Radiation Protection and Environment

: 2023  |  Volume : 46  |  Issue : 5  |  Page : 230--299

Theme 5. Nuclear instrumentation and system development


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 Abstract - 51136: GEANT4 based Monte Carlo simulation including optical photon transport for a Gd3Ga3Al2O12:Ce single crystal scintillator

Pratip Mitra1, Mohit Tyagi2,3, R. G. Thomas3,4, A. Vinod Kumar1,3, S. C. Gadkari5

1Environmental Monitoring and Assessment Division, Bhabha Atomic Research Centre, 2Technical Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, 4Nuclear Physics Division, Bhabha Atomic Research Centre, 5Mahaavir Jyoti, Sector 10, Kharghar, Navi Mumbai, Maharashtra, India

E-mail: [email protected]

Single crystal Gd3Ga3Al2O12:Ce (GGAG:Ce) is a good scintillation material for gamma spectrometry, owing to its high density, high effective atomic number, high light yield, fast decay time, high detection efficiency, good energy resolution, non-hygroscopic nature, etc. In this paper, GEANT4-based Monte Carlo simulation has been carried out for a 2“ϕ×2“L cylindrical GGAG:Ce scintillator. Pulse-height spectrum is generated from coupled radiation transport and optical photon transport calculations. Experimentally recorded gamma spectrum has been compared with simulated spectrum. Experimental measurements were made with a 2“ϕ×2“L cylindrical GGAG:Ce single crystal scintillator (make Epic Crystal), optically coupled to a 2“ϕ Hamamatsu R1306 PMT. Output of the PMT was fed to a CAEN DT5790M dual digital pulse analyzer. High voltage of 1000 V was also provided to the PMT from DT5790M. Energy spectra were generated by digital pulse processing and acquired on a desktop via the data acquisition software CoMPASS. 137Cs test source was used for the experiments. During all the experiments, the source were placed at 300 mm from the detector's front surface and spectra were recorded for 100 s. The simulated assembly was modelled after the experimental one and is shown in [Figure 1]. The assembly was placed in air. A divergent beam of 662 keV gamma photons, originating from a point source placed on the axis of the cylinder at 300 mm from its front surface, directed towards the negative z-axis, and entirely covering the front surface of the assembly, was incident on it. The simulation methodology is similar to that reported earlier[1] for a CsI(Tl) scintillator, except for the values of the crystal specific parameters. Radiation transport was followed by generation of optical photons and their transport up to the scintillator-photodetector interface. The number of optical photons detected at this interface is equivalent to the number of photoelectrons produced in the photodetector. This number was scored on an event-by-event basis and then converted to pulse height. Simulations were performed for 7×104 primary gamma photons. A pulse-height tally was generated for each primary event. 'ROOT' was used to create a pulse-height histogram from this tally. The pulse-height scale was finally converted to energy scale. [Figure 2] presents the comparison of the simulated spectrum and the background subtracted experimental spectrum. Good agreement between the two can be observed. The slight mismatch of energy resolution between the two may be attributed to the fact that the simulated spectrum contains spectral broadening arising due to optical photon generation and transport statistics, whereas the experimental spectrum additionally contains contributions arising from gain fluctuations of the PMT, variation of response of the scintillator over its volume, electronic noise, drift in operating parameters over the course of measurement, etc. The overestimation in the Compton and backscatter regions of the experimental spectrum may be due to the contribution of the scattered radiation from the detector's ancillary equipment and surrounding materials in the laboratory, which could not be exactly modelled in the simulation. The validated simulation model would be useful for studying the effect of various scintillation, optical, physical and surface parameters of the detector on the resulting gamma spectra, a job which is otherwise difficult to carry out experimentally. The study would help in the design of radiation monitoring instruments.{Figure 1}{Figure 2}{Figure 3}{Figure 4}{Table 1}

Keywords: Energy resolution, gamma spectrometry, GEANT4, GGAG:Ce, optical photon transport


Mitra P, et al. IEEE Trans Nucl Sci 2019;66:1870-8.

 Abstract - 51142: Design and development of 85Kr monitor using in-house prepared plastic scintillator

K. C. Ajoy, R. Surya, P. Mahendran, N. Yuvaraj, A. Dhanasekaran, R. Santhanam, R. Mathiyarasu, D. Ponraju

Health Physics Section, Health and Industrial Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Release of 85Kr through the stack in CORAL reprocessing facility is monitored using a flow-through chamber co-axially fitted with an imported 180 x 25 mm cylindrical plastic scintillator (PS). In-house development of thin film plastic scintillatorhas been initiated by considering procurement and maintenance constraints in imported PS. Parametric studies such as bias voltage, chamber height, chamber dia and flow rate were carried out for optimization of parameters and effective operation of Kr monitor. The in-house prepared PS is optically coupled to a 2” PMT that serves as detector in the monitor. The detector's operating voltage is selected based on Minimum Detectable Activity (MDA), background and efficiency. The minimum acceptable detection efficiency is fixed at 1 % as the present system's diameter was selected using this criterion. A 137Cs beta source is chosen to determine the detector efficiency as its end point energy is very close to 85Kr. The 137Cs source is moved vertically to identify the distance at which the efficiency reaches 1%. A similar exercise is carried out by moving the source horizontally at a constant vertical distance. The sampling chamber is fabricated in SS316 and the detector is mounted. For the selected operating voltage, the volumetric efficiency is estimated by injecting 85Kr of known concentration. The leak rate of the chamber at atmospheric pressure is derived from the rate of decrease of count rate in the monitor. The variation in 85Kr concentration in the chamber as a function of time (t) at various sampling rates is modeled using the standard first-order differential equation. [Figure 1] shows the variation of efficiency and MDA values of the detector with operating voltages of PMT. The optimum bias voltage is found to be 850 volts. The volumetric efficiency of the system is 4.5 % and the leak rate is 0.054 h-1. [Figure 2] shows the modeled response of the detector for an input rate of 1 Bqs-1. It is observed that the saturated concentration is reached within 10 s for a constant input for the entire modeled flow rates. Thus, selecting counting time beyond 10 s, the monitor response will be independent of change in input concentration. [Table 1] shows optimized parameters of the present and proposed85Kr monitor. The Minimum Detectable Concentration (MDC) for the present and proposed monitor is same. Though the system is designed and developed for 85Kr monitoring, it can be used as a gross β- fission product noble radioactive gas monitor. The monitor can operate even at high85Kr concentration environments, as the PS produced pulse width is <10 μs.

Keywords: Counting system, fission product noble gas, plastic scintillator, pulse height spectrum


Baskar S, et al. Monte Carlo simulation of calibration factor of a plastic scintillator based 85Kr stack monitoring system. Proc Conf 29th IARP Nat Conf Recent Adv Radiat Dosimetry 2010.

 Abstract - 51149: Design and development of active interrogation systems for fissile material quantification

P. M. Dighe1,2, M. Vinod1, S. G. Thombare1, P. V. Bhatnagar1

1Electronics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Anushaktinagar, Mumbai, Maharashtra, India

E-mail: [email protected]

Fissile material assay is conventionally carried out by measuring natural radiations emitted from the isotopes. This method is called as passive technique. For most of the requirements, passive techniques are sufficiently accurate. However, in special applications where fissile material is heavily shielded or has to be detected in the presence of high gamma background, active interrogation method is employed. In active non-destructive analysis (ANDA) neutron or photon source is used to induce fission in the fissile material and high intensity neutron and gamma radiations produced after fission is measured and analysed for characterization. In all ANDA techniques measurements are based on after fission detection of prompt neutrons, delayed neutrons, or delayed photons in order to establish the quantity of fissile material present in the given container. Electronics Division has been developing photofission and active neutron-based detection systems for hull drum monitoring at fuel reprocessing labs. The systems comprise of Helium-3 neutron detectors, BGO scintillators, front end pulse processing electronics and PC based digital data acquisition (DAQ) system. Photofission method: Sets of photofission experiments[1] were conducted at the 10 MeV LINAC facility. A scaled down model of hull drum with 500 mm OD x 1000 mm length that can be rotated by 360° angle and fixed at a particular position by lock key has been fabricated for the experiments. The instrumentation setup consists of multiple Triple Channel HV NIM modules, Dual channel pre-amplifier cum spectroscopy amplifier NIM modules, Hex Channel USB MCS NIM modules and USB MCA modules as shown in [Figure 1]. With photofission delayed neutrons, one gram of fissile material could be quantified in a hull drum with an accuracy of ~ 5%. Active neutron method: Electronics Division in collaboration with PP&EMD and APPD is conducting active neutron experiments in Hall 4, BARC. The detection system is identical to that of the system used for photofission. In single pulse source active neutron experiments for achieving measurable quantification, it is proposed to measure the prompt fission neutrons instead of delayed neutrons. The fission neutrons are nearly two-decade higher intensity compared to delayed neutrons. The fission neutrons can be measured with Differential Die Away (DDA) technique. Main challenge of DDA technique is to subtract the source neutron signal and measure fission neutrons with micro second dwell time. To improve and rule out EMI/EMC effect and for improved performance, modifications have been carried out in the detector signal processing electronics. Enhanced detector mountable electronics with optical transmitter/receivers have been designed and lab tested. The system has 0.5μs shaping time and acquires counts with <0.1ms dwell time. Simulation studies for active neutron method: Theoretical estimations and simulations are carried out as shown in [Figure 2] to estimate the Helium-3 proportional counter response in the proposed active neutron experiments. The absolute efficiency of the neutron detection assembly (~10%) estimated from simulations and measured experimentally validated within 5% accuracy. The same simulation model utilized, estimated for1Kg of natU sample with 1x10^9 neutron per pulse of source can produce 9100 prompt fissions that decay with 0.1 ms dwell time for DDA. Summary and path forward: The photofission experiments successfully demonstrated the utilization of the active method for fissile material mass measurements in large drums. The portable active neutron system validation is in progress.{Figure 5}{Figure 6}

Keywords: Active neutron, digital data acquisition, neutron detectors, photofission


Dighe PM, Vinod M, Thombare SG, Prafulla S, Kamble NR, Kamble LP, et al. Nucl Instrum Methods Phys Res A 2019;946:162624.

 Abstract - 51171: A network based dual channel dual phosphor counting system

A. Dhanasekaran, K. C. Ajoy, R. Santhanam, R. Mathiyarasu, D. Ponraju, N. T. Bineesh1, K. P. Desheeb1, Amudhu Ramesh Kumar1, M. S. Gopikrishna1, Geo Mathews1

Health Physics Section, Health and Industrial Safety Division, Coimbatore, 1Radiation Instrumentation Section, Reprocessing Process and Maintenance Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

The role and responsibility of health physicists working in nuclear facilities are to establish a proper radiation protection program and develop the infrastructure required for its implementation. The operational health physics (OHP) community is expected to react swiftly to the situations and produce a reliable analysis outcome of the different samples acquired from the field. A radiation counting system is one of the instruments used by OHP for analysing the samples. A variety of microcontroller-based counting systems are commercially available. They lag in features such as in-built quality control check, categorized data storage, and results in radiation protection-related quantities. Developing a counting system operable through an Ethernet network would be a complete solution to these constraints posed. The paper briefs the design and development of a radiation counting system operable through a Local Area Network (LAN) with dual phosphor[1] as the detector for simultaneous measurement of α and β radioactivity present in the sample. The electronic architecture of the developed system is shown in [Figure 1]. The Ej-444 dual phosphor detector is coupled to a 2” Photo Multiplier Tube (PMT). A commercial high voltage module provides the necessary bias to the PMT tube. The signal from the PMT is fed to the in-house developed signal conditioning board (SCB), which amplifies, segregates and generates digital outputs for α and β,γ interactions occurring in the detector. The digital outputs from the SCB are fed to the Digital Input Output Module (DIO), operated in frequency measurement mode. DIO module transmits the measured frequency values through a standard Modbus RS485 on request from the server. A generic RS485 to Ethernet converter connects the system to the Local Area Network (LAN). The operating software of the system is developed in the visual studio platform. For various counting requirements, different input forms such as “background“, “calibration“, “air sample” and “liquid planchette” are provided to facilitate the counting with in-built quality control measures. In the background counting mode, the software automatically calculates acceptable upper and lower background limits based on the previous ten measurements average. The data is stored only when the current background measurement falls within it. The user can generate a calibration source database and during calibration of the system, the decay corrected activity value of the source will be used for efficiency calculation. The air samples to be counted may originate from Continuous Air Monitors, Spot air samplers, Personal Air samplers or effluent monitors. Depending upon the user input details, the software provides the results in radiation protection quantities such as DACh, DAC or Bq. The liquid planchette results are directly displayed in MBq/m3 along with the AERB Liquid waste category (AxBx).Each counting mode is linked to a unique excel database to store the results. The system is calibrated with standard α andβγ sources and has efficiency, Minimum Detectable Activity (MDA) values comparable to that of any commercially available dual phosphor systems [Table 1]. The α to β crosstalk of the system is 2-3 % while the β to α crosstalk is <0.01% which is well within the acceptable range. The system is made of commercially available modules of industrial standard which makes the troubleshooting and maintenance easier.{Figure 7}{Table 2}

Keywords: Counting system, crosstalk, dual phosphor, IO module, modbus, scintillator


Dhanasekaran A, et al. Twenty Ninth IARP National Conference on Recent Advances in Radiation Dosimetry. Proceedings Conference; 2010.

 Abstract - 51178: Y2O3 loaded plastic scinitllators for radiation detection

N. Yuvaraj, A. Dhanasekaran, K. C. Ajoy, R. Santhanam, R. Mathiyarasu, D. Ponraju

Health Physics Section, Health and Industrial Safety Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

A solution casting method is developed in-house for preparing thin-film plastic scintillators and the various parameters such as the composition of ingredients, mixing ratio, bias voltage and thickness required for β detection are optimized.[1] Various studies are carried out for improving the light output of PS and one among them is loading phosphors with good optical properties, into the PS medium. Yttrium compounds preferably doped with cerium have shown Radioluminescence with impressive luminescence decay time and low afterglow. Yttrium oxide is reported as a self-activated scintillator. It has many compatible properties with that of plastic scintillators, such as fast luminescence decay time (28 ns), emission wavelength (370 nm) and refractive index (1.93). Additional physical properties like high effective Z (36.7) and density (5.04 g/cm3) make it suitable for improving the gamma detection capability. Polystyrene beads and predetermined quantities of PPO and POPOP are dissolved p-xylene in standard nylon containers. Y2O3 is purchased from sigma-aldrich with purity >99.0% in pure anhydrous powder is weighed in a standard analytical balance (Saffron scales, SES205) with 100 μg accuracy and added to the solution. The solution is coated over transparency sheets made of cellulose acetate using a thin wire coater to get the desired thickness of ~ 250 μ. The prepared wet films were allowed to dry at room temperature and peeled off from the transparency sheet after drying. The composite scintillator sheet is then cut into six different circular pieces of area ~20 cm2 to study the uniformity of response among them. [Figure 1] shows the photograph of PS with no loading and Y2O3 loaded PS. The Y2O3 loaded PS is coupled with a photomultiplier tube (R878) using silicone based optical grease to study the pulse height spectrum (PHS). Two layers of Mylar films (0.3 mg/cm2) are used in the front face of the scintillator. The photomultiplier tube is connected to Digital Multi-Channel Analyzer (DMCA). Standard α, β, and γ sources with accuracy ranging from 3 – 10% are used to acquire PHS in conventional sample counting geometry. The PHS is acquired in such a way that the net counts are at least 100000 to achieve better counting statistics. [Figure 2] shows the 238Pu alpha spectrum acquired using the scintillators. The Y2O3 loaded PS spectrum is the average of counts obtained with six different pieces and its ± σ in the respective channel counts. The peak channel which is a representative of light output is higher for Y2O3 loaded PS than other scintillators. The quantitative comparison of average β and γ detection efficiencies of different scintillators is shown in [Table 1]. It is observable that Y2O3 loaded PS has 0.1 % as detection efficiency for 241Am gamma energy of 60 keV whereas other two scintillators showed no response (*NR). Similarly, for the 133Ba gamma source Y2O3 loaded PS showed higher efficiency because of 80 keV gamma energy. Thus Y2O3 loaded PS has comparable β efficiencies with an added response to low energy gammas (<100 keV).{Figure 8}{Figure 9}{Table 3}

Keywords: Counting system, phosphor, plastic scintillator, pulse height spectrum


Dhanasekaran A, et al. Proceedings Conferences IARP National Conference on Advances in Radiation Measurement Systems and Techniques; 2014. p. 91.

 Abstract - 51179: Composite thin-film plastic scintillators for thermal neutron detection

A. Dhanasekaran, N. Yuvaraj, K. C. Ajoy, Kalvala Rajakrishna1, M. T. Jose1

Health Physics Section, Health and Industrial Safety Division, Indira Gandhi Centre for Atomic Research, 1Homi Bhabha National Institute, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Plastic scintillators (PS) loaded with chemical compounds, or luminescent micro crystals are called composite scintillators. The composite scintillator's radiation detection properties can be tuned by loading chemical compounds. PS acts as a fast neutron detector due to its hydrogen content but lacks in sensitivity toward the thermal neutron detection. Thus, thermal neutron sensitivity can be increased by loading thermal neutron-sensitive chemical compounds to the PS. Gd2O2S: Tb, Gd2O3 and Gd(BO2)3:Tb are selected as suitable compounds for loading after analysing their physical properties and each chemical compound is loaded into the PS by 14.3 weight %. The prepared composite PS film thickness is measured to be in the range of 200 – 300μ. The composite scintillators are coupled to a 2” dia Photo Multiplier Tubes (PMT) with the front face covered with two layers of mylar with a thickness density of 0.3 mg/cm2. The detector system is plugged into a digital multi-channel analyser (DMCA) and Pulse Height Spectrum (PHS) is acquired by keeping a 37 GBq Am-Be neutron source at 15 cm distance. The high-density polyethylene (HDPE) block is used as a moderator and PHS is acquired by increasing thickness in steps of 2.5 cm. The Am-Be source is covered with a 1” thick lead block to reduce the gamma interference. The comparison of PHS acquired for 200 s with different PS without a moderator is shown in [Figure 1]. A typical fast neutron scattering spectrum is observed in all the scintillators and the shape of PHS is similar for all scintillators. The PHS of gadolinium-loaded PS showed a higher-end channel, indicating a slight increase in the light output. When a gadolinium atom absorbs a neutron, it emits internal conversion electrons and the emission probability is 0.61/absorption. 76% of the gadolinium isotopes' internal conversion electrons are emitted in the energy region of 60 –250 keV. Thus, a differential spectrum of moderated and unmoderated neutron spectrum is expected to show a peak in the lower energy region of the PHS.


The differential PHS of moderator thickness 7.5 cm and unmoderated for different PS is shown in [Figure 2]. The highest response is observed in the Gd2O3 loaded PS. [Figure 3] shows the cps/nv values normalized to unmoderated cps/nv values observed in the 3He detector (IAEA-TRS-403), PS loaded with Gd2O3 and unloaded PS. The response of the Gd2O3 loaded PS is similar to that of the 3He detector, i.e., the response increases with increasing moderator thickness of 7.5 cm and starts decreasing beyond that. Thus, the optimal moderator thickness for the configuration is 5 to 7.5 cm. The PS showed a slight increase in response at 2.5 cm moderator thickness and the response decreased continuously beyond that. Notably, β source responses of these composite scintillators have not degraded compared to unloaded PS. The in-house developed composite scintillator is expected to serve as a versatile detector that could be used as conventional PS for beta detection or neutron detectors when covered with a suitable thickness of lead and HDPE.{Figure 10}{Figure 11}{Figure 12}

Keywords: Composite scintillators, detectors, neutron, plastic scintillator, pulse height spectrum

 Abstract - 51205: Development of on-line 129I stack monitoring system using NaI(Tl) detector

Pankaj Kumar1, Anish A. Ansari1, D. K. Pandey1, J. P. N. Pandey1, G. Ganesh1, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Bhabha Atomic Research Centre, Mumbai, Maharashtra India

E-mail: [email protected]

Introduction: Iodine-129, half-life 15.7×106 years, is formed in Spent Nuclear Fuel (SNF) as a result of nuclear fission reaction at nuclear reactor with fission yield 0.658 % (PHWR) and remains in the fuel matrix till its reprocessing. It undergoes beta decay with 154.4 keV to unstable 129Xe, followed by 39.6 KeV(7.52%) low energy gamma along with L-X-rays (29 & 33 KeV). During reprocessing of SNF about 95 to 99% of Iodine is released into the environment in vapor or gaseous form through the tall stack after proper treatment and filtration.[1] As a part of regulatory requirement, continuous monitoring of 129I & other radio-nuclides required to be carried out in stack discharges. However, due to non-availability of on-line monitoring system, off-line monitoring of 129I at reprocessing plant is being carried out by sampling of stack air using standard activated charcoal cartridge (>99% collection efficiency) and activity is estimated by analysing the cartridges with the help of calibrated HPGe spectrometer. Existing off-line monitoring system is upgraded by indigenously developed on-line monitor, using 3“x3” NaI(Tl) detector coupled to Single Channel Analyser (SCA), which is in line with International standard IEC 60761-1 & IEC 60761-4. After development, the monitor is successfully installed in stack effluent sampling circuit at a reprocessing plant and is being in use for on-line monitoring of 129I.

Materials and Methods: Development of sampling Head for on-line monitoring: A sampling head is indigenously designed, developed and fabricated which consists of a holders for housing iodine cartridge, 3“x3” NaI(Tl) detector facing the charcoal cartridge, air inlet & outlet tubing for stack air sampling and spring action paddle for easy replacement of cartridge. One mm thick Teflon sheet is fitted in leak tight manner over the sampling cavity as shown in [Figure 1]. The distance between detector & cartridge is optimized for maximum counting efficiency and background is minimized by shielding the detector & optimization of LLD & ULD of SCA. Efficiency Calibration of Monitor: A secondary standard source of charcoal cartridge having 20.3 KBq 129I activity was prepared using a calibrated ORTEC make N-type HPGe spectrometer.[2] After energy calibration, absolute efficiency of on-line monitoring system was determined at various distance from detector as shown in [Figure 2]. Optimized efficiency (ƞ) of 4.2% at a distance of 1.5 cm with calibration factor 24 Bq/cps is established. Measurement of 129I activity by on-line monitor: After successful installation of online monitor, all sampled cartridges were analyzed for 129I activity and measured values were compared with off-line observations. Results of estimated activity for 5 such samples are presented in [Table 1]. MDA for on-line 129I monitoring system = 153 ± 12.36 Bq & MDA for off-line measurement = 8 ± 1.21 Bq.

Results and Discussion: Activity of 129I estimated by on-line monitoring system was found in good agreement with off-line measurement with variation ±10%. In addition to that 129I on-line monitor real time data can be used as an indicator for complete dissolution of spent fuel matrix as shown in [Figure 3]. Thus, indigenously developed NaI(Tl) detector based on-line monitoring system is capable of continuous monitoring of 129I in stack effluents of reprocessing facilities.{Figure 13}{Figure 14}{Figure 15}{Table 4}

Keywords: Efficiency calibration, Iodine-129, on-line monitoring, reprocessing plant


Sakurai T, et al. Trapping and measuring 129I in cartridge filters. J Nucl Sci Technol 1997;34:211-6.Chinnaesakki S, et al. Low energy gamma spectrometry for 129I measurement in charcoal. IARP 2020;200.

 Abstract - 51206: Development of a multi-shell moderating assembly for neutron spectrometry in particle accelerators: Generation of response matrix with Au foils as detectors

S. P. Tripathy1,2, G. S. Sahoo1, Sabyasachi Paul1, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Neutron spectrometry in accelerator radiation environment is complicated due to the presence of intense gamma field, RF interference with the electronics involved in active measurement techniques, space and irradiation time limitations, etc.[1] In this regard, a single sphere moderating assembly has been designed and fabricated, in which the passive thermal neutron detectors can be used. In this work, the use of Au foils in the moderating assembly has been explored for neutron spectrometry. [Figure 1] presents the cross-sectional view of the spectrometer design having 11 HDPE (high density polyethylene) hemispheres with diameters from 2” to 12“. The Au foils can be attached to the upper surface of each hemisphere, as shown in [Figure 1]. The complete single sphere neutron spectrometer (SSNS) is shown in [Figure 2]. The system can measure the neutrons from thermal to ~100 MeV. The response matrix for the system has been generated using FLUKA Monte Carlo simulation code (Version 2020.0)[2] as explained below. Parallel beams of mono-energetic neutrons from 0.01 eV to 100 MeV, in 65 energy intervals, were transported through the SSNS having 12 Au foils (1” dia, 0.2 mm thick). Each simulation was carried out for 2.5x108 particle histories to retain the statistical uncertainty below 2%. As per the irradiation geometry in this case, the foil placed on the surface of 12” sphere is the bare detector (without any moderation), and the foil at the centre is at a moderation depth of 6“, corresponding to the sphere with diameter of 12“. An USRBIN estimator with region binning was used to score the photon fluence inside these Au foils which are produced by 197Au(n,γ) reaction. The low energy neutron cross section of Au was selected using the LOW-MAT card and the transport of photons was activated by the use of EMF card. The peaks observed in the response matrix [Figure 3] are due to the resonances in the cross-sections at lower energies. The net activities induced in the Au foils (at different moderation depths) and the corresponding response functions need to be solved mathematically (called “unfolding / deconvolution“) to obtain the neutron spectra. This in-house built SSNS is a portable system (~10 kg) and found to be convenient for neutron spectrometry in accelerator radiation environment. Further work is in progress for real field measurements.{Figure 16}{Figure 17}{Figure 18}

Keywords: Au foils, response matrix, single sphere neutron spectrometer


Tripathy S. Sol State Phenomena 2015;238:1-15.Ferrari A, et al. CERN-2005-10, INFN/TC_05/11, SLAC-R-773, CERN, Switzerland; 2005.

