the maxima of a thermoluminescence (TL) glow peak, with an increase in heating rate, is a well-known and readily observed phenomena with its mathematical treatment/formalism in the paper from Randal and Wilkins in 1945. The occurrence of TL glow peak is observed at high temperature and earlier in time domain with increase in heating rate. The explanation of this phenomenon is not obvious at first instance and hasn't been addressed satisfactorily in the literature. The shift in TL peak with the heating rate has been ignored as far as its physical interpretation is concerned. This communication suggests an approach to explaining this observation, and an attempt has been made to associate a physical significance to it using the concept that evolves from black-body radiation, a well-known and thoroughly explored quantum phenomenon in modern physics.

Radioactive materials are used widely in various applications, which in turn have resulted in its large scale availability to the various end users. Though in India, there is strict regulatory control on obtaining radioactive material and their use, there have been reported cases of radioactive material detected as steel contaminant in the public domain elsewhere. This led to the analysis on estimation of the radiation field during the postulated event of radionuclides ⁶⁰ Co and ¹³⁷ Cs getting into the alloys of steel. The postulated cases of radioactive material getting mixed during the alloy making are studied for two case studies to ascertain the detection and capability of identifying the radioactive material even in trace levels, are presented in this paper. For ⁶⁰ Co and ¹³⁷ Cs nuclides - either by design or unintentionally, as low as 10 mg gaining entry into the matrix of 100 kg of the alloy during its making, it is estimated and shown that the radioactive materials can easily be detected. It is feasible due to the use of sensitive radiation monitors available at the Emergency Response Centers-Department of Atomic Energy, which are capable of detecting radiation field above the natural radiation background, the event can be detected, and the consequences can be minimized.

In mixed oxide (MOX) fuel fabrication facilities, the fabrication of MOX fuel involves various metallurgical operations such as mixing and milling of weighed quantities of uranium and plutonium oxides, precompaction, granulation of precompacts, final compaction of granules, sintering of green pellets, followed by fuel pin fabrication. All these operations are carried out in glove boxes, which have complex geometries due to housing of various equipments. Plutonium, a source of neutrons, is handled in large quantities in various forms such as powder, granules, pellets and pellets encapsulated in pins during these operations. It is obvious that extensive knowledge on the neutron spectral distributions in workplace is required from radiation protection and shielding point of view. In this paper, a brief introduction to the source of neutrons in MOX fuel handling facilities, studies that include experimental measurements of neutron spectra of various forms of MOX, contribution of neutron fluence in various energy groups and its dose equivalent, establishment of a simulation procedure for glove boxes handling Pu using FLUKA Monte Carlo codes, comparison of simulation results with the actual experimental measurements are presented. The results indicate that bare PuO ₂ and MOX sources require 3 mm of lead shield to eliminate gamma interference in MICROSPEC with N-probe neutron spectrometer. Furthermore, the studies reveal that dose equivalent contribution from MOX pellets is very significant in the energy group of 1.0-1.5 MeV unlike PuO ₂ powder and fuel pins, which exhibit significant contribution in the energy groups of both 1.0-1.5 MeV and 2.0-3.0 MeV. Also, the fuel pins show high neutron fluence in the energy group of 0.0-0.01 MeV, but they do not contribute significantly to dose equivalent. A good agreement between experimentally measured data and FLUKA simulated results has been observed.

Activated charcoal is a well-known adsorber of ²²² Rn and ²²⁰ Rn gases. This property can be effectively used for remediation of these gases in the workplaces of uranium and thorium processing facilities. However, the adsorption on charcoal is sensitive to variation in temperature and humidity. The successful designing and characterization of adsorption systems require an adequate understanding of these sensitivities. The study has been carried out towards this end, to delineate the effect of relative humidity on the efficacy of ²²⁰ Rn mitigations in a charcoal bed. Air carrying ²²⁰ Rn from a Pylon source was passed through a column filled with coconut shell-based granular activated charcoal. The relative humidity of the air was controlled, and the transmission characteristics were examined at relative humidity varying from 45% to 60%. The mitigation factor was found to decrease significantly with an increase of humidity in the air.

A silicon PIN diode-based electrostatic collection type online real-time instrument has been developed for simultaneous measurement of radon ( ²²² Rn) and thoron ( ²²⁰ Rn). The system, discussed in this paper, utilizes a hemispherical metal chamber (volume 1 L) for active air sampling. Estimation of ²²² Rn/ ²²⁰ Rn concentration is carried out through alpha spectroscopy of electro-deposited polonium atoms on the detector surface. The system description and the characterization studies carried out with this instrument are presented here. Its performance has been tested with reference equipments. The instrument showed sensitivity of 0.408 counts per hour (CPH)/(Bq/m ³ ) and 0.169 CPH/(Bq/m ³ ) for radon and thoron measurements, respectively, at an optimized collection voltage of + 1.6 kV and relative humidity <10%.