 Abstract - 51256: Discrimination of low energy radiations in presence of high energies using phoswich detector of single crystal scintillators

Sonu1,2, M. Sonawane, Mohit Tyagi1,2

1Technical Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

In various applications such as medical diagnostics, academics and in nuclear industries the radiation detectors have to be used in mixed field of radiations. Depending on the size of detector and intensities of different energies radiation, the pulse height spectra may not be able to discriminate low energies in presence of high energies due to the continuum of Compton scattering and backscatter from surrounding materials etc. Gas or thin film based phoswich detectors are used to discriminate low energies but have disadvantage of the poor efficiencies. Single crystals have advantage of higher efficiency in compact size because of the higher atomic density. This work shows the Pulse Shape Discrimination (PSD) used in phoswich detector of Gd3Ga3Al2O12: Ce (GGAG) and CsI:Tl single crystals to discriminate the low energy gamma from high energy gamma.[1] The phoswich detector was fabricated using processed samples of two single crystals of GGAG (1 mm × 2 inch) and CsI (2 inch × 2 inch). Single crystal of GGAG: Ce was grown using Czochralski technique and CsI: Tl single crystal was grown using 4 zones Bridgeman method with translation rate of 1 mm/hr. The low energy gamma may stopped in GGAG depending on its thickness and only high energy gamma pass through GGAG and get deposited in CsI. Because of large difference in decay time of these crystals, interaction occurring in front and back crystals are well separated using PSD with high figure of merit. To avoid any entrance of low energies from side surfaces of CsI the side surface of the detector was covered with 1 mm thick lead. [Figure 1a] shows the pulse shape discrimination between low energy and high energy gamma in phoswich detector combination of CsI and GGAG from Cs137 (10kBq) and Eu152 (10kBq) sources. The upper band shows the energy deposited in CsI and lower band shows the energy deposited in GGAG. [Figure 1]b shows the projection of upper band (CsI) on x-axis in which higher energy gamma from Cs137 and Eu152 sources are deposited. [Figure 1]c shows the projection of lower band (GGAG) on x-axis, which shows that the low energy gamma 32keV from Cs137 source, 80keV and 122keV from Eu152 source are completely deposited in the front GGAG crystal, which otherwise suppressed in compton and backscatter peak energies.{Figure 19}

Keywords: Phoswich, pulse shape discrimination, scintillation, single crystal


We are thankful to Mr. Swapnil for crucible sealing, and Dr. Shashwati Sen, Head, CTS and Dr. L. M. Pant, Head ,TPD for providing their support.


Tyagi M, Rawat S, Gourishetty AK, Gadkari SC. Nucl Instrum Methods Phys Res A 2020;951:162982.

 Abstract - 51257: Development and performance of a MAPMT-based neutron detector system using waveform digitizer

Binqing Zhao, Yifan Zhang, Longfei Chen, Wanting Gui, Cailin Wang, Jinkui Zhao

Songshan Lake Materials Laboratory, Dongguan City, Guangdong Province, China

E-mail: [email protected]

In order to study the performance of the neutron sensitive scintillator,[1] a Multi-anode Photomultiplier Tube (MAPMT) based neutron detector system using waveform digitizer is developed in this paper. The schematic diagram of the neutron detector system is shown in [Figure 1]. The neutron Anger detector system uses GS20 lithium glass with a thickness of 1.5 mm and an effective area of 51.8 mm × 51.8 mm. The top and end faces of GS20 scintillator are polished. A 1.5 mm light guide is added between the scintillator and the photodetector, and the light guide material is acrylic (PMMA). The photodetector is a Hamamatsu H12700 series MAPMT. The array size is 8×8, each pixel size is 6 mm × 6 mm, and the total effective detection area is 48.5 mm × 48.5 mm. The scintillator, the light guide and the photodetector are all coupled by optical silicone oil. The 64-channel energy signal of the detector is input into two waveform digitizers through the transimpedance amplifier module, and the DY12-channel signal generated by the multi-anode photomultiplier tube is amplified and then outputs the trigger signal as the fast trigger signal of the waveform digitizer after passing through the constant fraction discriminator (CFD) module. Finally, the input pulses in the 64-channel waveform digitizer is sampled and digitized, and then stored into 64 txt format data files. The neutron incident position is calculated by the center of gravity method.

Energy Spectrum: As shown in [Figure 2], the number of events below a threshold value is counted to ensure less background events and more neutron events. In this paper, the trigger threshold value of CFD is set at -200mV, and gamma background events are automatically filtered in data acquisition.

Two-dimensional Imaging: In order to further analyze the two-dimensional imaging performance of the detector system, a B4C mask with two square holes is designed and manufactured, which contains 30% boron. The size of each square hole is 2.5 mm×4 mm, and the edge spacing of the square holes is 2.5 mm. The neutron image obtained by the detector is shown in [Figure 3]. Two squares can be clearly displayed in the image, and oblique long stripes appear in four corners of the image, and the boundary of each square is blurred, which may be caused by the poor directivity of the emitted neutron after the 252Cf neutron source is slowed down. Some discrete neutron points appear in the image, which may be caused by scattering of some neutrons and thermal noise of multi-anode photomultiplier tube. [Figure 4] is the projection distribution of two squares in X direction. Two relatively flat peaks can be clearly seen, and the neutron distribution in the right square is more uniform. The position resolution is about 2 mm (FWHM).

A multi-anode photomultiplier tube and waveform digitizer are used to develop a neutron Anger detector system, which is verified by GS20 scintillator. The preliminary test results show that the detector has good two-dimensional imaging ability. The neutron Anger detector system in the paper can be used to study the preformance of neutron scintillator materials. This work is supported by the Key-Area Research and Development Program of Guangdong Province (2020B0303090001) and Key Basic and Applied Research of Guangdong-Dongguan Joint Program (22S603N111).{Figure 20}{Figure 21}{Figure 22}{Figure 23}

Keywords: Anger camera, MAPMT, neutron detector, neutron scintillator, waveform digitizer


Riedel RA, et al. Nucl Instrum Methods Phys Res A 2015;794:224-33.

 Abstract - 51270: Design and development of area radiation monitor for Indian Cargo Scanner Facility

Shivam Agarwal, Amit Jain, Probal Chaudhury

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: GM-tube based Area Gamma Monitors are widely used for monitoring the dose rate generated by radioactive sources. However, the same monitors cannot be used for measuring the short, pulsed radiation field generated by accelerators. An accelerator with beam pulse width of 3-4 μs and pulse repetition rate of 200Hz is used in Indian Cargo Scanner Facility (ICSF) to scan cargo containers.[1] A Plastic Scintillator detector-based compact, AC-powered and wall-mountable area radiation monitor (ARM) with Modbus communication protocol has been designed and developed to monitor the accelerator-generated radiation fields at ICSF.

Block Diagram of ARM: The block diagram of ARM is shown in [Figure 1]. A 2” x 2” plastic scintillator, optically coupled with the photomultiplier tube (PMT), has been used in the design. Negative HV has been used to bias the PMT. The PMT generates the periodic, short current pulses proportional to the periodic, short radiation pulses generated by the accelerator. The output of PMT is directly coupled to the current to voltage converter (CVC) module. CVC module is a low input bias current, fast pulse integration amplifier with FET-input Op-Amp, integrating capacitors, and low leakage FET switches. It integrates the input current generated by the PMT for a set time and stores the resulting voltage on the integrating capacitor. The output voltage follows Equation 1.


where Cint = Integration Capacitance, tint = Integration Time, Iin (t) = Input Current, V0 = Output Voltage

The output voltage is held using the CVC module's hold switch during measurement. Once the voltage is measured, the integration capacitor is discharged using the discharge switch. The output of the CVC module is fed to the microcontroller module. The microcontroller module uses a PIC16F15323 microcontroller, an 8-bit microcontroller with 10-bit Analog to Digital Converter (ADC). The microcontroller controls the discharge switch & hold switch and reads the output voltage of the CVC module. The microcontroller sends the read voltage to the Single-Board Computer (SBC) over Universal Asynchronous Receiver/Transmitter (UART). A Linux Operating System (OS) based Beagle bone black SBC[2] has been used in the ARM. A Graphical User Interface (GUI)-based application has been developed in python language. The application reads the voltage data over UART and displays the dose rate (Read Voltage * Calibration Factor) on a 7” High-Definition Multimedia Interface (HDMI) display. The SBC sends dose rate data over the Ethernet using the MOD Bus protocol for readout at the remote console of ICSF. A universal serial bus (USB) interface has also been provided in ARM for the system configuration. ARM has a potential free Normally-Open Relay Contact. The relay is energized in normal conditions and de-energized when the dose rate crosses the set threshold. The relay remains in the de-energized state until the ARM is restarted. The radiation dose rate is stored locally along with the timestamp. Over-measurement range indication has also been provided in the system.

Results and Discussions: The ARM system has successfully been designed, developed, installed and tested at ICSF. As the system is integrating the current generated by the PMT, the system indicates the change in measured dose rate for the change in pulse height of the radiation burst and/or the change in pulse repetition rate of radiation burst and/or the change in pulse width of the radiation burst generated by the accelerator. Unlike the developed ARM, the GM tube does not indicate the change in dose rate for the change in pulse height and/or the change in pulse width of the radiation burst generated by the accelerator. The system measures the dose rate up to 100μSv/h and sends it over the Ethernet using the Modbus protocol for the dose rate read out at the central console.{Figure 24}

Keywords: Current to voltage converter, GM tube, ICSF, single board computer


Biju, et al. Shielding Evaluation and Radiological Safety aspects in the Design and Commissioning of Indian Cargo Scanner Facility, BARC Report No. BARC/2021/I/018; 2021.Coley G. BBB System Reference Manual; 2013. Available from: https://cdn-shop.adafruit.com/.

 Abstract - 51281: Application of internet of things technologies for environmental radiological monitoring

K. Sreekumar, S. K. Jha, R. Sujata, S. Ajeshkumar, M. S. Kulkarni

Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Outdoor dose rate measurements are carried out for various purposes such as (a) Environmental baseline studies prior to operation of any nuclear or radiological installation (b) During operational stage to ensure that the installations do not release any radioactivity to the environment (c) Emergency situations where presence of radioactive substances or any air borne radioactive material in environment is anticipated (d) Mineral resources survey, especially for Thorium and Uranium resources and (e) Academic studies. This paper discusses the application of Internet of Things (IoT) technologies for automation in environmental radiation survey, GPS tagging , ambient gamma Spectrum analysis and real time uplinking with a central server. Internet of Things (IoT) describes physical objects or groups with sensors, processing electronics, software that connects with other similar devices and exchange data over internet or any other connecting networks. Devices are connected directly to the network. MQ Telemetry Transport (MQTT) is a lightweight open messaging protocol used in IOT technology. In MQTT protocol, mobile network is used for low bandwidth data transfer. Indian Network of Environmental Radiation Tracking System (INERTS) is an indigenous platform being developed by Health Physics Unit, IREL (India) Limited, Manavalakurichi under Health Physics Division, BARC. The system consist of 05 Number of networked GM Surveymeters and 02 Number of networked Portable Gamma Spectrometers cum Survevmeters (Scintillator based). GM Surveymeters generate GPS tagged radiation dose rate data and upload to the server and Gamma Spectrometers generate both GPS tagged ambient gamma spectrum and dose rate. In both cases, additional readout and storage is provided with tuned android phones. This work is the up gradation of INERTS platform developed a few years earlier. The gamma spectometry features were added. The range of surveymeters is 0.01 -100 μSv/h, typical range measured in environment and emergency scenarios. The output from radiation detector module is averaged for 20 second for stable data. The GPS tagged dose rate data and ambient gamma spectrum data are transferred to INERTS Server through MQTT protocol. A system consisting of five GM detector based radiation survey meters and two portable gamma Spectrometer cum survey meter were designed and fabricated. All these instruments were linked with central server at Health Physics Unit, IREL (India) Limited, Manavalakurichi for real time environmental radiation mapping and data logging. Using these instruments , radiation survey and mapping of different locations of Thiruvananthapuram, Kanyakumari and Leh (Ladakh) were carried out and valuable dose rate data generated.{Figure 25}{Figure 26}{Figure 27}

Keywords: Dose rate, GPS, network, radiation

 Abstract - 51302: Validation of gamma ray spectrometry measurements for environmental radioactivity analysis

B. Arun1, I. Vijayalakshimi1, S. Viswanathan1, M. Menaka1, V. Subramanian1,2, B. Venkatraman1,2

1Safety, Quality, and Resource Management Group, 2Homi Bhabha National Institute, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Introduction: High Purity Germanium (HPGe) detector is widely used technique for evaluating the gamma emitting radio-isotopes in the environmental samples. Standard sources with traceability are employed for efficiency calibration of detector. Radio-analytical laboratories for measurements and calibrations are continually requested to provide evidence on the quality of their operations. This is mandatory in cases where radioactivity certification is involved, which supports the decision. The good quality in measurements is represented in terms of the accuracy and precision. Inter-comparison and proficiency exercises are useful to evaluate the performance of the analytical test method and measurement systems. The efficiency calibrations and measurements are validated by participating in proficiency test exercise. In this paper, results obtained in proficiency test with Terrestrial Environment Laboratory (TEL) of International atomic energy agency (IAEA) on environmental radioactivity analysis are presented.

Evaluation Method: High purity Germanium detector (HPGe) with 50% relative efficiency is used in this study. The results of proficiency tests IAEA-TEL-2020-03 and IAEA-TEL-2021-03 conducted by IAEA are reported in this paper. The results submitted by the laboratory are evaluated for accuracy and precision as per IAEA criteria. The final score is accepted if both accuracy and precision is accepted, “not accepted” if accuracy is not accepted and “warning” if accuracy is accepted and precision is not accepted.

Results: The results obtained for liquid samples for different radionuclides at different activity levels are reported in the [Table 1]. The reported values are accepted with respect to accuracy and precision values for all the radionuclides except 22Na. For estimation of 22Na, due to non availability of standard reference source for efficiency calibrations, mean efficiency value of 60Co (1173 keV & 1332 keV) was used. The target values and reported values for food matrix (Fish and Bamboo samples) are given in the [Table 2]. The reported values are accurate with respect to the target values for anthropogenic radio nuclides (134Cs and 137Cs). The measurements of 137Cs in fish sample are not precise as observed from the [Table 2].

Conclusions: The participation in proficiency exercises has demonstrated competence in radionuclide estimations in the environmental samples. The reported activities are accurate with respect to target values shows that the traceability of reference standards used for calibrations. The precision in the measurements verifies the combined associated uncertainty calculations.{Table 5}{Table 6}

Keywords: Environmental samples, HPGe, proficiency test


Iurian AR, Cosma C. A practical experimental approach for the determination of gamma-emitting radionuclides in environmental samples. Nucl Instrum Methods Phys Res A 2014;763:132-6.Yasser MG, Héctor CA, Carlos-Hernández MA, Carlos ND. Validation of an efficiency calibration procedure for a coaxial n-type and a well-type HPGe detector used for the measurement of environmental radioactivity. Nucl Instrum Methods Phys Res A 2016;818:51-6.

 Abstract - 51306: Determination of angular dependence factor for gamma radiation monitoring instruments

Pew Basu, M. Menaka, B. Venkatraman

Safety, Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

In the present study, the response of the detectors when in the radiation field and radiation falls on them at certain angle of incidence has been studied. This study will be helpful to understand the directional dependence of gamma radiation monitoring instruments. Towards this, the ambient dose equivalent H*(10, α) at an angle of incidence α and angular dependence factor H*(10, α)/H*(10, 0o) has been experimentally determined for the gamma rays of different energies and for incidence angles between ± 180o. The experiment has been carried out at Regional Calibration Facility (RCF) of Radiological & Environmental Safety Division (RESD), Indira Gandhi Centre for Atomic Research (IGCAR). The instruments used were Automess make handheld Teletector (Model No.: 6150AD/H) and Atomtex make Survey meter (Model No.: AT6130). The gamma radiation survey instruments were irradiated at a distance of 1.5 m from the source. Three collimated sources such as 10 mCi 137Cs, 1 Ci 241Am, and 1 Ci 60Co were used. The instruments were first placed at an incidence angle of radiation 00. Since changing the position and orientation of the source is not possible here, the instruments were rotated relative to their vertical axis by rotating the calibration bench. The angle of incidence was varied between ± 1800 and H*(10, α) was recorded. It is observed that the fluctuation in the angle response of both the instruments is maximum for low energy gamma radiations such as 60 keV photons from Am-241 as compared to high energy gamma radiations of 662 keV and 1250 keV from Cs-137 and Co-60 respectively. The reason behind the decrease in response with the rotation of the detector from the angle 00 is due to the effective attenuation of radiation field seen by detector/radiation monitor from different angles. The relative responses of Automess and Atomtex gamma radiation monitors with incidence angle between ± 180o has been plotted and shown in [Figure 1] and [Figure 2] respectively. In case of Automess instrument, H*(10, α)/H*(10, 0o) varies from 0.06−1.0 for Am-241, 0.52−1.0 for Cs-137, and 0.62−1.0 for Co-60. Hence, the maximum fluctuation goes up to 38% for Co-60 field, 48% for Cs-137 field, and 94% for Am-241 field for α between ± 180o. For α between ± 90o, the maximum fluctuation goes up to 12% for Co-60 field, 16% for Cs-137 field, and 45% for Am-241 field. For α between ± 45o, the maximum fluctuation goes up to 6% for Co-60 field, 9% for Cs-137 field, and 26% for Am-241 field. In case of Atomtex instrument, H*(10, α)/H*(10, 0o) varies from 0.1−1.0 for Am-241, 0.48−1.0 for Cs-137, and 0.57−1.0 for Co-60. Hence, the maximum fluctuation goes up to 43% for Co-60 field, 52% for Cs-137 field, and 90% for Am-241 field for α between ± 180o. For α between ± 90o, the maximum fluctuation goes up to 22% for Co-60 field, 34% for Cs-137 field, and 45% for Am-241 field. For α between ± 45o, the maximum fluctuation goes up to 14% for Co-60 field, 21% for Cs-137 field, and 28% for Am-241 field. Hence, the maximum fluctuation for the gamma radiation field varying from 60 keV to 1250 keV is within 30% for α between ± 450 and within 45% for α between ± 900. This study will be helpful to establish the directional dependence test in near future for the calibration of gamma radiation monitoring instruments at RCF.{Figure 28}{Figure 29}

Keywords: Ambient dose equivalent, angle response, atomtex survey meter, automess teletector, gamma monitoring instruments

 Abstract - 51311: Study of geometric efficiency variation for activity computation in portable TDCR system

M. K. Sharma1,2, M. S. Kulkarni1,2, Shivam Agarwal2, Sagar Baraiya2, D. B. Kulkarni2, Aatef Kamal Shaikh2

1Homi Bhabha National Institute, 2Radiation Safety Systems Division, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

The Triple to Double Coincidence Ratio (TDCR) method based on liquid scintillation (LSC) counting is standard method used worldwide for absolute measurement of activity of pure-beta and some electron-capture radionuclides. The three important parameter in this method to compute the activity are; the statistical law describing the number of photoelectrons created at each photocathode, the radionuclide spectrum and the scintillator non linearity model due to the ionisation quenching effect.[1] In the TDCR method, the detection efficiency of the LSC counter for double and triple coincidences (εD and εT) is theoretically computed from the statistical model of physical phenomena in the liquid scintillation detector. In TDCR method, if the model and the parameters are correctly evaluated, one must find the same source activity under different counting conditions. It is essential to change the efficiency of the LSC counter and analyse the activity measurement's quality to ascertain the model's correctness. In a well-designed counter, the counting efficiency is optimum, and the only remaining possibility is to decrease it. Several methods could be used for that purpose, namely phototube defocusing, grey filters and source chemical quenching. In the in-house developed FPGA-based portable TDCR system, the geometric efficiency variation is implemented based on the vial position in the chamber. A vial-holding portable chamber is placed above the counting chamber to position the vial in the optical chamber using a shaft, as shown in [Figure 1]. By pressing the shaft, the position of the vial is adjusted in the optical chamber. The optical chamber is designed considering a 20 ml vial filled with a 15 ml solution. The LS standard vial is held by the cap between pressure sensitive Teflon collet and the lift rod seat. The vial is lowered in the Counting Zone by pushing the shaft down. The collet is attached at the end of the spring loaded shaft. After the vial is seated on the lift rod cup and held in the collet, the chamber slide door is closed to make the chamber light tight. Pointer attached to the shaft indicates position of vial against the center of the detectors (PMTs) in the detection zone. After counting is over, by pressing the lock lever, the vial is raised to the vial chamber by the spring-loaded lift rod. The vial is removed by opening the chamber slide door. The effect of geometric efficiency variation on activity computation was studied using a toluene-based Eckert & Zeigler Analytics unquenched tritium standard in a 20 ml glass vial with a measurement uncertainty less than 4%. The HV of the three PMTs was set to the recommended voltage of -900 V. The efficiency variation was carried out by moving the vial up and down vertically and reproducibly around the centre of the PMTs in nine steps inside the optical chamber. The acquisition time is set as 60 s, and each activity computation/measurement was repeated ten times. The [Table 1] shows the vial position, efficiency and activity measurement result for various positions of the vial in the optical chamber. As expected, the efficiency of LSC counter is maximum at the centre position as the meniscus of the solution faces the PMT centre. A relative movement from centre position reduces the efficiency of the LSC counter. However, with change in the geometric position, the activity measurement results were found to be consistent.{Figure 30}{Table 7}

Keywords: Coincidence, efficiency, FPGA, TDCR


Cassette P, Broda R, Hainos D, Terlikowska T. Appl Radiat Isot 2000;52:643-8.

 Abstract - 51313: Development of online radiation monitoring system for exhaust filter banks

U. V. Deokar, A. R. Khot, Ashish Singh, Sonali Khurana, G. Ganesh, M. S. Kulkarni

Health Physics Division, Homi Bhabha National Institute, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Waste Immobilization Plant was designed for vitrification of High Level (HL) liquid waste from Reprocessing Plant using Metallic Melter. The decommissioning of plant was planned. The vitrification cell of plant contains highly contaminated equipment. During dismantling and cutting of these equipment, significant air-born particulate activity will be generated and which will result into increase in Radiation Level (RL) on Exhaust Filter Banks. For continuous monitoring and control of RL on Exhaust Filter Banks, Online Radiation Monitoring System was developed and installed. The system facilitate the operator to take decision in advance to changeover Filter Bank before it crosses permissible radiation level limit. Also it is helpful in saving of collective dose during replacement and disposal of filters. Online radiation monitoring system comprises of two large area side-window GM detectors installed on exhaust duct at 1.7 meter distance from the Centre of the filter bank. The multiplication factors in between contact RL on filter bank and online monitor reading was established by actual measurements and by shielding calculations based on Point Kernel method.[1] By connecting online radiation monitors to Centralized Radiation Potation Console (CRPC), is providing continuous radiological status of Filter Banks in Control Room and in Health Physics Shift Room. Due to remote online radiation monitoring, number of entries in amber area for frequent radiation survey are reduced, which resulted in considerable saving in collective dose & secondary waste. The multiplication factor of 19.55 was evaluated between online monitor reading and contact radiation level of exhaust filters. For evolution of multiplication factor correlation study was carried out by taking several radiation measurements of filter bank in different intervals. The graphical representation of statistical variation between calculated RL and estimated RL is given in [Figure 1] which is less than ± 10%. The results of measurements are presented in [Table 1]. In-cell air activity is gating loaded on exhaust filters, to know the concentration of airborne activity inside high level liquid waste treatment Cell during decommissioning, the continuous in-cell air monitor was installed. Based on in-cell activity concentration dismantling operation will be decided, as well as we will come to know the pattern of activity gating loaded on exhaust filter banks. The reliability and accuracy of the online radiation monitoring system was confirmed with actual radiation measurements and by theoretical calculations.[2] Conclusion: Online Radiation Monitoring System facilitate the operator to take decision in advance to changeover Filter Bank before it crosses permissible radiation level limit. The system is continuously logging dose rate of exhaust filter banks in Control Room and in HP Shift Room. Due to remote online radiation monitoring, number of entries in amber area are reduced, which resulted in considerable saving in collective dose & secondary waste.{Figure 31}{Table 8}

Keywords: Decommissioning, multiplication factor, point kernel method, vitrification


Deokar UV, Kulakrni VV, Purohit RG. Assessment of Radioactivity Content of Waste Packages at Waste Management Facilities. Tarapur: IANCAS-BRNS; 2010.Etherington H. Nuclear Engineering Handbook. 1st ed., Vol. 7. Mcgrow-Hill. p. 105-6.

 Abstract - 51347: Radiochemical synthesis of Eosin-Y and its application as radiation indicator in medical sterilisation

Sachin G. Mhatre, V. Sathian, Probal Chaudhury

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Radiation Sterilisation is carried out using ionising radiation for medical products to make them pathogen free (Medical Sterilisation). To distinguish radiation processed and unprocessed products radiation indicators are used. Radiation indicators (stickers) are chemical systems which are sticked on external packaging surface and change colour on receiving dose(threshold). As per international irradiation guidelines, it is mandatory to attach radiation indicator on every package to be irradiated. For medical sterilisation, a few indigenous radiation indicators are developed but are not available commercially. Most radiation indicators used in India are imported. Electrophilic halogenation of phenolic compounds is carried out using catalysts (Lewis acids) unless it has activating groups. Lewis acids form highly electrophilic complex that are highly reactive towards halogens. Halogenation of phenols is faster in polar solvents and alkaline environment due to the dissociation of phenol, as phenoxide ions are more susceptible to electrophilic attack.[1] Fluorescein, a xanthene dye is readily brominated to eosin-Y, a distinctly eye catching fluorescent red dye. This reaction is used for qualitative test presence of bromine. Eosin-Y is synthesized commercially by adding liquid bromine in alcoholic fluorescein solution at low temperatures. Electro-chemical synthesis of eosin-Y is carried out by passing electric current through solution of potassium bromide, fluorescein, sodium bicarbonate and acetate.[2] A new radiochemical method for eosin-Y synthesis was attempted based on chemical synthesis of erythrosine and other established methods of eosin synthesis.[3] Since bromine is more reactive, it was expected that it will selectively halogenate fluorescein in solid matrix like PVA. Alkaline solutions of potassium bromide, bromate and fluorescein were irradiated to kGy doses using Co-60 gamma radiations. Bromine is generated by action of (water) radiolysis products on bromide bromate solution. Generated bromine converts fluorescein to eosin-Y (halogenation) with the help of carbonate, in the presence of acetate. Ammonium acetate, a well-known brominating agent in many organic syntheses and was selected to enhance the rate of reaction. Polyvinyl alcohol (9%) films with glycerine (5%) produce films with good water retention properties. PVA films were made with the above composition. The cast films are packed in plastic pouches to retain moisture and avoid contact of oxidising chemicals. Films are fluorescent yellow when unirradiated and fluorescent red when irradiated on receiving radiation dose of around 15 kGy [Figure 1]. These films can be used as radiation indicators for medical sterilisation. The films are to be used within a month from preparation due to their limited stability.{Figure 32}

Keywords: Bromine, eosin-Y, fluorescein, polyvinyl alcohol, radiation indicators


McCullagh JV, Kelly A. Daggett synthesis of triarylmethane and xanthene dyes using electrophilic aromatic substitution reactions. J Chem Educ 2007;84:1799.Jagannathan E, Anantharaman PN. Electrochemical Preparation of Erythrosin and Eosin. Electrochem 3(1); 1987. p. 29-31.Volpe M, Cimmino A, Pezzella A, Palma A. Process for Synthesizing Halogenated Derivatives of Fluorescein for Use in the Production of Non-Volatile Memory Devices. US Patent no US20080061289; 13 March, 2008.

 Abstract - 51350: Simulation and experimental verification of ventilation filter dose rate mapping system

Sanjay Singh, Manish Tiwari1, K. R. Rathish, Jyoti Diwan1, M. K. Suresh Kumar1

Health Physics Division, Bhabha Atomic Research Centre, 1Waste Management Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: In the radiological plants, ventilation filter bank is an important engineered safety barrier to prevent atmospheric radioactivity release from the plant.[1] Increase in the pressure drop across the filter and radiation field, due to deposition of radioactive dust particles during continuous operation, limits the usable life of the filters. Therefore, dose rate of each filter in the filters bank needs to be measured for their optimum utilization. An online filter dose rate mapping system is developed and felt promising for this purpose.[2] Here, we present Monte-Carlo based simulation to arrive at the response factor, which relates the measured radiation field by the system to the actual radiation field of the filters and its verification with experimentally measured data.

Materials and Methods: The dose rate mapping system consists of Geiger Muller detector based gamma dose rate meter, located centrally at 1 meter from the cluster of 9 filters placed in a 3 x 3 matrix [Figure 1]. Measured dose rate by this system is translated into average contact dose rate of filters by multiplying a theoretically generated response factor which is ratio of dose rate on the filters in contact and dose rate at detector location as in system. For experimental verification of simulation results of the response factor, an actual filter bank which consists of 90 filters placed in 10 modules, was measured manually using pre-calibrated survey meter and the average contact radiation field of each module was calculated and using data of the dose rate mapping system, the response factor is calculated and compared.

Results and Discussion: First, dose rate data of each module was analyzed and found that co-efficient of variation is less than 20%. This implied the deposition is uniform within the nine filters of each ten module. With this consideration, in the simulation each filter is uniformly loaded with same activity of Cs-137 and response of detectors are calculated. However, for illustration, relative response (with respect dose rate of only one fully loaded filter module) of only 5 detectors is shown in [Figure 2]. An S-shaped response seen in [Figure 2] is due to contribution of the gamma flux from other modules to a particular detector. This contribution is minimized by providing collimator around each detector. After providing suitable collimators, response factor is re-calculated. It is seen in [Table 1] that the variation between measured and calculated values does not differ significantly.

Conclusions: Good agreement between calculated and measured response factors indicates filter dose rate mapping system is a good alternative of the manual measurement of ventilation filters. In addition, real time dose rate data can be used for effective utilization of the filters. Based on the dose data, uniform deposition among the filters can be achieved just by changing the dampers direction, which regulate the flow streams at the filter inlets.{Figure 33}{Figure 34}{Table 9}


AERB. Radiation Protection for Nuclear Facilities, AERB/NF/SM/ O-2(Rev 4). Mumbai, India; AERB; 2005.Tiwari MK, Kanojia AK, Raval RV, Shreekumar PR, Sur N, Wadhwa S, et al. Proceedings, IARPNC-2020 Conferences on Radiation Safety in Nuclear and Core Industries, Health Care and Environment; 2020-2021.

 Abstract - 51352: Commissioning experience of liquid scintillation counting system in PFBR

Vidhya Sivasailanathan, N. Suriya Murthy, Allu Ananth

Health Physics Unit, Prototype Fast Breeder Reactor, BHAVINI, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Introduction: Health Physics laboratory in PFBR is established by specifying, procuring, installing and commissioning the nuclear counting systems like alpha and beta counting systems together with radiation protection instruments such as radiation and contamination monitors. A liquid scintillation counter (LSC), Hidex 300SL was procured and installed in PFBR. The ground water analysis for presence of tritium (3H1) and 14C in the sample water can be analyzed using LSC. This manuscript discusses the efficiency calibration and estimation of MDA of the LSC.

Materials and Methods: The Hidex 300SL is a liquid scintillation analyzer with an automatic sample changer and a triple photo multiplier tube detection assembly. It is used to measure liquid beta samples in all typical scintillation vials, from 7 to 20 mL loaded in sample racks.[1] The instrument uses automatic sample calibration activities using the TDCR absolute counting method. It runs with an embedded software controlled by MikroWin 2000 Hidex 300SL user interface software. This software controls all the instrument functions as well as providing data processing and analysis, and reporting functions. In Mikrowin technology, the values that are available for use as raw data matrices are referred to as labels. Temperature, Time, counts, CPM, DPM, TDCR, QPI, QPE, Chemi, DTime and End time are the values available.

Efficiency Calibration of the Counting System: The system was connected to power supply and the Mikrowin software was installed. The counting system was calibrated using the standard liquid sources provided by the supplier along with the system. The source kit consists of two radioactive sources (Tritium and Carbon-14) and one background. The tritium source with activity of 204500 dpm as on 01.10.2007 and Carbon-14 source with activity of 102400 dpm as on 01.10.2007 were used. The sources were contained in sealed vials. The efficiency was found by applying the current activity of the standard sources. In Hidex, the efficiency is determined using TDCR (Triple to Double Coincidence Ratio) technology. The efficiency for unquenched standards was calculated to be 70.3% for tritium and 96.65% for 14C. The counting of tritium and 14C standard vials was done.[2]

Results and Discussion: The instrument was calibrated initially and the minimum detectable limit is estimated prior to analysis of the liquid samples. The following steps were carried out. Glass vials are better than plastic vials as the background counts will be reduced while using the former. The background sample can be made in a ratio of 8:12 or 5:5 ml while mixing with the scintillation solution based on the requirement. The cocktail (Aqualight-B) solution is used as the scintillator. 5 ml of distilled water (tritium free) is pipetted out in the plastic vial. Same quantity of cocktail (Aqualight-B) was pipetted out in the same vial containing distilled water in comparatively dark ambience. It was mixed thoroughly and kept overnight as leak tight in a cool ambience. The template with desirely modified values was created and with the required matrices as labels and saved with a file name. The sample loaded location was checked and the counting time was fixed as one hour. The range of “0-200“was selected which is suitable range for industrial grade low beta and the count repeat was chosen as 5. Then couting was started and the results were viewed in the “results” page. The complete data file was stored with an extension (.dat). The MDA was calculated using the following formula.


For representing the activity in litres (Bq/l), the value of the activity is divided by 1000. This usage will be convenient whenever the activity of the sample is of very less value. To compare the limits of the activity to identify the category of the liquid sample as per the limits recommended by the regulator for category of liquid effluents, which is referred in Bq/m3, the value in Bq/l needs to be converted to Bq/m3 [1m3=1000 litre and 1 litre = 10-3m3]. The background sample was counted for one hour duration and the values are shown in [Table 1]. Minimum Detectable Activity for the sample counted for 1 hour was obtained as 0.029 Bq/ml. The Figure of Merit (FoM), was found as 47.31. FoM is defined as the square of the counting efficiency divided by the background in the region of interest (E2/B), It is considered to be the best indicator of determining sensitivity of the measurement of the isotopes. Lower the MDA favours desirable higher FoM.{Table 10}


User Manual of Hidex 300 SL.Procedure for Calibration and Estimation of MDA for Analyzing Tritium in Water Samples Using LSC, PFBR/01150/Proc/46; 2021.

 Abstract - 51354: Design and development of rapidly deployable environmental radiation monitoring system

M. V. R. Narsaiah, Shashank Saindane, Nitin Bhosale, Vaibhav Bhujbal, S. Murali

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Radiological emergency preparedness requires proper instrumentation for quick impact assessment for initiating the countermeasures[1] and response to mitigate the consequences. The Rapidly Deployable Environmental Radiation Monitoring System (RDERMS) is a recent development for measuring radiation level in an affected area during radiological emergency scenarios. It is a compact battery-operated low-cost Geiger-Muller (GM) detector based portable environmental monitoring system. It provides the dose rates of the area continuously along with timestamp. It is a rugged environmental system with a self-sufficient power and mode of data transfer to a remote station located at few hundreds of meters of the line of site. It works on a wireless mesh networking system[2] for data transmission.

Architecture and Working Principle of the System: The newly designed RDERMS has four basic components: Detector, Electronic Interface, Microcontroller firmware and Zigbee transmitter. The optional components are PC, software for data analysis and presentation. It is a stand-alone unit with two GM detector tubes of different sizes with an acquisition time interval of two minutes. Both tubes are energy compensated with internal copper lining and external perforated lead wrapping in the range 100 keV to 1.2 MeV. The larger GM tube is for measuring low radiation field (< 0.1 mGy/h) while smaller GM tube for high radiation field (< 10 mGy/h). The HV unit is always “ON” for larger GM tube and when the reading exceeds 0.05 mGy/hr small GM tube gets activated electronically. The response is linear with sensitivity 180cps for 10 μGy/h. The Wireless transceiver turns on periodically for 10 sec to send data package. It is normally off during acquisition to save power. The system's photograph and schematic diagram showing interface component of RDERMS is shown in [Figure 1].

Methodology: RDERMS consists of data acquisition, data receiver, data communication and display modules. The receiver module is configured with PC through the serial communication port. Data acquisition module decides the sequence of the radiation monitoring unit to be called based on MAC-id and Radmon id stored in database. The acquired data contains dose rate, temperature of the enclosure, battery voltage, date and time, Unit ID and GM tube number. RDERMS contains ten units used as network of Radiation Early Warning System (REWS) to monitor radiological status of BARC site. The response of two such units to measure radiation level during the month of June-2022 is shown in [Figure 2]. The systems showed 69-242 nGy/h background radiation level at deployed location.

Response and Calibration: The system is calibrated with a 3.15 MBq standard 137Cs source and where contact measured dose rate was 2.74 mGy/hr and calculated dose rate was 2.66 mGy/hr. The source position aligned with centre of detector to expose maximum area of GM tube. The measured dose rate and calculated dose rate for identical source detector configuration tabulated in [Table 1].

Conclusion: The system can be deployed at a short notice and is very useful during emergency conditions to transfer data, assess the dose rate and its trend with timestamp at the remote station or an incident command post with the limited resources in emergency conditions. The System has wider scope for expansion, can be adopted for city's important installations to monitor in advertent movement of Radioactive Material.{Figure 35}{Figure 36}{Table 11}

Keywords: Continuous operation, low-cost, rapid deployable radiation monitoring


Saindane, et al. On-line environmental radiological data acquisition system for nuclear facilities and sites. Bull Radiat Prot 2003;26:1-2.Saindane, et al. Systems for Detection of Illicit Movement of Radioactive Material Inside/Outside the Facility. Vienna: IAEA International Conference; 2018.

 Abstract - 51358: Source term estimation using particle filter algorithm based on dose rate

Amit Silswal, Jis Romal Jose, Mukesh Sharma, Shashank Saindane, Probal Chaudhury

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Source term of a radionuclide is an important parameter for the radiological impact assessment following the routine or accidental release. In this regard attempts have been made by the researchers for the estimation of source term by using inverse modelling. The stochastic nature of the involved physical processes in the atmospheric dispersion of the released radionuclide and uncertainties in the meteorological data poses a challenge for the accurate prediction of the source term. For improving the accuracy in source term prediction, various studies had been carried out by using different methods such as bootstrap and bipivot method, iterative ensemble Kalman filter[1] etc. This paper discusses about the particle filter based algorithm for the source term estimation using dose rate measurements from the installed gamma monitors.

Methods: The statement for the current problem is: Suppose a gamma emitting radionuclide is continuosly being released from the stack of a nuclear facility. The dose rate from this radioactive plume is continuously monitored by the detectors installed at various locations around the facility. Meteorologcal data such as wind speed and stability class is measured by the local meteorological station. Source term i.e release rate of the radionuclide be estiamted using particle filter algorithm based on the dose rate and meteorological parameters. Particle filter is a Sequential Monte Carlo based technique to estimate the hidden states of a system. In the current problem, hidden states are source term and wind direction. A particle represents the potential release rate and wind direction. Particle filter involves, generation and initialization of particles, and prediction of the state of the particles according to the discrete time stochastic model known as the System model. Each predicted state of the particle is then weighted accoding to the likelihood of the particle being true state using the Meausrement model. The steps in the algorithm involves.

Obtain dose rate measurement (Dmeasure) from the detectors which register dose rate above the minimum detection level.Initialize the particles (No. of particles: Npar): Each particle X represents a potential release rate (Q) and wind direction (θ): X = [Q, θ]TFor the detectors detect the radioactive plume

Predict the state of the particles using System model.Estimate weight of the particles using Measurement model. Measurement model for the ith detector is given


Where Di is the dose rate in [INSIDE:1], E is the energy of the gamma ray photon in MeV, μa is the energy absorption coefficient (m-1), Bis the buildup factor, Ri location of the detector, μ is the attenuation coefficient (m-1 ), χ is the Concentration of radionuclide in plume (Bq/m3).

Particle weight is defined as :


Resample the particles as per the weights.Estimate source term by taking mean of the particles:


Results: The developed algorithm has been tested using a simulated release and dose rate for release height 100m, gamma energy 1.29 MeV. The parameters used in the simulation are: Npar = 20000, ranges of possible release rate and wind direction are [10, 2500] MBq/sec and [0-359] degrees respectively.

The algorithm was run 30 times and [Figure 1] shows the estimated and true source term. The average percentage error in the source term estimation is less than 8 %.{Figure 37}

Keywords: Particle filter, source term


Zhang X, Su GF, Chen JG, Raskob W, Yuan HY, Huang QY. Iterative ensemble Kalman filter for atmospheric dispersion in nuclear accidents: An application to Kincaid tracer experiment. J Hazard Mater 2015;297:329-39.

 Abstract - 51390: Development of server-based web application for radiation surveillance system

Sagar Baraiya, Puneet Jindal, Shivam Agarwal, Amit Jain, M. K. Sharma, Probal Chaudhury

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: The Radiation Surveillance System (RSS) is designed to detect the unauthorized movement of radioactive material using plastic scintillator detectors. The RSS generates an alarm along with the event video footage when the dose rate crosses the set threshold. It consists of three modules; the detector module, the data logger module and the server-based web application module. The data logger module records and sends the live dose rate data measured by the detector module to the server-based web application. This paper discusses the development of server-based web application module.

System Description: A server-based web application has been designed and developed for the Radiation Surveillance System. It provides database management and the web-based graphical user interface (GUI) to the authorized users of RSS. It receives the data from the data logger over HTTP, stores in the database, and presents in graphical form on a browser-based GUI.

Application Software Architecture: The software architecture of the server-based web application is shown in [Figure 1]. The application comprises four modules; a front-end application, an application server, a database server and a “Nginx” web server. The front-end application is a static website hosted using the “Nginx“[1] web server and runs on a web browser. It consists of HTML, CSS and Javascript files, providing an intuitive GUI to the users. This application has been developed using the typescript-based framework “Angular“.[2] It communicates with the application server using the HTTP and WebSocket protocols. The application server has been developed to handle the client requests sent from the front-end applications and database management. Various APIs have been developed for the different functionalities such as user authentication, receiving data from RSS, uploading data to the database, fetching data from the database, etc. The application server is divided into two modules; “auth” module and “gui” module. The “auth” module implements the functionalities related to user account management, whereas the “gui” module implements the functionalities related to the radiation data and the alarm videos. The application server has been developed using the python based web framework “FastAPI.“[3] PostgreSQL, an open-source Relational Database Management System (RDBMS), has been used for data storage. The role-based access control has been implemented with three types of logical users; data logger, data viewer and super user. The data viewer can access the current and previous dose rate profiles and the alarm event's videos footages received from the data loggers. The user account management feature has been provided to the superuser. The “Nginx” server has been configured as an API gateway and forwards requests from the clients to the application server. The data loggers send dose rate data in the normal conditions and event video footage in case of radiation alarms to the application server using POST request over HTTP. In addition, the application stores the received data into the database and broadcasts over WebSocket connections to the front-end application.

Result and Discussion: The application has been tested with the simulated field conditions and found to be working satisfactorily.{Figure 38}{Figure 39}

Keywords: Radiation, server, surveillance, web application


Available from: https://nginx.org/.Available from: https://angular.io/.Available from: https://fastapi.tiangolo.com/.

 Abstract - 51396: Development of android apps for the detection of orphan gamma radioactive source using smartphone camera with timebound access

Shaikh Aatef Kamal, Jis Romal, M. K. Sharma, Probal Chaudhury

Radiation Safety Systems Division, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: When information about a suspected radioactive material is received by the radiological emergency response authorities, it is preferred that the presence of a gamma radioactive source and its approximate strength is quickly confirmed. It will help the authorities to plan further course of action. CMOS camera of smartphone can be used for detecting the presence of gamma radioactive source and estimating approximate dose rate.[1] Present abstract describes the development of three standalone android apps to detect the presence of suspected orphan gamma radioactive source in the public domain using smartphone camera and communicate the encrypted measurement data to the authorities. The apps are intended to be used for the preliminary investigation in the absence of specialised radiation monitoring instruments at the incident site. The person in field having access to suspected source will be using his smartphone to make the measurement. The result of the measurement will not be visible to the person to avoid the misinterpretation of the result. The encrypted result is communicated to the authority for the analysis to confirm the presence of the gamma radiation and approximate strength of the gamma field.

Design Details: Three android apps have been developed namely “RadRecon“, “KeyGen“, and “RadReader“. The radiological emergency response officials share the installation file (.apk) of “RadRecon” app to the person in field. The official then uses “KeyGen” app to generate license key for the RadRecon app and shares it with the person via SMS/E-mail or WhatsApp. The person in field installs the app, follows the instructions displayed in the app, and finally shares the encrypted results file with the official for decryption and interpretation using “RadReader” app. The flow of operation is depicted in [Figure 1]. The details of the apps are as follows: “RadRecon“ is the app which is provided to the person in field. This app only works with a valid license key and stops working after the expiry of the license. The user has to install the app, enter the provided license key, cover the front camera of smartphone with a few layers of any opaque tape and initialise the app when prompted. After successful initialisation, the person takes the measurements by placing the phone near to the suspected material. The measurement results are stored in the encrypted database. User shares the encrypted results file via E-mail/WhatsApp or any other file sharing medium with concerned official. SQLCipher database that adds 256-bit AES encryption to database files on the standard SQLite is used for the database encryption. “KeyGen“ app is used by the radiological emergency response official to generate the license key for RadRecon. The official enters validity in days and hours along with user identification information (name, location etc) which is embedded in the key. This key along with the installation file (.apk) of RadRecon is shared with the person in field. “RadReader“ app is used by the official to decrypt the measurement file sent by the person in field. The decrypted data contains the exposure level and user identification information which will be used for identifying the person who has shared the file. The results are saved in local database on smartphone and can be exported and shared in plain text format. “RadReader” software is also developed for windows platform to run on desktop computers. [Figure 2] shows the applications installed on android device.

Conclusion: Use of smartphone camera-based radiation detector enables confirmation of the presence of gamma source and estimate approximate strength of the gamma field in the absence of specialized radiation monitoring instruments. Timebound license key reduces the risk of unauthorised distribution of the app and encryption of measurement data to avoid misinterpretation of the results.{Figure 40}{Figure 41}

Keywords: Android, camera, CMOS, encryption


Shaikh A, Sharma MK, Kulkarni MS, Romal J, Gupta A, Chaudhury P. Radiat Prot Environ 2017;40:116-20.

 Abstract - 51397: Design and development of single board computer based contamination mapping and image acquisition module for contamination monitor

Shuchita Bahadur, Vaishali M. Thakur, Amit Jain, M. K. Sharma, Probal Chaudhury

Radiation Safety Systems' Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Personnel working at nuclear and radiological facilities may inadvertently get contaminated with β/γ radiation sources in their routine work. The purpose of whole body contamination monitoring is to reckon the hazard from suspected contamination over body without manual frisking. Hence, the Whole Body Contamination Monitor (WBCM) was designed and developed for representation of 'contaminated versus clean' to the personnel monitored and also to the radiation safety officer in form of a comprehensive and reportable compilation.

Materials and Methods: The architecture of WBCM can be divided into Detector System (DS) and Mapping and Acquisition Module (MAM) as shown in [Figure 1]. The detector system consists of plastic scintillator detectors and associated electronics as shown in [Figure 2]. The Mapping and Acquisition module consists of Raspberry Pi Single Board Computer (SBC) and ancillaries. There are 13 plastic scintillator detectors placed for gross coverage of contamination on whole body i.e. head, feet, hands, front and back. The count data from all detectors alongside status of occupancy is sent to SBC by pulse processing hardware over RS-232. The occupancy state of the portal is marked “occupied” as per data received from pulse processing hardware. For a typical acquisition time of 10 seconds, the Minimum Detectable Activity (MDA) was calculated to be 45Bq using a 204Tl βsource which in turn corresponds to contamination of 0.15Bq cm-2. The software identifies the “alarm” or “no alarm” state as per the set thresholds and likewise renders colours as shown in [Figure 3]. If occupied, a photograph of the personnel is captured using a camera [1] installed on the wall of the portal and displayed on interface for each acquisition cycle. Software also compiles day-wise reports of contamination data with the photograph of monitored personnel in pdf format.

Results and Discussion: As shown in the figure, the system could successfully mark the suspected location as “contaminated” on the phantom. The system integrates the contamination data to a lucid and well-rounded illustration that completes the personnel monitoring and data is rendered as a visual cue to the user.{Figure 42}{Figure 43}{Figure 44}

Keywords: Acquisition, contamination, image, mapping, single board computer, β/γ


Available from: https://www.farnell.com/datasheets/2056180.Available from: https://pyserial.readthedocs.io/en/latest/pyserial.html.