The prototype two element eye-lens dosimeter badge based on indigenously developed α-Al ₂ O ₃ :C optically stimulated luminescence dosimeter was investigated comprehensively for its suitability for eye-lens monitoring applications. The badge is calibrated to measure the eye-lens dose in terms of H _p (3). The minimum measurable dose using the eye-lens dosimeter badge is observed to be ~ 35 μSv. This prototype eye-lens dosimeter badge was found to be suitable for measuring doses from X-rays, beta and gamma radiations to the eye-lens. The satisfactory performance of the prototype two element eye-lens dosimeter badge along with its attractive features such as multiple readout, less processing time, very good beta response uniquely position it for monitoring the eye-lens dose are presented.

Strontium (Sr) is a by-product of the nuclear fission of uranium and plutonium in nuclear reactors and nuclear weapons. It is an important radionuclide in spent nuclear fuel and radioactive waste and is considered as one of the most hazardous constituents of nuclear wastes. Also during a nuclear/radiological emergency, workers as well as members of the public may get internally contaminated with Sr. Hence, a truly rapid radio-bioassay method will be needed to screen a large number of people for assessment of internal contamination and facilitate decision making on subsequent medical intervention. The current precipitation method used at Bioassay Lab., BARC, Trombay is quite lengthy and laborious. Efforts are being made to optimize bioassay methods at BARC using solid extraction chromatography (SEC) technique for emergency response. The present work reports standardization of SEC using Sr-Spec (Sr-Specific, make Eichrom Technologies, Darien, Illinois, USA) resin for rapid estimation of Sr in urine samples. Adsorbed Sr is eluted from the resin using 0.05M HNO ₃ , precipitated as carbonate and estimated using a low background gas flow Geiger-Müller counter (make ECIL, Hyderabad, India). The average recovery of stable Sr is ~80% and comparable to radioactive recovery observed in spiked samples. The minimum detectable activity (MDA) achieved by this technique is ~4 mBq/mL for ⁹⁰ Sr that is well below the required generic emergency action level of 19 Bq for 20 mL urine sample. The advantage of this method is its simplicity, fast turnaround time and an increase in the sample throughput.

Interventional radiology procedures are used for diagnosing certain medical conditions. The radiologists and medical professionals are exposed to ionizing radiation from X-rays of the equipments and also from scattered radiation during these procedures. The radiation exposure to the eye is more important to be assessed while performing such procedures. ICRP has revised the annual dose limit to the lens of the eye from 150 mSv to 20 mSv. In view of this revision, a study was carried out to evaluate the dose to the lens of the eye during interventional radiology. The paper gives the details of calibration of TLDs using a head phantom, predict annual equivalent dose and also highlight the dependence of dose on the position of TLD on the head. It is observed the predicted annual equivalent doses to the lens of eye are in the range of 25 mGy to 37 mGy. The selection of dosimeter placement may also result in an uncertainty of -14% to 20%.

A prototype two element eye lens dosimeter badge based on highly sensitive α-Al ₂ O ₃ :C optically stimulated luminescence dosimeters (OSLDs) was designed and developed. The badge consists of a plastic card in which two thin α-Al ₂ O ₃ :C optically stimulated luminescence (OSL) discs are placed. The OSLDs in the plastic card (when inserted into plastic cassette) are covered with energy compensation filters made up of 0.3 mm thick Cu and 1.35 mm thick Teflon discs positioned symmetrically on both sides of the dosimeter. The OSLD badge is useful for monitoring doses from photons and beta particles. In this paper, theoretical studies using Monte Carlo method as well as using the analytical technique have been performed to study the energy response of the bare α-Al ₂ O ₃ :C based disc dosimeters. These dosimeter discs have been found to exhibit over-response by a factor of ~ 3.4 at ~ 33 keV photon energy, whereas, beyond 80 keV photon energy, the response is nearly energy independent. Studies have also been performed to find the energy response of the α-Al ₂ O ₃ :C disc dosimeters under different metal filters, viz., Al, Cu, Sn etc., and under various thicknesses of Teflon. From theoretical simulations, it has been found that 0.3 mm thick Cu is sufficient to correct the over-response in lower energy region within acceptable limits. Further, Teflon disc (DuPont, USA) having thickness of 1.35 mm is found to be the optimized choice as filter for the second dosimeter disc. It is worth mentioning that the ratio of the response of the OSL disc under Teflon to that under Cu filter indicates average energy of X-ray photons and same is used to correct the over-response as well as to estimate the quantity H _p (3). Also for higher photon energy region, the readout of the dosimeter disc under Teflon filter directly measures the quantity H _p (3). Same holds true for beta particles having maximum beta energy, E_max beyond 0.7 MeV.