 Abstract - 51405: Web based remote radiation monitoring portal for nuclear facilities

Vaibhav Bhujbal, M. V. R. Narsaiah, Shashank Saindane, Nitin Bhosale, S. Murali

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Safe handling and regular monitoring of nuclear material in Nuclear Power Plants, universities, agriculture, medical industries and various other industries, is the utmost imperative aspect to evade the regulatory violation. Also, detecting any attempts of inadvertent activity, unauthorized trafficking[1] of the radioactive material and radiation safety of radiation workers is realized as a radioactive-material-handling facility's prime responsibility in recent decade. Real time[2] radiological status information with an easy access and appealing visualization on a multi-platform, is important for quick radiological impact assessment at short notice. The developed radiation monitoring web portal is web-based system application encompasses state of the art browser-based interface with elementary salient features for data analysis and taking necessary remedial actions. The real time remote monitoring system consists radiation detectors connected to web server for data acquisition and client web browser for data display, connected through TCP/IP network.

Architecture and Working Principle: Radiation monitoring system [Figure 1] is a network of GM tube-based radiation detector monitoring devices, web server and standard web devices (i.e. computers, laptops, smartphones etc.). Web server receives real time data from radiation monitoring devices and stores it in database. Radiation monitoring devices has an inbuild memory database to store data in offline condition. After restoring connectivity, stored data sent to web server instantly. Algorithms written on web server communicates and fetches data from radiation monitoring devices asynchronously to avoid delayed response, if any due to break down of any of radiation monitoring device. Open source MySQL database server and stored procedures has been used to store the data. Java 2 Platform, Enterprise Edition(J2EE) standard platform has been used to develop the web portal. J2EE has enterprise features such as distributed computing and web services. Web portal is a collection of many servlets and Java Server Pages (JSPs). These block of java codes/pages executes on application server after URL request receives from client browser and response send back to client browser in Hyper Text Markup Language (HTML). One of the page response has been shown in [Figure 2]. The decision makers can request for real time radiation dose rate data from any standard web devices, connected to network & thus get faster access to radiation dose rate data. The stringent encryption algorithm and protocols in web portal have ensured data security and data protection. Methodology: The portal is a web application, developed using NetBeans IDE and it is comprising of MySQL database, java based servlets, JSPs at backend and jQuery, JavaScript, Bootstrap, CSS and HTML at frontend. It has salient features like user account/profile setting, delegating user-level responsibilities, visual and audio alarm indication, email services for alarm and reporting, alarm management, configuration and inventory management of radiation monitoring devices, chart representation, appealing georeferenced map dashboard [Figure 2], authority contact directory, availability and working status of radiation monitoring device on dashboard etc. These are some elementary features have incorporated as per system requirements. This radiation monitoring system can be implemented over Wide Area Network (WAN), Local Area Network (LAN)[3] or organizational network.

Conclusion: Web portal has a flexibility over a design as per the state of the art requirements with minimal efforts, makes it readily available for organizations handling radioactive material. The system can be deployed at any facility for radiation safety of radiation workers, detecting any attempt of inadvertent activity, unauthorized trafficking of the radioactive material.{Figure 45}{Figure 46}

Keywords: JAVA, MySQL, radiation monitoring, real time data, web application, web portal


Saindane S, et al. Systems for Detection of Illicit Movement of Radioactive Material Inside/Outside the Facility; IAEA International Conference; 2018.Bhujbal V, et al. (IARPNC-2020), a Development of Real Time Radiation Early Warning System for Nuclear Facility Using SDLC; 2020.Narsaiah MV, et al. Local Area Network Based Radiation Early Warning System for Nuclear Facilities; The International Conference on Decision Support System for Early Warning and Mitigation of Disaster (Durgapur), West Bengal; 2014.

 Abstract - 51415: Design and development of a floor scan contamination monitor

Anand Raman, Govinda Mukherjee, S. Murali, Probol Chaudhury

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

All nuclear facilities are mandated with the regulatory requirements which govern the permissible contamination levels on various physical entities, which include the physical extremities of the working personnel, their clothing and shoes and also the contamination levels on the floor of the working areas. The limits pertaining to those prescribed by AERB corresponds to 3.7 Bq/cm2 for beta/gamma contamination. This paper describes the design and development of a comprehensive Floor scan contamination monitor mounted on a portable platform. The system is intended to provide a rapid beta/gamma contamination status of the monitored surface. The Floor Scan floor contamination monitor is intended for the measurement of surface contamination by beta and gamma emitting radionuclides on floors. The monitor can be used in nuclear power plants, radio-chemical plants, research institutes with reactors, PET centres, etc. For different workplaces, an appropriate one or two channel model can be utilised. The monitor is equipped with scintillation detector(s) with integrated photomultipliers and signal processing electronics, and a mechanism providing easy, fast, and precise height adjustment for the convenience of the user. It is propelled by the operator on four castor wheels. The detector's protective grille provides good protection against foil damage while still allowing good penetration of alpha and low energy beta particles. The device is presently available with one detector (detection area 525 cm2). The measurements are shown on a display unit as a digital readout. If contamination higher than a pre-set threshold is detected, the acoustic alarm will sound. Two different alarm threshold levels that can be set by the user has been provided in the system. The system comprises of two detectors, (1) a wide area Thin film plastic scintillation detector and (2) a GM detector. The 150 x 350 x 0,5 mm plastic scintillation detector mounted in optical contact with an embedded 2 pi sensitive, 28 mm diameter, PMT with a 100 Mohm divider and connections via flying leads RG – 174 wires. The detector matrix is housed in a light tight vinyl protected enclosure. The entrance window for the radiation particle is through an 8.8 mg/cm2 layer of mylar-reflector interface. The GM detector incorporated in the system for ambient gamma dose rate measurement has a sensitivity of Sensitivity of Geiger Muller detector: 180 cps/.01 mGy/hr. The detector signals are processed by indigenously designed and developed amplifier modules and translated in to corresponding TTL signals. These signals are further input to a high end microcontroller module (DSPIC) and the results depicted a TFT display. The system is powered by a Rechargeable Li-Po battery (11.1V) charged from mains (230V/50Hz) cable. The display provides (1) BeG: Beta count rate in cps (2) Gm: gamma count rate in cps and (3) Max count rate: Gamma count rate from GM counter. A reset button provides user option to reset the system. The embedded firmware in the microcontroller module provides a graphical user interface, with provision for various modes of monitoring. The algorithm, incorporated in the firmware caters to beta/gamma contamination monitored on the surface in the presence of a significant ambient gamma radiation dose rate to the extent of 1 mSv/hr. The response of the system in a few instances of stimulated contamination is given below:


It is observed that the beta/gamma sensitivity of the plastic detector to ambient background is approximately 30-50 cps for a measured field of about 5-10 μR/hr giving a rough estimate of 0.1 cps/cm sq as the background. Any beta contamination above this value becomes significantly highlighted in the display.


Keywords: Floor contamination, thin film plastic scintillation detector


Kumar A, Raman A, Babu DA, Sharma DN. Development of a plastic scintillator based large area ground surface contamination monitor. Radiat Prot Environ 2011;34:41-3.

 Abstract - 51430: Determination of Fano factor of an HPGe Gamma-ray detector

S. Chinnaesakki1, M. R. Dhumale1, S. V. Bara1, S. K. Jha1,2, M. S. Kulkarni1,2

1Health Physics Division, 2Homi Bhabha National Institute, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

HPGe detector has been used in gamma-ray spectrometry extensively in many areas of nuclear science and technology. The overriding characteristics of HPGe is their superior energy resolution. Energy resolution in terms of full width at half maximum (FWHM) can be measured using gamma-ray sources. It depends upon the number of charge carriers produced by the incident gamma photon, electronics noise and charge collection efficiency. The relation between FWHM (WT) and the compounding factors can be expressed as follows:[1],[2]


WT – Total width of the peak (FWHM),

WE - Electronic noise,

WS – Statistical fluctuation in charge carriers, and

WC - Fluctuations due to incomplete charge collection.

Statistical fluctuation is given by:


(E – Incident gamma energy in keV, ε is the energy required to create one electron-hole pair). WC - Fluctuation in charge collection efficiency. Eqn. 2, indicates that resolution improves as number of charge carriers increases. The average energy (ε) needed to produce one electron-hope pair in HPGe is 2.96 eV, although the band gap (Eg) is 0.74 eV. This difference between ε and Eg indicates that there is competing processes which do not produce charge carriers. Variance in charge carrier production, if the process is only statistical in nature, as per the Poisson statistics,


The observed variance is smaller than predicted by the above equation. The Fano factor F is introduced as an adjustment factor to relate the observed variance, it is given by:


The Fano factor for detector with good resolution should be as small as possible. It is estimated by experimental FWHM. This paper discusses the determination of Fano factor of an p-type HPGe with 50 % relative efficiency. The detector has been coupled with 64 k DSP MCA module. Standard reference materials have been used for the calibration of the system covering the energy range 46 keV to 2614 keV. FWHM in keV were obtained using the point sources (241Am, 133Ba, 57Co, 60Co, 152Eu and 22Na) placed independently at a distance of 25 cm from the detector top. As shown in [Figure 1], Fitted (FWHM)2 as a function of energy, is given by

(FWHM)2 = 0.7411 + 1.57 × 10-3E + 5.2509 × 10-7E2 (5) The coefficient of this fit can be used to calculate the components of eqn.1. From the above equation,

WE)2 = 0.7411 , (WS)2 = 1.57 × 10-3E, and

(WC)2 = 5.2509 × 10-7E2 . The fitted values of (WT) 2, (WS) 2, (WE) 2, and (WC) 2, were calculated and plotted as shown in [Figure 2]. Fano factor (F) was calculated using equation 4.

It is well known that (WE)2 is independent of E, (WS)2 is proportional to E,also evident from Eq. 2, and (WC)2 is proportional to E2.[2] The determined F in this study was found to be 0.096, which is well within the range of literature values. F value less than 0.1 indicates the better energy resolution of this semi coaxial HPGe than the conventional coaxial HPGe. {Figure 47}{Figure 48}

Keywords: Fano factor, FWHM, HPGe


Harkness-Brennan LJ, et al. Nucl Instrum Methods Phys Res A 2014;760:28.Knoll GF. Radiation Detection and Measurement. 4th ed. New York: Wiley; 2010.

 Abstract - 51441: Development of advanced compact radiation monitoring system

Ashutosh Gupta, Shuchita Bahadur, JisRomal Jose, M. K. Sharma, Probal Chaudhury

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Radioactive sources are widely used in various medical and industrial facilities. Mis-handling or theft of these sources may lead to radiological emergencies. To cater such emergencies and routine radiation monitoring surveys, Advanced Compact Radiation Monitoring System (ACRMS) has been designed and developed. ACRMS is a battery operated stand-alone compact radiation monitoring system that uses two ambient dose equivalent Geiger Muller tubes to measure the gamma radiation, displays it over 7” touch screen LCD along with recent trend in the radiation level, current date, time & GPS co-ordinates and archives the same.

Materials and Methods: ACRMS is a commercially available power bank operated Raspberry Pi based system as shown in the [Figure 1]. An HV module mounted in the peripheral module provides High voltage (500V) for two GM detectors (Low sensitivity and high sensitivity). A python application has been developed for RaspberryPi to acquire the GM tube pulses over GPIOs for user settable acquisition time. Application also interfaces with GPS connected over USB/serial to obtain positional co-ordinates. The system has on-board battery backed Real Time Clock (RTC) to obtain date & time information. A Tkinter based GUI[1] containing multiple tabs has been developed in python as shown in [Figure 2]. After every acquisition time, counts are averaged over user settable window size and corresponding colour coded dose rate & integrated dose, current date & time and GPS co-ordinates are updated in home tab of the GUI. Circular analog panels in log scale have also been developed in Python to display dose rate and integrated dose. Dose rate histogram with user settable histogram size is also displayed to show the recent trend in the dose rate. MAP view tab has been provided to depict the variation of the real time dose rate over GIS MAPs as shown in [Figure 3]. Algorithm has been developed to acquire the static map images, build the map using those images[2] along with Zoom in & out capabilities and to put colour coded markers based on current dose rate. Based on the prevailing dose rate, system automatically selects GM detector to measure the dose rate and integrated dose. Provision has been provided to transfer the real time dose rate data tagged with date, time and position over Serial port. At the end of the survey, dose rate data tagged with positional and timing information are stored in a CSV file. A pdf report containing dose rate profile along with all the details of the alarm is also generated at the end of the survey. The system supports offline analysis of the saved CSV file and generates the dose rate time profile for the same. Admin tab allows the user to view and alter the password protected system parameters such as acquisition time, date & time, Alarm threshold level etc.

Results and Discussion: RaspberryPi SBC based ACRMS has been developed catering to the operating range from background to 5mSv/h. The system has a detachable power bank and can work continuously for more than 8 hours.{Figure 49}{Figure 50}{Figure 51}


Available from: https://docs.python.org/3/library/tkinter.html.Available from: https://developers.google.com/maps/documentation/javascript/coordinates.

 Abstract - 51447: Validation study of in house prepared quench standards for tritium counting using liquid scintillation analyser

Ajay Kumar Gocher, S. N. Tiwari, I. V. Saradhi1, A. Vinod Kumar2

Environmental Survey Laboratory, EMAD, BARC, Rawatbhata, Rajasthan, 1EMAD, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Tritium (3H) is most important radionuclide which is being monitored rigorously at PHWR sites using Liquid scintillation Analyser (LSA). Ready to use Di-isopropyl naphthalene (DIN) based certified quenched standards of M/s Perkin Elmer are being used for calibration of LSA at ESL, Rawatbhata, Rajasthan. In this study, an attempt was made to prepare in-house quenched standards for calibration of LSA and its validation by counting of known standard activity of tritium.

Instrument: Liquid scintillation Analyser (Quantulus 1220™) of Perkin Elmer make was used for tritium measurement. The spectra was acquired and evaluated by WinQ and EASY View software respectively. The counting protocol of the Quantulus 1220 system was set as follows: configuration: 3H (low energy beta), send spectra: 12 MCA, coincidence bias: low, Counting window (in EASY View for sample counts): 10-250. Quench level in each sample is automatically calculated by exposing the sample to the Eu-152 external standard and calculation the SQPE value (spectral quench parameter of the external standard) from the generated Compton spectrum.

Tritium Standard Preparation: For tritium counting, water samples are mixed with the DIN based scintillation cocktail (Ultima Gold of Perkin Elmer make). It is known that water molecules of sample itself behave as chemical quencher therefore ten no. of quenched standards are prepared by adding increasing amount of tritium free water (1,2… 10 ml) and standard tritium activity (6360 DPM) to every vial containing 10 ml of ultima gold cocktail. Effect of change in volume (gross volume changes from 11 ml to 20 ml in plastic vial) on counting efficiency is insignificant therefore counting efficiency is considered almost constant with respect to volume variation of these prepared quenched standards. There is decrease in efficiency only due to increase in quenching by adding increasing amount of water volume. Lowest water volume (1 ml) will have least quench (maximum SQPE) and highest water volume (10 ml) will have maximum quench (least SQPE). Polyethylene vials were used because they are superior to glass vials due to low background count. Certified quenched standards of M/s Perkin Elmer were used for validation of quench curve generated from in-house prepared quenched standards.

The counting detection efficiency for 3H was thus calculated as follows:


SQPE (quench parameter) vs tritium efficiency data were plotted and illustrate in [Figure 1]. Very good correlation (R2= 0.99) was observed between SQPE (quench parameter) and efficiency of prepared quench standards. Second order polynomial functions was used to fit the efficiency and SQPE values obtained by experiment. Equation generated from calibration curves was used to find out efficiency unknown sample based on quench level and thus activity was determined. Activity results of known standards using above quench curve, which was obtained from in-house prepared quenched standards, are shown in [Table 1]. It was observed that all results of known standards are found with in + 5 % from actual activity. This validates that in-house prepared quenched standards can also be used in place of imported quenched standards for calibration of LSA to determine the tritium radioactivity in environmental water samples.{Figure 52}{Table 12}

Keywords: 3H, Efficiency, liquid scintillation analyser, quenching


Kessler MJ. The liquid scintillation analysis (LSA Manual). In: Science and Technology. Packard Instrument Company.L'Annunziata MF. Handbook of Radioactivity Analysis. 2nd ed. San Diego: Academic Press; 2003.

 Abstract - 51460: Development of automated delayed neutron coincidence system for special nuclear material detection

S. Chandrasekaran, S. Viswanathan, C. V. Srinivas, B. Venkatraman

Environmental Assessment Section, Environment Assessment Division, Safety, Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Special Nuclear Material (SNM) detection and quantification assumes importance in alpha waste from nuclear facilities. Delayed neutron counting method is the conventional method employed for this purpose. It is based on the measurement of spontaneous neutron emission of 240Pu. However, the presence of other actinides in the waste would complicate the counting which results into errors in waste assay. The present work uses coincidence technique in which it will be differentiating the neutron originating from 240Pu and other actinides. An optimized neutron counter design involves selection of neutron detector, array of detectors, moderating assembly and geometry. Monte Carlo based computer code is employed for design optimization. Helium filled gas proportional counter is chosen for neutron measurement because of its high sensitivity and higher thermal neutron cross section [5830 barns]. The detector assembly is made of High Density Poly Ethylene (HDPE) with the provision for 20 Nos. of 3He neutron proportional counters (50 cm length; 2.54 cm dia.) arranged in a circular pitch around the central cavity. It is cylindrical one with dimensions 30 cm diameter and 50 cm height. It is surrounded by HDPE cylinder which has 7.5 cm thick. The sample to be assayed is placed in the central well of the counter. Coincidence counting of time correlated fission neutrons is a powerful technique for distinguishing the fission neutrons from (α, n) and background neutrons. Rossi Alpha distribution obtained from the distribution of arrival times of the neutron pulses from the detector is the basis for obtaining the random (A) and real coincidences (R). The difference of (R+A) and (A) gives the real coincidences (R) and is proportional to the effective 240Pu mass.{Figure 53}{Figure 54}{Figure 55}

Keywords: Coincidence counting, Monte Carlo simulation, neutron, plutonium

 Abstract - 51474: Study of alpha radioactivity in beach sand samples using polymeric solid state nuclear track detector

S. S. Chavan, H. K. Bagla

Department of Nuclear and Radiochemistry, Kishinchand Chellaram College, Churchagte, Mumbai, Maharashtra, India

E-mail: [email protected], [email protected]

Different ecological, geological, and environmental formations like, rocks, soils, sand, plant life, aquatic forms and airborne materials are present in natural environment of earth. Pure radioactivity in different soil and beach sand samples arises from 238U, 232Th and natural 40K. The Polymeric detector such as Allyl Diglycol Polycarbonate (CR-39) as Solid-State Nuclear Track Detector has been employed in this work to study alpha radioactivity and alpha track detection from different Beach sand samples. For the existing investigation, sand samples had been collected from two distinctive seashores of southern Coastal India. CR-39 detector pieces have been exposed for various times with sand samples for alpha track detection and measurements. After exposure of detector pieces with sand samples, chemical etching used to be carried out by employing newly introduced 5% Tetra Ethyl Ammonium Bromide (TEAB) etchant at 60°C for 6 hrs. Track density (Td) and track diameter for collected sand samples were measured. Alpha radioactivity due to Radionuclides such as 238U & 232Th from sand samples were additionally measured by using EDXRF & ICP-MS. The introduction of new chemical etchant accurately improves uniformity, enhances track density, and tracks formation. Therefore, it can be concluded that TEAB is exceptionally high-quality chemical etchant for the alpha track revelation and detection. Experimental: Different sand samples were collected from different beaches from Mumbai to Kerala Black beach region of southern Coastal India. Collected samples were then dried in an oven at about 100°C for 3-6 H to remove moisture from it and then sieved by using 100 μm mesh size sieving filters to obtain uniform particle size of samples. Both samples were then packed in an airtight polythene container and stored for 2-3-weeks to attain radioactive equilibrium of alpha emitters and particles present in sand samples and their other decay elements. For the present investigation, the pieces of CR-39 detector with area 1X 1.5 cm2 was inserted in about 1 g of each sand sample and 10 mL of each sea water sample in different containers. After exposure, CR-39 detectors pieces were etched chemically with 6 M NaOH and newly introduced etchant with addition of 5% Tetraethyl ammonium bromide with 6 M NaOH at 60°C for different etching time. Observations and counting of obtained alpha tracks were done using optical microscope on 40X magnification. The average number of tracks per cm2 was obtained and from total number of tracks in addition of track density for all samples were measured. The Atomic Force, Microscope, Spinning Disc Confocal Microscope and Scanning Electron Microscope has been employed to study the surface of CR-39 detector. Track density (Td) and track diameter for sand samples were calculated. Track detection was studied from sea water filtrate sample, but accountable results were not obtained. 238U & 232Th presence by EDXRF and their content were also measured by ICP-MS.{Figure 56}

Keywords: Alpha, CR-39, Sand, SSNTD, TEAB


Chavan SS, Bagla HK. Measurements of alpha radioactivity in thermal power plant effluents employing CR-39 detector based improved alpha track detection method. J Environ Radioact 2021;233:106574.

 Abstract - 51489: Use of LaBr(Ce) scintillator for the estimation of 131I produced from the irradiation of 130Te at KAMINI

R. Akila, A. Dhanasekaran, R. Mathiyarasu, D. Ponraju

Health and Industrial Safety Division, SQRMG, IGCAR, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

The scintillation detector NaI(Tl), is widely used over the decades and finds application in many fields. In recent years, the use of LaBr(Ce) scintillators for various applications is explored. As compared to NaI(Tl), LaBr(Ce) have better scintillation property, resolution, light yield, decay time and density. LaBr(Ce) has 1.6 times light output and decay time is 10 times more than that of NaI(Tl). The resolution of LaBr(Ce) is 3% and that of NaI(Tl) is 7% of 137Cs 662 keV peak. The density of LaBr (Ce) is 5.06 g/cm3 and hence efficiency is also 1.5 times that of NaI(Tl). Because of these salient features, it has been recognized as a valuable alternate for NaI(Tl) in many applications. In this paper the application of LaBr(Ce) for the estimation of activity induced by neutron irradiation is discussed. The in-house production of 131I, to be used for calibration of stack effluent monitoring system, by irradiating 130Te in KAMINI reactor was carried out. The feasibility of use of LaBr(Ce) for the estimation of 131I activity produced by irradiation is explored and this activity is compared with that of NaI(Tl) based measurements. The detector BrilLanCe380 is 1“x1” crystals of Saint Gobain make with associated electronics of CAEN make gamma stream stand-alone, tube base MCA including high voltage power supply and charge sensitive preamplifier. The data obtained is analyzed using MC2 software. The energy and efficiency calibration of the LaBr(Ce) was done using 152Eu standard reference source. The tellurium oxide of mass 1g was irradiated in Pneumatic Sample Fast Transfer System (PFTS) irradiation location at KAMINI at a power of 1kWt for 5 minutes. The sample was retrieved and the dose rate on the sample was 200μSv/h. The sample was counted for 1000s immediately after irradiation with sample to detector distance being 30 cm. The second counting of the sample was also carried out after ~ 3 days delay for 131I for 20000s. The predominant peaks of 131Te, i.e., 149, 452 and 492 keV were only present. The peaks due to 131mTe and 131I are not visible as the counting time was only 1000s. [Figure 1] shows the spectrum acquired after a three-day delay by keeping the sample in contact with LaBr(Ce) detector. All the predominant peaks of 131Te, 131mTe and 131I are present. The efficiency for 364 keV photo peak regions corresponding to 131I is found to be 0.0237 cps/Bq and the decay corrected 131I activity is calculated to be 2096 Bq. Later the sample was analyzed in NaI (Tl) scintillator system for up to 20 days at different decay intervals. [Figure 2] shows the spectrum obtained in NaI(Tl). The back calculated 131I activity from the NaI(Tl) system at the time when the sample was counted in LaBr(Ce) system is 2236 Bq. The deviation in estimated activity between the two different systems is 6.05%. It is observed that LaBr(Ce) having superior features than that of NaI(Tl), can be used in the estimation of activity induced by neutron irradiation. The estimated activities of the scintillators are comparable.{Figure 57}{Figure 58}

Keywords: Efficiency, gamma spectrometry, induced activity, LaBr(Ce)


Bettiol M, Preziosi E, Borrazzo C, Polito C, Cinti MN, Pellegrini R, et al. Nucl Instrum Methods Phys Res A 2018;912:154-7.Saint Gobain Crystals, BrillianceTM380 Scintillation Material, Rev. 3; 2007.Milbrath BD, Choate BJ, Fast JE, Hensley WK, Kouzes RT, Schweppe JE. Nucl Instrum Methods Phys Res A 2007;572:774-84.

 Abstract - 51512: Rapid optimisation of BEGe detector geometry by establishing an empirical formula for full energy peak efficiency using Monte Carlo simulation and machine learning algorithm

Riya Dey1,2, S. Chinnaesakki1, M. R. Dhumale1, S. V. Bara1, K. D. Singh1, S. K. Jha1,2, M. S. Kulkarni1,2, S. Anand1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Monte Carlo (MC) based detector simulations are now widely adopted in gamma-ray spectrometry enabling quick estimation of full-energy peak efficiency (FEPE) values. The general practice that is followed globally is to optimise various parameters of the detector using multiple MC simulations till simulated FEPE at various gamma energies matches the experimental values since specifications provided by manufacturer are often very limited. This optimization method requires numerous simulations and changing each parameter with small incremental values can lead to long computational time. Hence, it is important to build up a method to quickly identify the initial guess values to optimise these parameters. To achieve this goal, the present study focused on exploring the dependency of FEPE for a Broad Energy High Purity Germanium (BEGe) type detector, on the geometry of the active region of the detector as well as on the incident gamma energy (E). FLUKA[1] MC-based code was used to model different detector geometries and corresponding FEPE values for point sources (60Co, 152Eu, 133Ba and 137Cs) placed axially at a source to the detector distance of 25 cm, were obtained. The “DETECT” card enables scoring (tally) of energy deposition on an event-by-event basis and thus net counts under the photo-peaks can be obtained and FEPE values can be calculated. These simulated FEPE values along with the detector geometry parameters (detector active volume and diameter to length ratio of the active region) and incident gamma energies were incorporated as an input to machine learning-based non-linear regression models. This input consists of FEPE values at 15 gamma energies ranging from 122 keV to 1.41 MeV for 16 geometrical configurations with active volume (Vac) ranging from 74 cc to 212 cc resulting 240 (= 16 x 15) sets of input variables. This geometry range covers all the commercially available BEGe detectors. The active volume of the detector was calculated as


where, Vde = total vol of all dead layers, and [INSIDE:2]; D and L are diameter and length of the detector respectively. The GEKKO optimization software[2] package was used for this purpose and a python-based code was prepared to find out an empirical formula for the prediction of FEPE values for point sources placed at 25 cm from the detector entrance window. Different nonlinear functions were tested for convergence and the functions were discarded if convergence was not achieved or if it yielded high sum of squared error (SSE) value. Following this approach, finally, a non-linear function having a form of equation 2, containing 12 parameters (a, b, c, d, e, f, g, h, k, l, p, q), yielded an SSE of 0.111 with maximum relative percentage deviation 4% in this energy and active volume range. Beyond this range, the validity of predicted FEPE values obtained from the empirical formula needs to be verified with experimental results.


This empirical relation was then used to carry out the optimisation of geometry of the over-square shaped p-type BEGe detector present in the lab. Instead of performing multiple simulations by changing the diameter and length, first, a separate python-based code was written where the empirical relation (Eq. (2)) was used to get a set of values of length and radius of the active region of the detector that can produce the intended relative FEPE. The relative deviation (RD) of empirically predicted FEPE from the experimental value has been plotted in [Figure 1] and it was found that for a length of 4.19 cm, a radius of 3.68 cm yielded an FEPE very close to the experimental value, where manufacturer quoted nominal value of radius was 3.95 cm. With this geometry (from Eq. (2)), FEPE values were estimated from FLUKA simulation and it was found that the simulated FEPE values were well matching with the experimental value (maximum deviation was 4%). The main advantage of this empirical formula is that instead of multiple MC simulations, only one (or few) simulation will be required to perform the detector optimization.{Figure 59}

Keywords: BEGe detector, FEPE, FLUKA, machine learning, nonlinear regression


Ahdida C, et al. Front Phy 2022;9:788253.Beal LD, et al. Processes 2018;6:106.

 Abstract - 51515: NaI(Tl) based on-line stack monitor to measure 133Xe in presence of 41Ar

C. N. Sunil, V. Narasimhanath, Utpalbera, M. Prashant, Mayank, T. Raghunath, S. Kumaravel, V. Ramakrishna, G. Ganesh, M. S. Kulkarni

Health Physics Division, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Effluents from nuclear industry are ardently controlled and monitored before they are discharged to the environment. This ensures releases are within the limits set by regulatory body and also minimises the radiation dose to all stake holders in the environment. Among various radioactive gases produced from a nuclear reactor such as fission products (FPs) and activation products (APs), those having long half-life and high energy poses higher radiation dose risk in public domain. The most significant among them are 133Xe (T1/2 : 5.3d, Eγ: 81keV)and 41Ar (T1/2 : 110m, Eγ: 1293 keV). An on-line gaseous effluent monitor for this purpose was designed, fabricated and calibrated in-house as given below. Gaseous effluents from a nuclear reactor are first passed through [Figure 1] glass fibre filter paper for trapping particulates and its activity is normally monitored by plastic scintillator. Further effluents pass through activated charcoal for trapping 131I which is measured by NaI(Tl) detector. Finally, the residual air with 41Ar and FPNG is fed to a gas chamber with NaI(Tl) as the counter. 133Xe is estimated in the presence of 41Ar by 3 upper level discriminators[1] set at energies 60keV, 110keV and 1MeV as shown in the block diagram [Figure 2]. Energy calibration of NaI(Tl) used in gas monitor was done by 133Ba, 137Cs, 22Na disc sources. Keeping one of the above sources at a fixed position, cumulative and differential spectrum [Figure 3] was obtained from which calibration factor was estimated. Keeping 22Na (1.27MeV) source at various radial and azimuthal positions inside the sampling chamber, effective efficiency of 0.6% for (41Ar) the whole air volume was measured. Compton contribution from high energy photons to 133Xe channel was observed to be 0.35% using 60Co. FLUKA simulations predicted 0.67% efficiency for 41Ar and 0.90% for Xe133. Efficiencies for 41Arand 133Xe were experimentally validated[2] by counting effluent samples from MAPS and comparing with calibrated[3] HPGe detector. Calculated MDAs for 41Ar and 133Xe are tabulated below. High MDA for 133Xe in presence of 41Ar is due to the Compton background created by 1293 keV gamma in the lower channel(60-110keV) meant for counting 133Xe.{Figure 60}{Figure 61}{Figure 62}{Table 13}

Keywords: Ar41, FPNG, NaI (Tl)


Knoll GF. Radiation Detection and Measurement. New York: John Wiley & Sons; 2000.Petr K, Dryak P. Appl Radiat Isot 2008;66:796-8.Feng X, et al. Appl Radiat Isot 2015;97:8-11.

 Abstract - 51516: Development of remote real time radiological data acquisition system

Govinda Mukherjee1, Anand Raman2, S Murali, Probal Chaudhury

EPRS, RSSD-BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

With objective of exploring the emergent technologies of accessing and controlling radiation monitoring systems remotely, a System capable of connecting with any web network using IP address has been developed. Web page controls reduce the hardware components hence enhancing the compactness and facilitates low current consumption. This system with dimensions that of a standard brick is referred to as Brick detector system. Brick detector is a regular AGM (detector zp1201) with enhanced capabilities. The system use ESP32 with micro-python to accomplish the objectives.

The system has two modes of working:

TFT ModeWi-Fi Mode

TFT Mode:

Display the dose rate in mR/h.Display the alarm set value in mR/h.

No Provision to set the alarm in TFT Mode. Whenever the dose rate >= set limit, the system alerts by audio visual indications.

Wi-Fi Mode:

Display the dose rate in mR/h.Display the alarm with the option to set it.Display's last 10 alarm history with date-time and dose level.

In this mode, the device acts as access point with Username: [xxxx_x_AP] and Password: [xxxxxxxxx]. Wireless range is limited to few meters only. One can extend the range using external Wi-Fi router or LAN cable. Once connected with PC/Smart TV/Mobile/Tab etc, where web browser is accessible, by entering the static IP: 192.168.x.x/Gamma.html one can get the device hosted page. Data from the device to the web browser is pipelined using Web Socket. In case of any breakage in the connection, the page controls are in-accessible and in the right up corner of the page instead of “–connected—” message it will display “–disconnected—“. In case of alarm, only the initial reading will register in the table. Next alarm data will append only after reset or dose rate comes below alarm level. The device casing is also designed in-house using 3D printing technique.

Brick Detector is calibrated from background to 2 R/h and field tested in AGRO facility at Gamma Fields BARC, Mumbai. This work was supported by Smt Wani, Shri. Vishal Kharvi from RSSD, the staff at AGRO facility and colleagues from RSSD Workshop.{Figure 63}{Figure 64}{Figure 65}

Keywords: AGM, embedded python, ESP32, remote access, Web Server, Web Socket, Wi Fi display


Tollervey NH. Programming with Micro Python: Embedded Programming with Microcontrollers and Python. O'Reilly; 2017.Sever Spanulescu. ESP32 Programming for the Internet of Things. 2nd ed. HTML, JavaScript, MQTT and WebSocket Solutions; 2020.

 Abstract - 51527: Evaluation of DWL values for radiation portal monitor using Monte-Carlo simulation

Vaibhav Bhardwaj, Ashish Arvind, Amit Bhatnagar, Joy Chakraborty, M. K. Sureshkumar

Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Personnel contamination check and clearance has got importance in several nuclear facilities i.e., reprocessing plant, waste immobilisation facility etc. as various process involved in such type of facilities deal with solution/dispersible form of radioisotopes leading to more chances of contamination. After completion of job, it becomes significant to check level of contamination while crossing various plant zones boundaries. At the final exit point of radiation facilities, radiation portal monitor (RPM) becomes essential requirement for personnel check to ensure that personnel leaving the plant are contamination free or contamination level are below derived working levels (DWL) set by AERB for personal clothes, shoes or on skin. Sensitivity of various types of portal monitor and their minimum detectable activity (MDA) plays an important role to ensure that regulatory guidelines are achievable and are being strictly followed. For this simulation, Monte-Carlo package FLUKA (FLUktuierendeKAskade) was used to calculate the efficiency of the RPM. The geometry of PMS-4021 Doorway monitor designed by ECIL (Electronics Corporation of India Limited) was simulated for this study. It consists of two single crystal PVT detectors of dimensions (4 cm × 28 cm × 198 cm). The overall dimension of portal monitor is 121 cm × 45 cm × 234 cm and the walking area inside portal monitor is 75 cm × 45 cm × 210 cm. For the simplicity of the simulation, an adult Bottle Mannequin Absorber (BOMAB) Phantom was simulated at the centre of the portal monitor. The Critical Level (LC) was established using a type I (false positive) error probability and is calculated using eq-1.


Where Tg is the personnel counting time in seconds, Tb is the background counting time in seconds, Rb is the background count rate in cps and zα corresponds to the false positive error probability. The lower limit of detection (LD) was established using a type II (false negative) error probability and is calculated using eq-2 and MDA can be estimated by ratio of Ld and efficiency.


Counting efficiency variation along vertical and horizontal direction using a point source of 137Cs is shown in [Figure 1]. It can be seen that the efficiency is minimum at the centre and increases in either direction. The vertical and horizontal efficiency of RPM was joined to form a 2-D contour shown in [Figure 1]b. A 137Cs source was placed at five different body positions (chest, head, lower arm, lower leg and abdomen) of the BOMAB phantom. Using DETECT card in FLUKA, efficiency was determined and was used to calculate MDA. Using these MDA values, surface contamination for skin averaged over 300 cm2 was calculated as shown in [Table 1]. The background count rate was 300 cps, zα was 1.645, zβ was 3.09 and background counting time was 50s.

Polar variation of efficiency for 137Cs along the center of the chest was also calculated and is shown in [Figure 2]. The source was kept adhered to the surface of BOMAB phantom. The DWL value set by AERB for skin contamination is 1.5 Bq/cm2 for beta contamination. Considering these parameters, it is inferred from the above table that the calculated values of surface contamination for 137Cs at various body positions are well below the DWL values set by AERB. Further studies with various beta sources will be carried out in the future.{Figure 66}{Figure 67}{Table 14}

Keywords: Counting efficiency, derived working levels, minimum detectable activity, radiation portal monitor


Currie LA. Anal Chem 1968;40:587-93.Battistoni G, et al. Ann Nucl Energy 2015;82:10-8.NF/SM/O-2 (Rev4). Radiation Protection for Nuclear Facilities. Mumbai: AERB; 2005.

 Abstract - 51531: Development of semi-automatic high power blue LED chip based optical bleaching setup for OSL dosimetry

Mukesh Uke, A. K. Singh, D. R. Mishra

Radiation Physics and Advisory Division, Bhabha Atomic Research Centre, Trombay, Mumbai, Maharashtra, India

E-mail: [email protected]

The semi-automatic optical bleaching setup using high power blue LED chip has essential application in Optically Stimulated Luminescence (OSL) based radiation dosimetry.[1] This setup has been used for the resetting of residual OSL signal from OSLD discs to the background level. This system is designed for automatic turns ON period of 15 minutes to complete the five bleaching cycles. The LED off periods and cooling fans are incorporated to avoid overheating of the LED chips to increase the working lifetime. The manual emergency stop is also provided for shutdown of system. An additional thermal safety is incorporated with temperature sensor based power cut off method. Total output power of the system is around 500 W with maximum light intensity of 20000 lux at the center areas. The bleaching period has been optimized for the indigenously developed OSL discs. The semi-automatic optical bleaching setup [Figure 1] consists of ten ultra-bright high power blue LED chip (50W each), power supply modules (35V, 10A), 2 mm thick sample holding glass tray, LM35 temperature sensor. The control unit consists of a heat sink temperature sensing, pulse stretcher, LED drivers and single 7-segment display. Two aluminum plates also act as a heat sink with ten high power blue LED chip have been mounted. An arrangement at the center of the setup has been made for the removable of sample holding glass tray to position the samples easily. Both metal plates are arranged at an equal distance from the top and bottom of the glass tray to keep uniform LED intensity at the sample position. The setup will be ON uniformly for 1 minute of bleaching and OFF uniformly for 2 minutes of LED cooling during each cycle. A display will be automatically updated 1 to 5 numbers after each bleaching cycle, and then seven segment digit will be blinking same number of times till the next cycle. If the LED heat sink temperature reached at 800C, power supply module will be shut down automatically. Polytetrafluoroethylene (PTFE) based Al2O3:C discs of 10 mm diameter,[2] containing phosphor to PTFE in the ratio of 1:3, were used for all the experiments. The OSL measurements were carried out on an automatic RISO TL–OSL-DA-15 reader system. The bleaching light intensity uniformity was measured at 23 points on the glass tray using light meter LX101A. The maximum light intensity of 20000 lux was found to be around the center areas (15 cm x 15 cm). Ten dosimeter discs (samples) of similar sensitivity were selected for the bleaching experiment. The continuous wave (CW) OSL of the discs were recorded for a background reference. Irradiated dosimeters of 100mGy dose were bleached for different bleaching period. The CW-OSL intensity vs. bleaching period was plotted as shown in [Figure 2]. It is found that CW-OSL signal decreased to ∼0.3 % of initial signal after 10 minutes of bleaching. The OSL discs were found to get completely reset within 30 minutes for the dose of 100mGy. The OSL phosphor has a slow decay component which takes relatively more time to bleach out. Bleaching time varies with the dose received by the OSL discs. The Al2O3:C discs exposed to dose of 10mGy can be reset to background level within 5 minutes for their reuse [Table 1].{Figure 68}{Figure 69}{Table 15}


Authors thank to Dr. D. K. Aswal, Director, HS&EG and Dr. B. K. Sapra, Head, RP&AD, BARC, Mumbai for their support during this study.


Yukihara EG, McKeever SW. Optically Stimulated Luminescence Fundamentals and Applications. UK: John Wiely & Sons Ltd; 2011.Rawat NS, Dhabekar B, Kulkarni MS, Muthe KP, Mishra DR, Soni A, et al. Optimization of CW-OSL parameters for improved dose detection threshold in Al2O3: C. Radiat Meas 2014;71:212-6.

 Abstract - 51565: Development of a portable gamma radiation spectrometer using indigenously grown CdZnTe single crystals

E. P. Amaladass1,2, P. Vijayakumar1, O. K. Sheela, R. M. Sarguna1, K. Ganesan1,2, Varsha Roy1, S. Ganesamoorthy 1,2, N. V. Chandra Shekar1,2

1Materials Science Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected], [email protected]

The demand for high-performance X-ray and gamma-ray detectors has been continuously growing due to their application in various fields such as national security, medical imaging, and environmental protection, etc. High purity germanium (HpGe) based detectors with high energy resolution are widely used for radiation monitoring. However, one of the significant drawbacks is the requirement of cryogens for operation. CdZnTe (CZT) detectors with wide bandgap (~1.57 eV), high atomic number, high resistivity (~1010 Ω-cm), and high electron mobility-lifetime product (mt ~ 5×10-3 cm2/V) are ideal candidate for room temperature radiation detection.[1] However, the growth of high-quality CZT single crystals involves several challenges due to inherent material properties. With optimized growth parameters CZT single crystals were successfully grown using an indigenously designed and fabricated travelling heater method (THM).[2] The wafers from the crystal boule were cut, lapped, and polished. Planar gold electrodes were deposited by an electroless process. The elements with electrodes were housed in an aluminium casing with a BNC connector and attached to an in-house developed portable gamma spectrometer, as shown in [Figure 1]. The photographs of grown crystal and processed elements with Au electrodes are shown in the inset of [Figure 1]. The photo peak of 241Am and 133Ba at 59.5 keV and 81 keV were resolved with an energy resolution of 8-9 %. The planar detectors made from CdZnTe suffer from the issue of poor hole transport properties. The induced charge is dependent on the interaction depth in the crystal. This limitation of the planar detector structure necessitates the usage of other electrode geometries that will compensate for the poor hole transport properties. Detector with a quasi-hemispherical (QH) electrode was prepared by electroding the five sides of the CdZnTe crystal of 10x10x5 mm3 in dimension. The photo peak of 137Cs was resolved at 14% resolution, as shown in [Figure 1]. Further developments in growing high-quality single crystal, device fabrication process, and signal analysis electronics are in progress to improve the sensitivity and resolution.{Figure 70}

Keywords: CdZnTe single crystals, gamma detector


Alam MD, Nasim SS, Hasan S. Prog Nucl Energy 2021;140:103918.Vijayakumar P, Amaladass EP, et al. AIP Conf Proc 2020;2265:030414.

 Abstract - 52199: Impact of supporting substrate for natLi(1H,n) system on emission neutron yields and dose equivalent at proton energies between 8-20 MeV

Sabyasachi Paul1, G. S. Sahoo1, S. P. Tripathy1,2, A. A. Shanbhag1, S. C. Sharma3, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, 3Nuclear Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

The natLi(1H,n) reaction emits quasi-monoenergetic neutrons from epithermal to few hundreds of MeV and the application domain extends from cancer treatment to calibration of high energy neutron detectors. In these applications, most of the times the Li target is not self-supported and requires a thick supporting substrate. It can be either high or low Z-materials based on the applications. For high 1H current requirements, metallic options (Au, Ta etc.) are preferred over low Z- (C, Al etc.) supporting substrates due to better thermal conductivity. In the present study, a comparison of the emitted neutrons from the natLi target (8 mg cm-2) with two separate supporting substrates, natTa and natC were carried out at incident 1H energies between 8-20 MeV. The experimental measurements were performed at 6 m facility of BARC-TIFR Pelletron-LINAC facility. The neutron yields at 0°, 90° directions with respect to incident 1H beam with natTa, natC targets and thin natLi foil were presented in [Table 1]. The natLi(1H,n) neutron yields were estimated by subtracting yields from separate measurements carried out with/without natLi foils on the supporting substrates. The neutron yields ratio from natLi target and supporting substrate were estimated to be ~1 for natC whereas ~10-2 for natTa. Experimentally the neutron yields from natLi is expected to be independent of the supporting substrate and reduction of the emission neutrons originated from supporting substrate is desired for precise identification of emission neutron signatures from natLi target. The variation in neutron yield from supporting substrates depends on constituent isotopes and possible neutron emission reactions. 13C (1.07%) is the major source of neutrons in case of natC,[1] whereas for natTa; 181Ta (99.98%) with multiple emission neutron channels 181Ta(1H,xn; x=1-3) become dominant contributor at proton energies between 8-20 MeV.[2] This reduces the neutron background by two orders of magnitude for natC with improved quantification and better spectral features of natLi(1H,n) emission neutrons. The ambient dose equivalent, H*(10) also reduced significantly with natC target. [Figure 1] (same legends used for both plots) indicated that with natC support, H*(10) reduced sharply (~60 times at 0° & ~40 times at 90°) in the entire 1H energy range. Study concludes that natC is a better supporting substrate compared to natTa for natLi(1H,n) reaction studies.{Figure 71}{Table 16}

Keywords: Ambient dose equivalent, dose minimisation, thick target neutron yield


Paul S, et al. Nucl Instrum Methods A 2022;957:163432.Paul S, et al. Nucl Instrum Methods A 2020;1034:166767.

 Abstract - 52204: Development of plastic scintillator detector based radiation monitor with source categorisation capability

Vaishali M. Thakur, Amit Jain, Probal Chaudhury

Radiation Safety System Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: The radiation protection and surveillance programme in nuclear industries require intelligent and sensitive radiation monitors, both installed and portable, for the environmental radiation monitoring. The radiation monitor should have high sensitivity and ability to discriminate Naturally Occurring Radioactive Material (NORM) against Industrial, medical use radioisotopes and Special Nuclear Material (SNM). This paper describes the development of plastic scintillator detector based radiation monitor (PSRM) using limited spectroscopy analysis for gamma ray measurements.

Materials and Methods: An Intelligent PSRM has been developed using a 50 cm long and 5.1 cm diameter detector interfaced to a Photomultiplier Tube (PMT). Detector PMT signal is fed to the pulse processing electronics through Pre-Amplifier. This signal is analysed by limited spectrometer for data acquisition time of 10 S and in built embedded software. The result is communicated to user on a display and serial RS-232 port to PC. This intelligent plastic scintillator detector [Figure 1] based radiation monitor operates on 12 V DC power supply. Methodology: The PSRM processes the detector signal and segregates it into 5 Energy Windows (EW) based on pulse height.[1] The energy calibration of the system is 260 keV/V The five EWs are set as 50-260 keV, 261-400 keV, 401-800 keV, 801-1500 keV and 1501-2600 keV. Gross counts and applying calibration factor dose rate is calculated. If gross counts exceed the threshold counts then the ratios of counts of low EW to high EW (R1 to R4) and percentage energy window counts contribution (PEWCC) from each window (EWp1 to EWp5) are calculated to discriminate NORM against radionuclides like 137Cs, SNM, Thorium and 60Co. At ~ 0.1 μSv/h background field typical values of ratios with 10% statistical variation are given in [Table 1]. The observations with different sources leads to conclusion that PECC from EWp1 is greater than 90% for low energy sources, 80-90% for 177Lu (208 keV) and 70-80% for medium energy sources 133Ba and 131I. Once the PECC from EWp1 is < 50 % and EWp2 (14%), EWp3 (20-30%) are higher than background PECC concludes the presence of 137Cs with ratios R1 and R2 less than background ratios. PECC from EWp4 is less than background for Thorium (12-16 %) and for 60Co is (22-32 %) higher than that of background. But PECC from EWp5 is greater than that of background for both thorium and 60Co. Ratios (R1 to R4) for gamma emitting radionuclides[2] with energy greater than 1 MeV are constant irrespective of the strength of the source. The ratios are used for the confirmation of presence of the thorium, 60Co and uranium ore having all R1 to R3 ratios lower than their respective background ratios. The ratio R4 discriminate it among Thorium (6-9), 60Co (4-7) and uranium ore (10-14) as compared to background ratio in the range 12-14.

Results and Discussion: The sensitivity of the PSRM for 137Cs is 1.5 CPS/nSv/hr with dose rate range of 0.25 to 100 μSv/hr. The monitor's energy response is within ±10%. It categorises the radionuclides as low (SNM, Radiopharmaceutical radionuclides (except 131I), medium (133Ba, 131I and 137Cs) high (60Co, Thorium) energy and NORM (Uranium ore, Sand, Soil, Granite and Fertilizer). It detects presence of gamma source with more than 90 % confidence above 250 nSv/h dose rate. The measured dose rate from the system was compared with commercially available Thermo scientific survey meter and is within ± 20%.

Keywords: Energy window, limited spectroscopy, naturally occurring radioactive material, plastic scintillator detector, special nuclear material


Ely JH, et al. Discrimination of naturally occurring radioactive material in plastic scintillator material. IEEE Trans Nucl Sci 2004;51:5.Anderson et al. Dis- criminating nuclear threats from benign sources in gamma-ray spectra using a spectral comparison ratio method. J Radioanal Nucl Chem 2008;276:713-8.

 Abstract - 52211: Simulation of peak tailing in alpha spectrometry based continuous air monitor with collimator using Geant4

Vivek Kaushik, D. P. Rath, A. Bhaktivinayagam , S. Anand, R. V. Kolekar

Health Physics Division, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Continuous air monitor (CAMs) based on alpha spectrometry is widely used in fuel fabrication facilities to detect very low level of airborne activity as required by regulatory authority. Monitoring of air activity is most challenging as Radon daughters 218Po and 214Po, and Thoron daughters 216Po, 212Bi and 212Po emit alpha particles, which interfere with the other long lived alpha emitting radionuclide detection in the region of interest. Peak tailing is the common problem in such open-air alpha spectrometry as higher energy alpha emitters interfere with radionuclide of concern. This tailing is because of angled alpha radiation entering the detector. Use of collimator is common solution to address such problem.[1] Present study have been done to simulate the drop in efficiency and tailing using Geant4 toolkit for the simulation of the passage of particle through matter.

Methodology: Geant4 is a platform for the simulation of passage of particle through matter using Monte Carlo methods.[3] Various classes of Geant4 are used to define geometry, source term and scoring. A passivated implanted planar silicon (PIPS) detector with active area of 450 mm2 , and a co-axial stainless steel collimator with fins length 20 mm, height 9 mm, and thickness 1 mm with a ring having 33 mm inner and 45 mm outer diameter is simulated in air, see [Figure 1]. In-build material definitions of Geant4 are used. Alpha particles with energies of 5.45 MeV (29%) and 5.49 MeV (71%) are used uniformly in a circular shape to simulate the filter paper containing radionuclide. Particle momentum is uniform in 2π geometry towards detector to avoid undesirable computing. Result and Discussion: Energy deposited in detector per event is logged in a branch-tuple of a root file using G4AnalysisManager. Root[4] is a framework for data processing/analysis. Logged data of both the cases, with and without collimator is plotted as line-histogram. In [Figure 2], the red line shows the energy deposition without collimator and blue line, with collimator. The mean energy deposited in detector is increased from 3.76±0.56 MeV to 4.04±0.18 due to collimator. The tailing has reduced from 52% to 14% in later case. Similarly, the efficiency has also reduced to one-third of the initial value. The resolution for both cases are found to be 300KeV and 200KeV. Conclusion: The alpha peak tailing is significantly reduced to 38%, compromising the efficiency of the detector. The FWHM has reduced to 67 % due to collimator. The simulation results are in consistency with the experimental observations (Rath, 2017).[1]{Figure 72}{Figure 73}{Figure 74}{Table 17}

Keywords: Alpha-spectrometry, Geant4, peak-tailing


Rath DP, Ashokkumar P, Verma OP, et al. Indigenous Development of Plutonium-in-Air Monitor NUCAR-2017; 2017.Mirion. Application Note: The Continuous Air Monitoring (CAM) PIPS Detector Properties and Application; 2011.Agostinelli S, Allison J, Amako K, et al. Nucl Instrum Meth Phys A 203;250-303. [Doi: 10.1016/S0168-9002(03)01368-8].Rene B, Rademakers F. Nucl Instrum Meth Phys Res A 1997;81-6. Available from: https://root.cer.

 Abstract - 52241: A rapid and robust technique for the measurement of 238 Pu in nuclear fuel samples

Sumana Paul, K. Sasi Bhushan and Preeti G. Goswami

Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Precise and accurate determination of the isotope 238Pu in Pu samples is highly crucial for several applications like quality control of nuclear fuels, nuclear material accounting, nuclear forensics, development of reactor physics codes, standardization of 238Pu as a spike for isotope dilution alpha spectrometry, identifying the source of Pu contamination in the environment etc.[1] The abundance of 238Pu in an irradiated uranium or uranium oxide fuel may vary from 0.01–5%,[2] depending on the fuel burn-up and various other parameters characteristic to the type of the nuclear reactor. Thermal ionization mass spectrometry is one of the most widely used techniques for precise and accurate determination of Pu isotopes.[3] However, it is not feasible to measure 238Pu with sufficient precision and accuracy by TIMS due to the isobaric interference of 238U, particularly in samples like nuclear fuel dissolver solution where the molar ratio of plutonium to uranium may vary from 200 to 1000.

In the present work, we have developed a robust yet simple and rapid method for the determination 238Pu in nuclear samples, containing large excess of 238U. A supported liquid membrane ([email protected]) was prepared by physical immobilization of the liquid extractant n,n′-dioctyl-α-hydroxyacetamide (DOHA) within the pores of commercially available, microporous polypropylene (PP) membrane. DOHA was dissolved in a room temperature ionic liquid (RTIL) viz. 1-hexyl-3-methylimidazolium bis(trifluoromethylsulfonyl)imide – the use of RTIL instead of the conventional dodecane solvent not only makes the process greener but also improves the separation factor for Pu(IV) and increases the equilibrium uptake capacity significantly. It can be seen form [Figure 1] that in 0.5 to 4 M HNO3, sorption of Pu(IV) by [email protected] membrane is significantly higher than Am(III) and U(VI). [email protected] was employed for selective pre-concentration of Pu from a variety of nuclear samples that contain large excess of uranium (Pu:U molar ratio ranging from 25 to 1000). In the next step, Pu pre-concentrated in [email protected] was stripped back into solution using hydroxylamine hydrochloride and Pu isotopic composition, including the atom% of 238Pu, in the sample was determined by TIMS. [Figure 2] gives a comparison of the measurement of 238Pu in a mixture of NIST SRM-947 Pu isotopic standard and natural uranium (Pu:U = 1:1000). It can be concluded from [Figure 2] that if the vaporization filament current is maintained at 2.4 A or higher, the measured 238Pu/239Pu atom ratio matches with the certified value with better than 0.5% accuracy. The method developed in the present work offered the analysis of 238Pu, in the presence of excess 238U, with a precision and accuracy better than alpha spectrometry as well as the conventional TIMS technique involving ion-exchange or extraction chromatographic resins for the separation of Pu. Also, Pu partitioning using [email protected] membrane offers several additional advantages compared to the traditional method e.g. minimum sample handling, fast separation and less personnel exposure.{Figure 75}{Figure 76}

Keywords: 238Pu, isotopic composition, thermal ionization mass spectrometry


Alamelu D, et al. Radiochim Acta 2005;93:259-63.Tiong LY, Tan S. J Radioanal Nucl Chem 2019;322:399-406.Paul S, et al. Anal Chem 2019;91:14383-91.

 Abstract - 52283: Influence of dust load on optimization of filter change cycle for alpha air monitors

Gopal P. Verma1, M. K. Sureshkumar1, S. K. Jha1,2, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Measurement of airborne alpha activity in actinides handling facility is an important aspect to achieve effective radiation protection. Alpha-particle continuous air monitors usually employees a PIPS detector for online acquisition of the alpha spectrum of the aerosol deposited on the filter paper.[1] As the instrument use a custom made algorithm to quantify the long lived isotopes in presence of the short lived natural alpha emitters from the acquired composite alpha spectrum, deterioration of the energy spectrum of the sample affects the sensitivity of the instrument. These instrument are usually designed to operate in clean laboratory environment, but our requirements also include operation of them in relatively dusty plant environments. In such conditions, it is mandatory to understand how-long the monitor could be operated without replacing the filter paper, before the dust load deteriorates the energy spectrum to cause loss of sensitivity and specificity for the radionuclide of interest. In this work, the SmartCAM (Lab Impex Systems) has been operated for different time interval in an actinide handling facility to achieve the optimum filter changing cycle. The present work discusses about the filter change criteria based upon the dust loading on the filters.

Materials and Methods: The continuous alpha in air monitors (Smart CAM) is placed and operated in an actinide handling facility. Ambient air is drawn through the filter using a vacuum pump at a flow rate of 50 lpm. The instrument is equipped with card mounted Glass-fiber filter papers of 5 cm diameter for collection of particulates. The activity deposited on the filter is measured by the two solid state PIPS detectors placed directly above the sample filter. It was operated for different operating hours to study the accumulated dust load on the filter papers. Estimation of accumulated dust load on the filter papers used for different time interval has been carried out. The used filters were also counted after sufficient delay in a ZnS (Ag) based counting system to confirm the presence of any long-lived isotopes. A series of such experiments have been carried out to know the dust load pattern on the filters and their interference on detection of manmade actinides.Results and Discussion: [Figure 1] shows a typical card mounted filter paper with dust load over it after field experiment. The amount of accumulated dust load on the filter paper at various operating interval in a typical experiment is given in [Table 1]. The accumulation of dust load on filters gradually increases with higher operational period of the SmartCAM. Thin-layer deposits of dust did not affect the alpha-particle energy resolution significantly.[2] It has been reported that serious reduction in energy resolution for such instruments results from dust load on filter exceeding 0.4 mg cm-2.[3] Thus, from the present set of experiments it is demonstrated that the SmartCAM could be operated up to 24 hrs without replacement of the filter paper. Increase in frequency of false alarm was observed for operation of the instrument beyond 24 hrs. without filter replacement, due alpha energy degradation of the Radon daughter products by the accumulated dust.{Figure 77}{Table 18}

Keywords: Airborne, dust loade, glass-fiber, SmartCAM


Maiello ML, Hoover MD. Radioactive Air Sampling Methods. USA: CRC Press, Taylor & Francis Group; 2011.Huang S, et al. Health Phys 2002;83:884-91.Stevens DC, Toureau AE. AERE-R-4249. UKAEA, Harwell, U.K; 1963.

 Abstract - 52339: Determination of working level of radon using three counts method and its comparison with WLM in atmosphere

Dibyendu Rana1, V. N. Jha1, R. L. Patnaik1, M. K. Singh1, S. K. Jha1,2, M. S. Kulkarni1,3

1Health Physics Division, Bhabha Atomic Research Centre, Departments of 2Chemical Sciences and 3Physical Sciences, Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

In view of the low activity concentration of radon and progeny in outdoor atmosphere, the daughter concentration as Working Level (WL) can be estimated using the Working Level Meter for extended sampling durations. For the three arbitrary intervals counting of the filter paper for gross α activity the WL in outdoor atmosphere has been estimated for a larger sampling duration. Present gross α activity based manual counting technique has been compared with the WL estimated using Si semiconductor based Working Level Meter. The results of the two estimates have been found in good agreement.

Materials and Methods: Working Level Meter (TN-WL-02) has been used to determine radon progeny WL in outdoor atmosphere (1 m above the ground). The working level can be determined by using the relation


Where, C is total counts in T hours and Cf is calibration factor ( Cf = 6.2 CPH/mWL, CPH is counts per hour) .For Working Level Meter, sample has been collected for 2.5 hours and the counts were converted into mWL using the above conversion factor. After a delay of td = 152 s the filter paper was counted (three count method) for 30 minutes at 5 minutes interval using ZnS (Ag) based alpha counter.

Three Counts Method: In view of the abundance of radon (222Rn) , the progeny (Po218, Pb214, Bi214) are presumed to be the major portion of short lived radionuclide present in natural atmosphere. During the counting, the effects of long lived radio-nuclides are considered insignificant.

q1, q2, q3 concentration of Po218, Pb214, Bi214 respectively in nuclei.m-3 and Xtjti integrated count in between ti and tj obey the following equation


In equation (2), ε is the efficiency of counter (in fraction) and V is the flow rate. All hij (i,j=1,2,3) depends on sample collection time ts, decay constant of radon progenies λ1, λ2 , λ3 and counting interval ti and tj.

Working level = (Xt2t1(13.7h11+7.69h21+7.69h31) +Xt3t2(13.7h12+7.69h22+7.69h32)+Xt3t2(13.7h31+7.69h32 +7.69h33)/(1.3×105 εV) in mWL (3)

With uncertainty,

σWL=(1.3×105 εV)-1 σXt2t1(13.7h11+7.69h21+7.69h31) +σXt3t2(13.7h12+7.69h22+7.69h32)+σXt3t2(13.7h31 + 7.69h32+7.69h33)0.5 in mWL (4)

Results: For 2.5 hours of collection time, total integrated counts in working level meter was 34 amounting to 2.19 mWL using the mentioned Cf .Using equation (3) , equation(4) and observed results from [Table 1], the estimated working level is provided in [Table 2]. The best-estimated working level from the above table is 2.05 mWL obtained by using WL = [INSIDE:3]. The uncertainty in best estimated value is 0.65 mWL obtained from, [INSIDE:4]. In the formulation of best estimation and uncertainty, [INSIDE:5].

Conclusion: The mWL estimated using simple three count methods for extended sampling period (2.5 h) resembles well with the alpha spectrometric based Working Level meter. The said technique can effectively be used for the outdoor atmosphere or areas having low level of radon / radon progeny. {Table 19}{Table 20}

Keywords: Radon progeny, three counts method, working level


Nazaroff WW. Optimizing the total-alpha three-count technique for measuring concentrations of radon progeny in residences. 1983.

 Abstract - 52442: Development of nonlinear square fitting based radioactive source localization algorithm in complex environment

Ashutosh Gupta, R. M. Suresh Babu1, Jis Romal Jose, M. K. Sharma, Probal Chaudhury

Radiation Safety Systems Division, Bhabha Atomic Research Centre, 1Health Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: In the nuclear industry and related applications, a problem of misplaced or lost gamma-ray sources is considered as a cardinal concern and hence, it necessitates quick source search and locate operation. Multiple attempts have been made to automatically locate a point source, but most of them are limited to the scenarios having no shield between source and detector.[1],[2] Statistical radiological data variation, inaccurate GPS positional information and presence of variable shields between the source and detector complicates auto locating the source. This paper investigates the feasibility and testing of Non-Linear Square (NLS) fitting based algorithm to auto locate the orphan point source in presence of environmental shields such as walls, trees, human etc.

Materials and Methods: Counts registered per second in a radiation monitor at a position (xi, yi, zi) due to a point radioactive source of strength S at (xs, ys, zs) can be written as:


Where μ represents the linear attenuation coefficient, B represents the buildup factor and [INSIDE:6]), distance between source and detector. A source locate algorithm has been developed in 'R' language based on Non-linear Least Square (NLS) fitting of Equation 1. Radiological measurement data synchronized with detector positional information referred as data points are fed to the algorithm. The algorithm divides the data into batches and NLS fitting is applied to each batch. Data points influenced by environmental shield such as trees, building etc are treated as outliers. These outliers are removed from the batch and NLS fitting is again carried out with the remaining batch. This process is repeated till no more outliers are present. Source location is estimated from each batch and the estimates with Euclidean distance lesser than the threshold are treated as single estimate. Hence, centroid of such source locations is considered as single estimate. Thus, a numbers of centroids are generated as the output of NLS fitting. Since estimates obtained from the batch having majority of the data points influenced by environmental shield lead to distant source location with higher activity than the actual source activity, estimated source with least activity is the most probable source. Hence, estimates are sorted in the order of their estimated activity with least activity source first. First sorted estimate is verified against all the input data points such that calculated radiological value corresponding to any of the input data points shall not be more than measured radiological value by a large margin. If first sorted estimate does not verify against the input data, then other estimates are verified in the sorted order and first verified estimate is considered as the estimated source.

Results and Discussion: The algorithm has been successfully developed in 'R' and tested in the field surveys. [Figure 1] depicts a field survey carried out in the presence of a 5.74 GBq Cs137 Point radioactive source shown as blue dot. Green dot and red dot represents measurement data points and estimated source location respectively. [Figure 1]b shows the improvement in the accuracy of the estimated source location with more relevant data points as compared to [Figure 1]a. Source could be located with positional error less than 10 meters.{Figure 78}


Bukartas A, Finck R, Wallin J, Rääf CL. A Bayesian method to localize lost gamma sources. Appl Radiat Isot 2019;145:142-7.Silswal A, et al. Regularized Particle Filter based algorithm for the state estimation of orphan gamma source in real time using a backpack gamma spectrometry system. Appl Radiat Isot 2021;169.

 Abstract - 52455: Methodology for improvement of minimum detectable activity of radioactive iodine measurement by using high purity germanium detector based gamma spectrometer at Kaiga Generating Station-3&4

G. S. Salunke, S. S. Managanvi1, Veerendra Danannavar1, N. M. Nayak, V. Udaykumar, Ashok Bhatia2, M. Seshaiah2, B. Vinod Kumar2, N. Karunakar3

Chemistry Control Laboratory, 1Health Physics Unit, 2Kaiga Generating Station-3&4, Nuclear Power Corporation of India Limited, Uttar Kannada, 3Center for Advanced Research in Environmental Radioactivity, Mangalore University, Mangalagangothri, Karnataka, India

E-mail: [email protected]

Qualitative and quantitative analysis of radiochemistry parameters is an essential safety requirement in a Nuclear Power Plant (NPP). It is carried out by Health Physics Section and Chemistry Control Laboratory. One of the key indicators to evaluate the plant performance is Fuel Reliability Index (FRI). The FRI is defined as the steady state primary coolant Iodine-131 (I-131) activity (Bq/g), corrected for the tramp uranium contribution and power level and normalized to a common purification rate. The significance of the FRI is that it is an indicator of the industry progress in achieving and maintaining high fuel integrity and to foster a healthy respect for preservation of fuel integrity. This paper details the Innovative methodology developed for improvement of Minimum Detectable Activity (MDA) in radioactive Iodine measurement using High Purity Germanium Detector (HPGe) based gamma spectrometer for the realistic assessment of the FRI at Kaiga Generating Station-3&4 (KGS-3&4), comprising of two 220 MWe PHWRs. The MDA of radioactive Iodine has direct bearing on FRI. The conventional method of Iodine activity measurement by preparing planchet of 2 ml system heavy water sample in Indian NPP indicates MDA up to 0.5 μCi/L (18.5 Bq/ml). In this method heavy water sample was evaporated using IR lamp which leads to loss of heavy water in every analysis. I-131 specific activity in the coolant system of KGS-3 and KGS-4 is approximately 1.0 μCi/lit (37.0 Bq/ml) and 0.1 μCi/lit (3.7 Bq/ml) respectively. With the existing method I-131 activity in the system sample lower than 0.5 μCi/L (18.5 Bq/ml) couldn't be measured. Hence the need aroused to precisely quantify the activity of I-131 and other isotopes for improvement of FRI. Generally, MDA can be improved by increasing the efficiency of detector, decreasing the background, increasing the counting time and increasing the sample volume. Since detector efficiency is the intrinsic property, the improvement is MDA is achieved by reducing the background count rates by providing the graded shielding, increasing the counting time and the sample volume. In this work, to optimize sample volume, certified reference standard sources of Am241, Ba133 and Eu152 in 10 ml, 60 ml, 500 ml and 1000 ml geometries to cover entire energy range of interest were used for the calibration of gamma spectrometer. Efficiency calibration was carried out from 3.0 cm to higher distances from detector. From the established efficiency curves, MDA was calculated for different energies/radionuclides. By following this methodology, sample volume of 60 ml with counting time of 2000s at 15.0 cm distance from the detector, MDA arrived is 0.003 μCi/l (0.11Bq/ml). For validation of the results, an inter comparison study was conducted at Center for Advanced Research in Environmental Radioactivity (CARER), Mangalore University. The specific activities of I-131 and other radionuclides in PHT samples of KGS-3 and KGS-4 measured at KGS-3&4 were in good agreement with the results arrived at CARER. The improvement made in MDA by conceptualization and making use of new techniques significantly helped in improving the station's FRI values. This also helped to detect other isotopes of very low levels in different reactor system samples. This methodology of liquid sample counting directly is adopted at KGS-3&4 for routine monitoring programme depending on the activity levels. This is a nondestructive method as heavy water from the sample is retrieved and helps in preservation of heavy water. In this methodology open handling of radioactive sample is reduced which reduces the radiation hazard.

Keywords: Certified reference standard sources, fuel integrity, fuel reliability index, graded shielding, minimum detectable activity, multi-channel analyzer, radiochemistry parameters


Cember H, Johnson TE. Introduction to Health Physics: Fourth Edition.IAEA Safety Standards Chemistry Programme for Water Cooled Nuclear Power Plant Dated; 2011.High Resolution Gamma-Ray Spectrometry Analyses for Normal Operations and Radiological Incident Response EPA United States Environmental Protection Agency Dated; 2019.

 Abstract - 52458: Study of working life of commercial cameras in the hot-cell environment in the perspective of radiation induce damages

M. K. Tiwari1, Sanjay Singh2, P. B. Bhandut3, Jyoti Diwan3, Anita Topkar1,4

3Waste Management Division, Bhabha Atomic Research Centre, 2Health Physics Division, Bhabha Atomic Research Centre, 4Electronics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India, 1Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Hot Cell is a high integrity enclosure for processing and handling of very high radioactive source. From the point of view of radiological protection, operations in the hot cell are done remotely. Remote Viewing system plays an important role inside the hot cell from the point of view of remote material handling activities as view from Radiation Shielding Window is limited. This system utilizes commercially available Charge Coupled Device (CCD) or CMOS Active Pixel Sensor (CMOS APS) based camera due to their technical suitability and cost effectiveness as well. However, commercially available camera fails due to radiation induced surface damage and bulk damage of Metal Oxide Semiconductor (MOS) devices.[1] In the present study, investigations were carried to evaluate Total Ionizing Dose (TID) sustainability of CMOS APS based camera in the hot cell of a high-level radioactive waste management facility and the results are presented in this paper.

Material and Methods: Phenomena is studied through both controlled and field experiments. In both the experiments, cameras as per the specifications shown in [Table 1] were used. Video data from the camera was recorded in a Network Video Recorder (NVR) for offline analysis. For the measurement of TID during field experiment, suitable radio-chromic film dosimeter was fixed on the camera surface close to its sensor location.

Results and Discussion: The controlled experiment was carried out using Co-60 source at the dose rate of 60 Gy/hr. Pre-irradiated and post irradiated images from the camera are shown in [Figure 1]. The degradation of image quality is clearly seen at TID of 130 Gy and subsequently, the communication of the camera with the NVR failed abruptly. Performance of the camera was evaluated using image analysis technique. The resulted mean square error (MSE), an indicator of image quality degradation is shown in [Figure 2]. For comparison, the working life of camera in actual field condition of hot cell, TID data for the failed camera are collected by processing the film dosimeter provided with the camera in the hot cell. The TID value for the camera was 438 Gy. This is more than those seen in control experiment where camera failed at TID of 130 Gy. This deviation could be due to the reasons (i) Dose rate of the of hot cell is lesser than the dose rate in the controlled experiment (ii) The camera in the hot cell was kept off whenever operation was not there. During this, camera recovers from the damage due to room temperature annealing.[2]

Conclusions: Working life of the commercial camera is estimated and found their sustainability up to a maximum TID of 438 Gy. In view of difference in result of controlled and field condition, further experiment is planned to investigate TID and dose rate dependency on the working life of the camera.{Figure 79}{Figure 80}{Table 21}

Keywords: CMOS active pixel sensor, dose assessment, radiation effect on camera, total ionizing dose


Oldham TR, McLean FB. IEEE Trans Nucl Sci 2003;50:483-99.Wang Z, Chen W, He B, Yao Z, Xiao Z, Sheng J, et al. AIP Adv 2015;5:107134.

 Abstract - 52560: Repeatability and reproducibility coefficient of ultra-trace uranium measurement in sub-surface water

S. K. Sahoo, J. S. Dubey1, G. P. Verma1, A. Gupta1, S. K. Jha1,2, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Repeatability and reproducibility are the two fundamental quality control parameters of an analytical technique which are the performance and acceptance indicators of the measurement results. Though both the terminologies are used synonymously by chemists but there is a fine demarcation between them according to ISO.[1] Repeatability is a measure of precision or dispersion characteristic of the measurement result of a measurand obtained with the same method by the same operator using the same equipment within short time interval. Whereas, reproducibility is a measure of precision or dispersion characteristic of the measurement result of a measurand obtained by different methods by different operators using different equipment. In the present study, the repeatability and reproducibility standard deviations were estimated for uranium estimation. Uranium standard solutions (0.1, 0.5, 1, 1.5 and 3 ppm) were prepared from the certified and NIST traceable uranium stock solution (1001 ±5 ppm). All the dilutions were carried out gravimetrically to reduce uncertainty in dilution. These diluted uranium standard solutions were analyzed using LED induced fluorimeter (LF), Adsorptive Striping Voltammetry (SV) and Liquid Scintillation Analyzer (LSA) following IS 14194(part3):2021, DIN38406-17:2009 and ISO 13169:2018, respectively. Ten aliquots of 1 ppm were analyzed in all the three techniques under repeatability condition. The repeated analysis of uranium standards in LF is given in [Figure 1]. The percentage dispersion was estimated to be 7.3, 3.9, 4.7, 2.8 and 3.9% for 0.1 ppm, 0.5 ppm, 1 ppm, 1.5 ppm and 3 ppm, respectively. Higher percentage dispersion at lower concentration is anticipated bur show a decreasing trend with concentration. The undulated percentage dispersion could be ascribed to random errors and statistical fluctuations on the fluorescence signal at different U concentration. The percentage deviation of all the uranium concentration well within 10%. The reproducibility dispersion of 1 ppm uranium concentration measured in three different analytically techniques having different principle of measurement is given in [Figure 2]. From the figure, it may be observed the standard deviation in LF is higher than SV which is higher than LSA. Lesser uncertainty in LSA due to direct counting of samples without any chemical/physical processing but relatively higher dispersion in LF and SV can be ascribed to analysis of samples after wet digestion and appropriate dilution of samples to bring in the calibration range.

Keywords: LED induced fluorimeter, liquid scintillation analyzer, repeatability, reproducibility, striping voltammetry, U{Figure 81}{Figure 82}


ISO 21748:2017: Guidance for the Use of Repeatability, Reproducibility and Trueness Estimates in Measurement Uncertainty Evaluation.

 Abstract - 52577: Comparison of airborne 14C release rate from Kaiga nuclear power plant with other nuclear power plants of the world

Bharath, K. Arya Krishnan, S. S. Manganvi1, D. Veerendra1, G. S. Salunke2, N. Karunakara

Centre for Advanced Research in Environmental Radioactivity, Mangalore University, Mangalagangothri, 1Health Physics Unit, KGS 3&4, Kaiga Generating Station, 2Station Chemist, KGS 3&4, Kaiga Generating Station, Uttara Kannada, Karnataka, India

E-mail: [email protected]

Introduction: Carbon-14 (14C) is a long-lived beta-emitting radionuclide (T1/2 = 5730 y). It is produced naturally in the upper atmosphere by cosmic ray interaction with nitrogen.[1] On the global scale, the anthropogenic source of this radionuclide is mainly the atmospheric nuclear weapon tests conducted in the 1950s and 1960s. Operation of the nuclear facilities contributes a very small fraction of the natural inventory of this radionuclide, and the increase in the specific activity, if any, is confined to the local scale. The Centre for Advanced Research in Environmental Radioactivity (CARER), Mangalore University, initiated a detailed study in the year 2014 on the standardization of methods for 14C measurements in stack effluents and environmental biota samples. This paper presents the annual discharge of 14C (in oxide form) through the gaseous route from the PHWR nuclear power plant (NPP) at Kaiga, India.

Materials and Methods: In PHWRs, the 14C from the reactors is released mainly in the form of CO2 through gaseous effluents. Gaseous effluent samples were collected from the common gaseous effluents stack of PHWR reactors 3 & 4 at Kaiga NPP. Samples of gaseous effluents were drawn through a set of two bubblers, interconnected to each other, containing 100 mL each of 2M ultra-pure NaOH solution (2NaOH + CO2→Na2CO3 + H2O). Before passing through the NaOH solution, the gaseous sample was bubbled through 1M HNO3 to trap 3H. The sampling duration was 24 h at a flow rate of 1 L m-1. After sampling, BaCl2 was added to the sampled solution, and BaCO3 was precipitated. The BaCO3 was taken in a specially designed CO2 regeneration apparatus, and the evolved CO2 was reabsorbed in a 10 ml + 10 ml mixture of CarbosorbE and Perma-flour scintillator (both from PerkinElmer, Inc.m USA). The mixture was transferred to a poly-ethylene vial and stored in the dark overnight to reduce the chemi-luminescence before subjecting for liquid scintillation spectrometric analysis.

Results and Discussion: The gaseous effluents from the common stack of two units of PHWR (each of 220 MWe) were monitored periodically for four years, during 2017-2020. From the specific activity of 14C in gaseous effluent, the annual release through the stack was computed considering the stack flow rate. The average normalised 14C release rate (880 MWe) was 1.3 TBq GWe-1a-1 (GM= 0.28) [Table 1]. A comparison of the release rate measured for Kaiga NPP with those reported for other NPPs of the world is also presented in [Table 1]. There is only one report available in the literature[4] for NPP in India, which is for the Rajasthan Atomic Power station. The reported release rate for this NPP is higher than that observed for the Kaiga NPP.{Table 22}

Keywords: Carbon-14, PHWR, release rate


Authors would like to thank the Board of Research in Nuclear Science (BRNS), Department of Atomic Energy, Government of India for funding the research programme. The technical support received from the staff members of ESL, Kaiga and HP unit, KGS 3&4 during sampling is also acknowledged.


Libby. Atmospheric helium-3 and radiocarbon from cosmic radiation. Phys Rev 1945;69:679-2.IAEA. Management of Waste Containing Tritium and C-14. Report IAEA 421, Vienna; 2004.Sohn, et al. An Estimate of C-14 inventory at Wolsong Nuclear power plant in the Republic of Korea. J Nucl Sci Technol 2003;40.Joshi, et al. Measurement of 14C emission rates from a pressurized heavy water reactor. J Health Phys 1986;52:787-91.

 Abstract - 52584: Development of phosphate functionalized membrane for environmental monitoring of uranium

Manish Shingole, Sumana Paul1

Homi Bhabha National Institute, Anushaktinagar, 1Fuel Chemistry Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Uranium is a radiotoxic and nephrotoxic element with a maximum permissible limit of 30 ppb in drinking water. This calls for the need to design and develop suitable materials for selective separation of uranium from water resources, when the contamination level is higher than the prescribed limit. In addition, there is a global concern over the limited resources for fossil fuels and nuclear energy is considered to be one of the most viable options to meet the fast increasing demand for energy all over the world. However, the sustainability of nuclear energy program depends upon the availability of nuclear fuel. Though seawater is the largest reservoir of uranium with a total global pool of ∼4.3 billion tons –recovery of uranium from seawater is challenging considering the ultra-low concentration (∼3.3 ppb) of uranium in sea water and the presence of high concentration of Na+, Ca2+, Mg2+ etc. A lot of researchers are working towards overcoming the challenges of extraction of uranium from seawater.[1] The present study focuses on the development of a phosphate functionalized membrane for the extraction of uranium from natural water. In the present work, the monomer bis[2-(methacryloyloxy)ethyl]phosphate (MEP) was grafted on the poly(ethersulfone) (PES) membrane by UV-radiation induced grafting method. The choice of the monomer was based on the fact that, MEP contains a phosphate group that is highly selective towards actinides.[2] Poly(ethersulfone) membrane was chosen as the substrate for grafting because PES has a porous structure that is crucial for fast and easy accessibility of the functional groups for complexation with uranyl ions present in the aqueous medium. The synthesis of the MEP grafted PES membrane ([email protected]) was a simple, rapid and one-pot process. The SEM image of [email protected] showed considerable porosity even after grafting of the monomer which is very crucial for quantitative sorption of uranium with improved kinetics. The phosphorus mapping by EDS, on the surface and along the cross-section of [email protected], indicated uniform distribution of the extractant MEP on the surface as well as within the pores of the PES membrane. [Figure 1] shows uranium extraction by [email protected] as a function of the pH of the equilibrating medium. It can be seen from [Figure 1] that, [email protected] can extract more than 95% of uranium from aqueous medium when pH is varied from 1 to 8. Also, more than 90% of uranium was extracted by [email protected] from the mixtures of uranium with various competing cations (Na+, Ca2+, Mg2+, Cd2+, Pb2+) and anions (Cl-, SCN-, NO3-, SO42-, CH3COO-, C2O22-). To explore the feasibility of employing [email protected] for the pre-concentration of uranium from various natural water samples, sorption experiments were carried out with four real samples viz. tap water, ground water, borewell water and sea water having pH = 7.3, 6.8, 4.9 and 8.1 respectively. The real water samples were spiked with known weight of 233U tracer (Eα=4.8 MeV) and then equilibrated with [email protected]. [Figure 2] shows the uranium uptake by [email protected] from the real water samples. It was found that more than 95% of uranium was extracted by [email protected] from all the real samples, without any pH adjustment. This indicates the high potential of employing [email protected] membrane for the pre-concentration of uranium from various natural water samples, including seawater.{Figure 83}{Figure 84}

Keywords: Membrane, natural water, radiation grafting, uranium


Singhal P, et al. J Hazard Mater 2020;384:121353-61. Paul S, et al. Anal Chim Acta 2015;878:54-62.

 Abstract - 52591: Determination of dose rate-to-gamma activity conversion factor for online gamma activity monitoring system

S. Selvaganapathy1, B. Suresh1, Atul Prakash2, S. Murugan1, G. Ganesh1, M. S. Kulkarni1,3

1Health Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Mumbai, Maharashtra, 2Waste Immobilisation Plant, INRP(K), NRB, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Introduction: Ion exchange process is one of the methods of treatment of radioactive liquid wherein selective extraction of radionuclide of interest from the solution is achieved. The efficacy of radionuclide retention in the ion-exchange resin column during treatment at any given time depends on the saturation level of radionuclide in the column. One way to estimate the same is by analysing the activity level in the effluent generated, post treatment. However, loading of IX column being a continuous process, frequent sampling and analysis for activity is time consuming. An alternate and efficient method would be online assessment of activity level in the effluent as it is being generated after IX treatment.[1] The study presents the design and response of a system capable of online assessment of gamma activity for varying activity concentrations in the liquid and as well determine the minimum activity concentration that the system is capable of detection.

Materials and Methods: A schematic of the Online Gamma Monitor (OGM) experimental setup for the study is presented in [Figure 1]. The OGM comprises of a) a stainless steel annular chamber through which the effluent generated post IX treatment is allowed to flow through, b) an external Geiger-Muller detector probe mounted in the centre of the annular chamber and c) wall-mounted electronic unit, which houses the associated instrumentation and display unit. The experiment is carried out by filling the chamber with 137Cs standard solution of varying activity concentrations (137Cs being the radionuclide of interest in ion-exchange treatment and also the predominant radionuclide in the radioactive liquid). The average of five OGM readings is noted for each trial and all the values are tabulated.

Results and Discussion: [Table 1] presents five randomly chosen values (out of 32 observations) of activity concentration of the liquid in the chamber, the corresponding average OGM reading and the normalized activity concentration per unit dose rate of. [Figure 2] depicts the response of the OGM for various activity concentrations. The normalized 137Cs activity concentration per unit dose rate of OGM reading is found to be fairly constant and its average value is estimated to be 72 ± 2 Bq/ml per μSv/h. Considering gamma sensitivity of 1.8 cps/μSv/h of the detector, LND-make 71210,[2] 15-second period (averaging time), an average background of 0.1 μSv/h and 3σ confidence level, the minimum detectable dose rate of the instrument is estimated to be 0.3 μSv/h. The corresponding 137Cs activity in solution is about 22 Bq/ml and helps in early detection of saturation of the IX column for elution.{Figure 85}{Figure 86}{Table 23}

Keywords: Ion-exchange, online gamma monitoring, radioactive liquid


Wingfield EC. A Beta Gamma Monitor for Liquid Streams. US Atomic Energy Commission; 1956.GM Detector Specification of LND-71210. Available from: https://www.lndinc.com/products/geiger-mueller-tubes/71210/.

 Abstract - 52592: Optimization of sampling method for testing of high efficiency particulate air filter banks

D. N. Sangeetha, S. Viswanathan, M. Menaka, V. Subramanian, B. Venkatraman

Safety, Quality, and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Introduction: The High Efficiency Particulate Air (HEPA) filter banks installed in the exhaust system ensures that the air released to the environment remains free from radioactive particulates. The methodology of sampling place a crucial role in evaluation of filter bank efficiency. The sampling needs to be isokinetic to avoid flow error in particle counter which is used for measuring aerosol concentration. This paper discuss about the study on sampling of aerosol concentration in an isokinetic and anisokinetic sampling conditions.

Methodology: The retention efficiency of HEPA filter is evaluated at IGCAR as per EN 1822 by injecting the approved test aerosols using aerosol generator and measuring the concentration of aerosol in upstream and downstream of filter banks using laser based optical particle counter. The sampling of aerosol in the airstream must be isokinetic, means that the concentration and size distribution of the aerosol entering the sampling tube is the same as that in the flowing stream. If the sampling is anisokinetic, it may be result in distortion of the size distribution and misrepresentation of the concentration. The anisokinetic sampling may be due to,(i) the probe is not aligned with the gas flow streamlines, (ii) the gas velocity in the probe exceeds the stream velocity & (iii) the velocity of the stream exceeds the velocity in the probe. Hence, for a free stream velocity Uo and a gas velocity U in the probe, the isokinetic condition for a properly aligned probe is U=Uo. The flow rates in the duct and tube must be directly proportional to their respective cross-sectional areas. The sampling probe of diameter Ds is used at a sampling flow rate Qs, in a circular duct of diameter Do carrying a flow rate Qo, [INSIDE:7]. Under the condition (iii) stated above, the inlet and outlet of the instrument were connected in a loop with duct in order to maintain same velocity pressure at the upstream and downstream junction of the instrument with duct flow.

Results and Discussion: An in-situ Filter bank test was conducted under anisokinetic sampling condition [Figure 1] when the instrument connected to inlet from downstream sampling point showing flow error. To achieve isokinetic sampling, the instrument outlet was connected back to upstream sampling point, when inlet is connected to downstream and vice versa as shown in [Figure 2]. This condition was evaluated by calculating flow Reynolds number and Stokes number such that flow through sampler is laminar and particle traversing in the flow stream are not deviated. The flow Reynolds number in the duct and sampler is found to be ~3 showing laminar flow condition and Stokes number (stk|) in the duct and in the instrument is 0.001809 and 0.0017065345 respectively showing stk tending towards zero, which means particles follow gas streamline. To validate the results of in-situ HEPA filter bank test similar in-house test was conducted using two fresh filters at HFTL and the results obtained were tabulated in [Table 1] and [Table 2]. The percentage efficiency is calculated as:

[INSIDE:8], Cu & Cd are Avg Upstream & Avg Downstream concentration. The upstream and downstream counts of injected aerosol were found to be similar counts in both with loop connected to instrument to duct at upstream and duct to instrument in downstream and vice versa. The efficiency of the filters is 99.99%.

Conclusion: The methodology adopted for sampling by connecting the instrument in a loop where outlet back to the upstream sampling point to ensure the isokinetic sampling when velocity of upstream exceeds the probe velocity condition. The observations with loop and without loop reveals, not having any impact on filter efficiency. Hence, the method of sampling by connecting the instrument in a loop can be adopted for ensuring isokinetic sampling and to maintain the gas velocity in the sampling probe.{Figure 87}{Figure 88}{Table 24}{Table 25}

Keywords: Anisokinetic sampling, efficiency test, high efficiency particulate air filter

 Abstract - 53138: Simulation study of front-end electronics design of Silicon Photomultipliers for optimal performance in gamma monitoring application

S. Srivastava1,2, A. V. Kumar1,2, A. Topkar2,3

1Environmental Monitoring and Assessment Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, 3Electronics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: In the present work, a SPICE simulation study was carried out to investigate the pulse response and noise response of Silicon photomultiplier (SiPM) along with front end electronics (FEE) comprising a voltage sensitive or charge sensitive amplification schemes. The study was targeted for the development of Gd3Ga3Al2O12:Ce,B (GGAG:Ce,B) scintillator detector based compact gamma monitor for environmental monitoring applications. Using this study, various design parameters such as amplifier gain, biasing resistance, feedback capacitance or resistance, etc., were optimized for best performance in terms of improved signal to noise ratio (SNR). The results of this simulation study are presented in this paper.

SPICE Model of SiPM: The SPICE electrical model for circuit simulation was realized using a voltage controlled switch (S_Trigger) for avalanche triggering and a current controlled switch (S_Sustain) during avalanche region [Figure 1]a. The SiPM pixels were grouped into fired and non-fired pixel groups.[1] In [Figure 1]a, the quench resistance, quench capacitance and depletion capacitance of the fired and non-fired pixel groups are indicated as RQ_F, CQ_F, CD_F and RQ_NF, CQ_NF, CD_NF, respectively. C_G and R represents the overall grid capacitance and contact resistance of the device, respectively. The values of these parameters of a commercially available SiPM (pixel area 15×15 μm2, total 3600 pixels) were incorporated into the SPICE model to simulate the SiPM pulse response. The simulated pulse response of the SiPM shows that the SiPM produces a current pulse of height 2 μA and width 100 ns with a rise time of less than 10 ns for single photon detection. As discussed later, a FEE was designed to convert the pulsed current response of SiPM into a voltage signal and its performance was studied using circuit simulations.

Front-end Electronics: The simplified schematic representation of voltage sensitive and charge sensitive amplification schemes interfaced with SiPM SPICE model are presented in [Figure 1]b and [Figure 1]c, respectively. Considering the output current pulse response of SiPM for single photon, it was found that a FEE with overall gain of about 20 V/mA and a bandwidth of 400 MHz is required to amplify the output pulses in the voltage mode readout. The simulation results also showed that a higher bias resistance [[Figure 1]b RSense ~ 470 Ω] increases the gain linearly without a significant increase in the noise voltage and thus helps to achieve better SNR. In case of a charge sensitive amplifier, the simulation results of the pulse response and noise analysis indicated that with an increase in the bias resistance, the noise voltage decreases. However, as the bias resistance approaches the quench resistance of the device, due to increase in the overall resistance in the avalanche breakdown region, the output pulse height decreases. The noise power was observed to be concentrated within 30 MHz bandwidth. The results of FEE readout parameter optimization through the simulation study are presented in [Table 1].

Conclusion: Using a comprehensive analysis of SiPM with FEE, the design of readout electronics for SiPM has been optimized. The optimized FEE with voltage amplification scheme has been fabricated to study the response of SiPM based gamma monitor.{Figure 89}{Table 26}

Keywords: Front-end electronics design, gamma monitoring, SiPM, SiPM modeling, SPICE modeling


Acerbi F, et al. Understanding and simulating SiPMs. Nucl Instrum Methods A 2019;926:616.

 Abstract - 53272: Design development and testing of a neutron activation based novel criticality accident alarm system

Deepa Sathian1,2, Kapil Deo3, A. K. Bakshi1,2, Umasankari Kannan2,3 and B. K. Sapra1,2

1Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, 3Reactor Physics Design Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

A criticality accident is the release of enormous amount of energy as a result of an accidentally produced self-sustained and divergent neutron chain reaction occurring in a quantity of fissile material. This accident can happen in places like nuclear fuel reprocessing plants, fuel fabrication and storage facilities, experimental research reactors and all other areas where fissile materials are handled. A criticality accident alarm system (CAAS) is an equipment intended to provide warning of a criticality accident with the detection of gamma radiation, neutrons or both from a criticality accident.[1] The primary purpose of the criticality alarm system is to detect radiation from criticality accidents and warn personnel present at the plant site for immediate evacuation, thus to reduce the probability of serious exposure to personnel. Most of the countries including India is using gamma radiation-based criticality alarm monitors which are vulnerable to false positive alarms in special scenarios [viz. A 10 mCi of 60Co source movement from 1m distance to 5 cm distance in 500 msec can produce a false alarm in the gamma based criticality alarm system]. To overcome this difficulty, a novel neutron detection-based criticality accident alarm system is designed, developed and tested in the current study. The newly designed novel criticality accident alarm system (CAAS) consists of a cylindrical stainless steel tank with a provision to keep a 2” x 2” NaI (Tl) gamma detector at the centre of the tank from outside as shown in the block diagram. The tank is filled with vanadium sulphate solution which is prepared by dissolving vanadium sulphate powder in de-mineralized light water. The dimension of the tank and the concentration of vanadium was optimized to give the required activity to generate the alarm in the system. The energy calibration of the detector system was carried out using 137Cs, 60Co and 40K gamma sources. The detection window was selected for vanadium gamma of 1.43 MeV by adjusting the LLD and ULD of the gamma detector in the window mode. Also, this CAAS system put in a proper lead shield can prevent the interference of other gammas without reducing the intensity of neutron field. The alarm system was tested with 252Cf, 241Am-Be, 252Cf+D2O and 241Am-B ISO standard neutron sources. The 52V induced activity in the tank was measured using the system attached NaI (Tl) gamma detector and evaluated efficiency. Monte carlo simulation of CAAS was carried out using FLUKA,[2] to estimate the induced saturation activity [NσФ] in the tank due to the neutron sources. Measured activities were compared with the theoretical estimates. A neutron activation based novel criticality accident alarm system was designed, and fabricated. The system response was tested with various neutron sources including fission neutron source (252Cf), as the neutron spectrum after a nuclear accident is generally fission spectrum. Alarm functioning was also tested by keeping the alarm threshold much below the calculated saturation activity. The chance for false alarm can be removed by setting the activity threshold high, which is difficult to achieve using laboratory neutron sources and also setting gamma window for vanadium peak. Monte Carlo simulation of CAAS was also carried out. This would aid in further optimization of the same. This unique neutron activation based CAAS can be implemented in all potential areas of accidental criticality after its complete performance evaluation.{Figure 90}{Figure 91}

Keywords: Criticality accident alarm system, fission neutron source, liquid activation, vanadium sulphate


The authors are grateful to Dr. D. K. Aswal, Director, HS&EG for his encouragement. Thanks are also due to all RSS members for their help and support during irradiation studies.


ANSI/ANS-8.3-1997; R2003; R2012; R2017 “Criticality Accident Alarm System“; 2017.Ferrari A, et al. CERN-2005-10, INFN/TC_05/11, SLAC-R-773; 2005.

 Abstract - 53583: Study on variation of efficiency with various matrices and detectors

I. Vijayalakshmi, B. Arun, S. Viswanathan, M. Menaka, V. Subramanian, B. Venkatraman

Safety, Quality, and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Introduction: High purity germanium detector based gamma detectors are used for estimation of radioactivity in various samples in various geometries. In order to estimate the activity in these samples efficiency and efficiency curve has to be established for various geometries and different matrices. The source distance, the crystal absorption cross-section and shape, the window and volume of the detector are important to determine the efficiency of the detector. HPGe detector offers very good resolution and is a good instrument for nuclide identification.

Materials and Methods: Two 50% RE HPGe detectors were used for the efficiency comparison of different make one is named as Detector A and other as Detector B. For NORM estimation in the environmental samples efficiency was established using IAEA stds RGK, RGU, RGTh packed in 250 ml plastic bottles. For Liquid Geometry 1 litre and 250 ml Liquid Eu-152 was used and efficiency curve was established. Eu-152 Disc source was kept at distance of 10cm from the detectors for establishing efficiency curve. Both the detectors were provided with 10cm thick lead shielding and is provided with graded shielding. IAEA std soil -375 was used for the efficiency of Cs-137 in the soil samples.

Results and Discussion: The efficiency of Natural occurring materials (NORM) is shown in the [Table 1]. The deviation in the NORM standards (K-40, U-238, Th-232) for the two detectors is from 0.44 to 0.55. The deviation in the IAEA standard Soil -375 is 0.002. The efficiency curve for the liquid standard Eu-152 250 ml is shown in the [Figure 1]. There is a slight difference in the efficiency at lower energies. The efficiency curve for the liquid standard Eu-152 (one litre) is shown in the [Figure 2]. In the [Figure 1] and [Figure 2] the efficiency variation is seen only in the lower energies. The efficiency of detector A is slightly higher than the detector B. This may be due to variation in the thickness of the end cap. The end cap thickness of detector A is 0.7 mm whereas the thickness of detector B is 1 mm.The sensitive area of detector A is 70.4 mm of diameter and depth of 57.3mm. The sensitive area of Detector B is 69 mm of diameter and depth of 59.8mm. Hence the volume of detector A is 222.93 cm3. The volume of Detector B is 223.9 cm3. In case of the Disc source Eu-152 the variation in the efficiency was due to the height of the source holder. The source holder of 10 cm was placed on the top of the detector. The source holder was exactly fitted with Detector A whereas in Detector B the source holder height was increased by 0.5 mm. Hence there is a variation in the efficiency between detector A and Detector B.

Conclusion: Study reveals that the detector has to be chosen for any particular application, considering the energy range, geometry of the sample and the sensitive area of the detector. Hence, the Crystal dimension plays the important role in the determining efficiency of the detectors as they can affect the solid angle and absolute counting efficiency. In this study there is no much variation in the volume of the detectors, hence variation in the efficiency for different matrices of the detectors is negligible.{Table 27}




Keywords: Efficiency, HPGe detector, IAEA stds

 Abstract - 53621: Design and performance test of an eye lens dosimeter C-lens

Xiafeng Wei, Liye Liu, Qinjian Cao, Yuan Zhao, Yunshi Xiao, Yan Jiao

Department of Health Physics, China Institute for Radiation Protection, China

E-mail: [email protected]

Recently, with the dose limit of eye lens reduced, a series of problems in research, monitoring, protection and evaluation, have gradually attracted attention. The dosimeter of the eye lens plays an important and foundational role in this circle. This paper mainly focuses on an eye lens dosimeter named C-lens. Firstly, the type of eye-lens dosimeter, one single LiF:Mg,Cu,P elements with a plastic shell, was chosen. Secondly, with MC simulation method, the energy response and angular response of the dosimeter were optimized and improved by adjusting the thickness, material and shell shape of the dosimeter. Then, the prototype and final dosimeter was completed by 3D printing and injection molding respectively. Finally, the radiation performance of C-lens was tested in a radiation metrology station, which is an IAEA Sub-standard Lab. The results show that the non-linear response was less than 6% within 0.01 to 100mSv, relative energy response (normalized to 661keV) for photons was between 0.80 and 1.25 in the range of 20.3 to 1250 keV, and the relative deviations of angular response(normalized to 0°) was less than 4%when the incident angle was less than 60°.{Figure 92}{Figure 93}{Table 28}

Keywords: Dosimeter, eye lens, Hp(3), personal dose

 Abstract - 54130: Elemental analysis of ancient pottery samples from recently excavated archaeological site of Tamil Nadu using scanning electron microscopy with energy dispersive spectroscopy

A. Tamilarasi, A. Chandrasekaran

Department of Physics, SSN College of Engineering, Kalavakkam, Chennai, Tamil Nadu, India

E-mail: [email protected]

Introduction: The characterization of archaeological materials such as pottery, bricks, stones, etc. leads to getting information about the fabrication techniques, the provenance of the raw material, chemical composition, the age of the material, and ancient civilization.[1] This present work, aims to ensure that the elemental concentration of the ancient pottery samples collected from recently excavated Neolithic archaeological sites of Tamil Nadu with the succour of Scanning Electron Microscopy equipped with Energy Dispersive Spectroscopy (SEM-EDS). Furthermore, the firing temperature, firing condition and firing atmosphere which is prevailed at the time of manufacturing were estimate based on the elemental concentration.


To identify the elemental concentration of i) the pottery samples

ii) To identify the provenance of the raw material the potteries used by ancient artesian.

iii) To estimate the firing temperature of the potteries at the time of manufacturing.

Abstract Review: The surface morphology of the potteries provides an extensive information of the raw materials based on the elements present in the potteries. The SEM image of the samples are shown in [Figure 1]. The concentration of calcium oxide is 2.33%, 3.06% and 2.04% in the collected samples KMI, PLR, and PNR respectively. It indicates that the raw material of the sample is non-calcareous clay because the presence of calcium element in the samples is less than 6%.[2] The topographic image of the pottery samples by SEM-EDX show the initial vitrification implies that the raw material is non-calcareous clay and these samples were fired in the oxidizing atmosphere with the temperature range between 600°C and 800°C. The concentration of the elements such as Al, Si, Pb, K, Ca, Fe, Cu, Zn, Hg are listed in the [Table 1]. This study is very informative for the identification of clay sources for pottery.{Figure 94}{Table 29}

Keywords: Ancient potteries, Neolithic archaeological site, Non-Calcareous clay, SEM-EDX, Topographic image


Manoharan C, Veeramuthu K, Venkatachalapathy R, Radhakrishnan T, Ilango R. Spectroscopic and ancient geomagnetic field intensity on archaeological pottery samples, India. Lith J Phys 2008;48:195-202.Maniatis Y, Tite MS. Technological examination of Neolithic-Bronze age pottery from central and southeast Europe and from the Near East. J Archaeol Sci 1981;8:59-76.

 Abstract - 54294: Parametric optimisation of BEGe detector for efficiency calibration by simulation

Anilkumar S. Pillai, Amar D. Pant, Amit Verma, A. Vinod Kumar

Environmental Monitoring and Assessment Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

The Broad Energy Germanium (BEGe) detectors are a special type of HPGe detectors, offering an optimised efficiency in broad energy range from 10 keV to 3 MeV with good energy resolution. This detector is an ideal choice for environmental radioactivity measurements. The efficiency calibration of the system is required to be done using standard radioactive sources of same geometrical dimensions, density and composition, as that of the sample. But it is not possible to have physical standards of all nuclides encountered in the laboratory. The alternate method is the application of Monte Carlo simulation techniques.[1],[2] The present work describes a methodology of simulation used for the BEGe detector and its validation with experimental measurements using standard reference sources. The physical parameters of the detector, such as diameter and length, has been optimised using experimental measurements of standard sources. Standard reference sources of 60Co and 152Eu with known activity have been used. The experimental set up consists of BEGe detector of BSI make with ORION MCA and InterWinner software for data acquisition and analysis. The relative efficiency reported by the manufacturer is 60 % for 1332 keV of 60Co. In order to make a detector model for efficiency calibration, it is essential to optimise the major detector dimensions like diameter and height based on the data given by the manufacturer. After using the dimensions given by the manufacturer, we observed a deviation of 25% in the simulated relative efficiency compared to the experimentally observed value using 60Co source. The major detector parameters studied for the optimisation was crystal diameter and height. Absolute detection efficiency of prominent gamma energies from 152Eu source was used for the optimisation of diameter and height of the detector. The optimised parameters were validated experimentally using source of 152Eu kept at a distance of 25 cm from the detector. From the measurements it is observed that the variation in the simulated and experimental efficiency values for the wide energy range is uniform. This indicate the requirement of correction in diameter of the detector contributing to the geometrical efficiency. The diameter of the detector was optimised to match the absolute experimental efficiency for the gamma energies of 152Eu. The ratio of absolute efficiency of prominent gamma energies of 152Eu estimated by simulation (εs) and experiment (εe), before and after optimisation is given in [Table 1].

The relative efficiency of the detector was then estimated using this optimised detector diameter and found to be in agreement with the reported value given by the manufacturer. The error in simulation is <3%, which is mainly due to uncertainty in activity of reference sources used for validation. The simulated spectra based on optimised diameter and experimental spectra of 60Co at 25 cm from the detector is shown in [Figure 1], validating the simulations. This optimised source-less simulation method was used for efficiency calibration of all the practical geometries, which are in general difficult to generate experimentally and is being used for regular radioactivity measurements.{Figure 95}{Table 30}

Keywords: BEGe, detector simulation, efficiency calibration, gamma spectrometry


Narayani K, Pant A, Amit K, Anilkumar S. Proceedings of the 33rd IARP Conference, Mumbai; 2018.Luis R, Bento J. Nucl Instrum Methods Phys Res A 2010;623:1014-9.

 Abstract - 54363: Innovation of mathematical model for quantitative estimation of radioactivity of samples having gas filled cylindrical geometry through HPGe based gamma spectrometry system

Ravi Kant Sharma, Pradeep Upadhyaya, Barun Mudgal, Sharad Agarwal, Neel Abhishek, D. P. Dwivedi, Mahesh Gautam, B. N. Kesharwani, Amandeep Kakkar, Rajesh Kumar, K. Venkata Ramana1, K. K. De1

Health Physics Unit, Rajasthan Atomic Power Station-5&6, Rawatbhata, Rajasthan, 1Health Safety and Environment Group, Directorate of Technical, HQ, NPCIL, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: At NPPs, HPGe based gamma spectrometry system is used for qualitative & quantitative estimation of radioactivity in different samples. Quantitative analysis of gaseous activity in nuclear plants has been a challenge for nuclear community due to limitations in availability of long lived standard gaseous sources. In order to estimate the radioactivity in a particular volumetric gaseous sample with the help of HPGe based gamma spectrometry system; the HPGe system needs to be calibrated with that particular gaseous volumetric geometry. In absence of standard gaseous sources, the efficiency calibration of HPGe based gamma spectrometry system for gaseous volumetric geometry can be achieved with the help of point source efficiencies & geometric factors of gaseous volumetric geometry. At HPU, RAPS-5&6, these geometric factors are achieved by using standard mathematical integration techniques for gaseous volumetric geometry followed by further computation of radioactivity of gaseous samples.

Materials and Methods: Detector used: N-type Coaxial HPGe Detector EGC 20-185-No 74067 with volume 94.6 cm3.


Here y = distance from base rack (Ref. Point),

R=Radius of cylindrical geometry

H= Height of cylindrical geometry

a=Distance between detector & base rack (Ref. Point)

If “Ep“ is the point source efficiency for any particular gamma energy then efficiency for gas filled cylindrical geometry (Eg) can be estimated as

Eg = Ep × G

Where G = Geometric factor for cylindrical geometry at any given gamma energy


Validation of efficiency Vs energy curve for gas filled cylindrical bottle was carried out by estimating the activity in a gaseous sample collected from Annular Gas Monitoring system (AGMS) of RAPS Unit-5 (PHWR) with the help of a dedicated sampling pump.

Result and Discussion: Observed radioactivity of the gas filled glass bottle by mathematical model was found within standard acceptance criterion (±15%) and hence the methodology is being widely accepted and implemented for quantitative estimation of gaseous samples through this mathematical model to overcome the limitations of unavailability of standard long lived gaseous sources in nuclear power plants.{Figure 96}{Table 31}

Keywords: Cylindrical, gas, model, volume


Cember H. Introduction to Health Physics. 4th ed. Knoll GF. Radiation Detection & Measurement. 3rd ed.IAEA. Calibration of Radiation Protection Monitoring Instruments. Safety Series No-16. IAEA; 2000.

 Abstract - 54417: Gamma spectroscopy for study of isotope composition and activity estimation in high active hull pieces

Rabindra Nath Juine1, V. Ramprasath1, S. K. Nayak1, K. Kannan1, Tarit Kumar Mandal1, G. M. Rao1, Anshuman Singha Roy1, G. Ganesh1, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: In the back end of nuclear cycle, PHWR fuels are reprocessed at reprocessing plants. The fuel bundles are chopped and dissolved in 13 M HNO3 and subsequently processed in Solvent Extraction Process using TBP for separation of Pu, U and Fission products. The dissolution process leaves behind zirconium cladding material as the major solid waste from reprocessing plants. The cladding material waste is termed as Hull. The hull pieces show high radiation field due to presence of trace quantity of undissolved fission products along with fissile material and activation products. Radio-nuclide identification and their activity estimation are important from safety point of view and for waste management strategy. For the present study, high active hull pieces were collected from SS hull drum after complete dissolution and samples were prepared for spectrum acquisition. HPGe was used to acquire Gamma spectrum for radionuclide identification and activity estimation.

Experimental Procedure: High active hull sample collected after chop-leach process were cut into small pieces for gamma spectrometry. Hull samples were distributed on disc in such a way that sample geometry is identical to the source used for efficiency calibration. Net weight of sample was measured by using high accuracy electronic weighing machine. CANBERRA GC3018 HPGe spectrometer with co-axial Ge detector pared with an external liquid nitrogen cooling system and controlled by DSA 100 MCA card connected with Genie 2000 computer based software was used for this present experiment.[1] Energy and efficiency calibration of HPGe spectrometer were performed using standard disc sources.[2] Then gamma spectrum of high active hull piece was acquired. Acquired gamma spectrum was further analyzed using Genie 2000 computer based software.

Results and Discussion: The gamma spectrum obtained from hull samples is shown in [Figure 1]. Radionuclides corresponding to the peaks have been identified and are marked in the spectrum. Analysis of the spectrum indicates the presence of Pb-212, Pb-214, Ac-228, and K-40 which can be attributed to background radiation. The identified fission products present in the hull samples are Sb-125 (T1/2 = 2.73 year), Cs-137 (T1/2 = 30 year), and Eu-154 (T1/2 = 8.8 year). The prominent activation product identified in the hull piece is Co-60 (T1/2 = 5.27 year). Presence of nuclides having substantial fission yields such as Cs-134 (T1/2 = 2.06 year), Ru-106 (T1/2 = 372 days) and Pr-144 (Ce-144, T1/2 = 285 days) could not be detected due to cooling period more than 10 years. Details of activity calculated from the spectrum and the percentage contribution of radionuclides are tabulated in [Table 1].

Conclusion: The hull sample have been subjected for gamma spectrometry using HPGe detector. The dominant radionuclides such as Cs-137, Sb-125, Eu-154, and Co-60 have been identified. Total activity calculated from hull sample with maximum contribution of 90.7% from Cs-137.{Figure 97}{Table 32}

Keywords: Activation product, fuel reprocessing, gamma spectrometry, hull sample, radionuclide


Suarez-Navarro JA, et al. Radiat Phy Chem 2020;171:108709.Wallbrink PJ, Walling DE. Handbook. Dordrecht: Springer; 2002.

 Abstract - 54490: Gamma assay of structural material samples irradiated in fast breeder test reactor with high dose rate using LaBr(Ce)

R. Akila, R. Sarangapani, R. Mathiyarasu, D. Ponraju

Health and Industrial Safety Division, SQRMG, IGCAR, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

The permanent core component, grid plate of Fast Breeder Test Reactor (FBTR) is made of SS316 and the life of the reactor is determined by the performance of grid plate. Hence to evaluate the material properties like tensile strength, toughness, ductility etc, specimen samples were irradiated at Fast Breeder Test Reactor (FBTR) at different displacement per atoms (dpa) which is a calculated value. Assay of these samples are carried out using LaBr (Ce) based gamma spectrometer. The dose rate on the samples was 50 – 70 mGy/h which were irradiated to different dpa varying from 2 to 7. The grid plate samples of size 8mm diameter and 0.5mm thick were irradiated in FBTR for 22 months in 4th ring. As the dose rate on the samples are very high, a collimation set-up was specially designed and fabricated. The collimation has an aperture of diameter 5 mm. The schematic of the experimental set up is shown in [Figure 1]. The result, thus obtained is used in waste management aspects like quantifying the total activity in the samples. This can also be used in assessing the dose during decommissioning of components associated in the primary circuit the fast reactor. The samples were assayed in a 1“x1” LaBr(Ce) crystal coupled to photomultiplier tube. The resolution of the detector is 3% for 137Cs 662keV peak. The acquired spectrum is analysed using MC2 software. The energy and efficiency calibration of the detector was carried out using 152Eu standard reference source. The samples were counted for a time interval of 1000s. The spectrum obtained after counting the samples were analyzed using the MC2 software. The radionuclides identified in the samples are 54Mn and 60Co. It is observed that 60Co contributes to 35 - 40% of the activity and the remaining is due to 54Mn. [Figure 2] shows the typical spectrum of the samples. Owing to its short half life of 312 days, its activity decreases rapidly as compared to 60Co. The specific activities of the irradiated samples were plotted against the dpa and are shown in [Figure 3]. It is observed that the activity increases with the dpa. 54Mn can be used to determine the fast neutron fluence to which the material is exposed. The activity estimated is useful in waste management. This activity can also be used for dose assessment and control during decommissioning of fast reactor. There is scope to establish the relation between the dpa and induced activity which can be explored with further studies on the samples irradiated to different dpa.{Figure 98}{Figure 99}{Figure 100}

Keywords: Gamma spectrometry dpa, Irradiated structural material, specific activity


Perez-Moreno JP, Bolıvar JP, Garcıa-Tenorio R, San Miguel E, Aguado JL, Mas L, et al. Radiat Phys Chem 2001;61:437-8.Diasa MS, Cardosoa V, Vaninb VR, Koskinas MF. Appl Radiat Isot 2004;60:683-7.

 Abstract - 54509: Measurement of minimum detectable activity for HPGe spectrometry system in analysis of low level radioactive samples

Lalit Kumar Vajpyee, Saurav, K. S. Babu, Sajin Prasad, Ranjit Sharma

Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Minimum Detectable activity (MDA) indicates the capability of the detector and method of measurement to measure the radioactivity in a sample. This paper discusses the MDA of the High Purity Germanium (HPGe) based gamma ray spectrometry system. During decommissioning process at a research reactor large volume of solid waste is produced, which is potentially contaminated with long lived gamma emitter like 137Cs and 60Co. By careful segregation of the inactive waste which would qualify for release into public domain as clearance level material, the volume of radioactive waste can be significantly reduced. The clearance material can be recycled further contributing to environmental impact reduction. The exemption criterion given by Atomic Energy Regulatory Board (AERB) for exemption of Bulk amount of solid materials[1] is 0.1 Bq/gm for 137Cs and 60Co radionuclides. For low level radioactivity measurements as per clearance criterion, detection methodology ideally should achieve 1/10th of the clearance value with 95% confidence. High resolution gamma ray spectrometry system consist of a p-type HPGe detector (Co-axial germanium crystal) having diameter of 68.5mm, a length of 51.8mm, an aluminium window of 0.6mm. This System has 50% relative efficiency with respect to 7.62 cm x 7.62 cm NaI (Tl). A bias voltage of +2500V is applied to detector. HPGe Detector system is housed in 100mm thick Lead Shielding of cylindrical shape with opening at top. Lead shielding is copper lined covered with Tin. Outer jacket of the shielding is low carbon steel. Minimum Detectable activity is defined by Gilmore, (2008)[2] as being the smallest quantity of activity that we are sure we can detect with a system in specific measurement conditions. Mathematical equation for the MDA by Currie[3] is as follows:



Where: B: peak area counts for time t sec, ε: Photo peak efficiency, γ: Gamma ray branching ratio and N: No. of the Channels in ROI and m is the no. of channels to the left and right of ROI used for background subtraction. Equation (1) is used in case no-peaked backgrounds and equation (2) has been used in case of peaked backgrounds. For estimation of MDA of this system, background counting time was set to 60000 Seconds, contact geometry of 20mL Glass Vial, efficiency curve has been plotted using standard sources viz. 133Ba,152Eu,137Cs and 60Co for same geometry and shown in [Figure 1]. MDA values for the long lived gamma emitters 137Cs and 60Co were estimated with 95% confidence level and found to be 1.51mBq and 4.49mBq respectively. Detector system with high detection efficiency and high gamma ray energy resolution results in low MDA values, which indicates its capabilities to measure low activity. Low level activity measurement methodology in this system with above MDA values provides confidence for estimation of the activity in the various systems of research reactor, which has to be decommissioned and released unconditionally in accordance with regulatory guidelines.{Figure 101}

Keywords: Decommissioning, HPGe, MDA


Atomic Energy Regulatory Board (AERB) Directive No. 01/2021.Gilmore G. Practical Gamma-ray Spectrometry. 2nd ed. John Wiley & Sons; 2008.Done L, Ioan MR. Minimum detectable activity in gamma spectrometry and its use in low level activity measurements. Appl Radiat Isot 2016;114:28-32.

 Abstract - 54520: Radionuclide characterization for dose control during handling of shielding section of irradiated fuel rods

S. Sajin Prasad, P. M. Yogesh, Lalit Vajpyee, Ranjit Sharma, Vikas Jain1

Health Physics Division, Bhabha Atomic Research Centre, 1Reactor Operations Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

The irradiated fuel assemblies which are removed from the reactor, are transferred to the Spent Fuel Storage Building (SFSB) for interim storage and cooling. The fuel assemblies are bisected in the general bay of SFSB. The fuel cluster portion containing stainless steel plug, aluminium shielding and a part of flow tube of fuel cluster is stored separately in general bay and subsequently sent to Waste Management Division (WMD) for reuse. The higher radiation field on the Al shielding section is allowed to reduce within the handling limits, by storing them underwater in the racks provided in the general bay of SFSB bay. They are taken out of the bay and flushed with water to remove transferable contamination and transported to WMD for further processing. The higher radiation field and radionuclide surface contamination on the shielding section poses a challenge during it's handling ,transfer and decontamination operations. This paper describes the characterization of radionuclide's present in the aluminium shielding section contributing to the occupational exposure during the transfer operations. The shielding section of fuel rods are made of Aluminium-1S of and elemental composition limits are specified for fuel fabrication, as per reactor material specifications standard. The aluminium of grade - 57S is used for the fabrication of split collar and flow tube of fuel cluster. When a fresh fuel rod assembly is loaded in the reactor, the split collar and top bulge portions are facing the reactor core and irradiated under the neutron flux ,during its residence time inside reactor .Due to the neutron activation of the elemental impurities in the aluminium, the radiation field on the split collar portion was found to be significant. The maximum radiation field on the shielding section is observed to be in the range of 10.0-15.0 mGyhr-1, four months after removal from reactor. At WMD, the bottom 1.20 metres of Al shield assembly ( consisting of flow tube ,spilt collar and a portion of Al shield) is removed by cutting and rest portion is kept for reduction in radiation field for a ~2 year period . Along with the neutron activation products, surface contamination due to fission products present in the bay water also contributes to the radiation field and contamination resulting in occupational exposure. In order to identify the main radionuclide's contributing to the radiation field, a small aluminium portion was cut from the split collar part of the shielding section. The sample matrix was decontaminated using EDTA solution, to eliminate the transferable contamination. The sample was transferred to 100ml of 5N NaOH solution and was allowed to dissolve completely. For characterization of radio nuclides present in the split collar sample matrix an HPGe detector based gamma spectrometer with suitable shielding and having relative efficiency of 50% and resolution of 1.9 keV at 1332 keV gamma energy of 60Co was used. The dissolved sample was prepared in standard glass vials of suitable geometry for high resolution gamma ray spectrometry.The energy and efficiency calibration of the detector system was determined from gamma rays emitted over the full energy range of interest from multi nuclide radioactivity sources namely 133Ba, 137Cs and 60Co for a wide energy range from 80 keV to 1332 keV. The energy efficiency values are fitted in Log polynomial of order 3 for a wide energy range up to 1.332 MeV.[1] The gamma spectrometric analysis result of the split collar sample is given in [Figure 1]. The gamma spectrometric analysis of split collar sample using HP Ge showed 51Cr,60Co and 65Zn as the major radio nuclides.60Co and 65Zn are long lived radio nuclides and contributes significantly to the residual radiation field, after storing the shielding section for two years also. The fission product radionuclide contamination in the spent fuel bay water and the activation products of on the shielding section are found to be the main sources of occupational exposure during the handling of the shielding section of fuel rod assembly. The radiation level can be reduced significantly by stringent quality control of impurity elemental composition in aluminium.{Figure 102}

Keywords: Characterization, dose, HPGe, split collar


Gilmore GR. Practical Gamma-Ray Spectrometry. West Sussex: John Wiley & Sons; 2008.

 Abstract - 54559: Simultaneously determination of ultra-trace level of U and Th in aqueous medium using Inductively Coupled Plasma – Optical Emission Spectrometer

S. K. Sahoo1, A. Gupta1, G. P. Verma1, J. S. Dubey1, S. K. Jha1,2, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Quantitative measurement of uranium and thorium is a fundamental requirement for nuclear fuel cycle facilities operation, radiation safety and assessment of radiation dose to occupational workers as well as members of the public. These two elements also govern the natural background radiation and exposure. Activity concentration and/or mass concentration (interrelated) of these two elements vary over a wide range depending on the matrices. There is gamut of analytical techniques available for simultaneous measurement of U and Th at higher concentration (Gamma Spectrometry, AAS, XRF) but the challenge in quantitative estimation at trace or ultra-trace level (ppb range). In the present study, an effort has been made to measure simultaneously U and Th at ppb level using Inductively Coupled Plasma – Optical Emission Spectrometer (ICP-OES). Parameters of the ICP-OES are optimised and three emission lines were selected to characterise Th (283.73 and 311.953 nm) and U (385.957 nm) in the aqueous solution (Manual of ICP OES). Mixed U and Th standards (5, 10, 15, 20 and 25 ppb) were prepared from the certified U (1001 ± 5 ppm) and Th (991 ± 4 ppm) stock solutions. All the dilutions were carried gravimetrically to reduce uncertainty in the dilution. Calibration of the ICP-OES was carried out sequentially using the freshly prepared U and Th standards. The calibration graphs are given in [Figure 1]a, [Figure 1]b, [Figure 1]c. The R2 (adj.) value for 283.73 nm (Th), 311.953 nm (Th) and 385.957 nm (U) calibration curves was found to be 0.99, 0.99 and 0.98 while the corresponding slope values was observed to be 68.6, 61.5, and 22.6 ints./ppb. Out of the two emission lines of Th, the Y-intercept of the linear calibration curve of the 311.953 nm was observed to be higher than 283.73 nm which indicate the later wavelength is more apropos for trace level measurement. After calibration of the ICP-OES, spike recovery studies were carried out within the calibration range to ensure the data quality [Table 1]. The percentage deviation was observed to be in the range of 1.3 – 9%, 2.4 – 10.7% and 5.4 -8.7% for 283.73 nm, 311.953 nm and 385.957 nm during measurement of Th and U in aqueous solution at ultra-trace level. The linear calibration curve and lower percentage deviation at ppb level are encouraging results and more field samples required to be analysed for development of a standard method.{Figure 103}{Table 33}

Keywords: Inductively coupled plasma – optical emission spectrometer, simultaneous measurement, Th, U


Manual of Analytical Jena PQ 9100 Elite ICP OES; 2022.

 Abstract - 54582: Numerical response of an oversquare LOAX HPGe detector for point source

Ram Sharma, M. Manohari1, O. Anna Lakshmi1, S. Murugan, G. Ganesh2, M. S. Kulkarani2,3

HPU, WIP, HPD, Bhabha Atomic Research Centre Facility, 1SQ&RMG, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, 2Health Physics Division, Bhabha Atomic Research Centre, 3Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Knowledge of full energy peak efficiency value is a prerequisite for the estimation of gamma activity of any samples using HPGe based gamma spectrometer. However, experimental determination of the efficiency of the HPGe detector is difficult in many circumstances due to the non-availability of the standard source of monoenergy, equivalent geometries and matrices, for short-lived radionuclides and for samples having energies greater than 1500 keV. In such cases, numerical simulations are the solutions. Since the characterization of each HPGe detector is unique, simulation of the individual detector's exact model of the detector is important for accurate quantitative measurement. So, in this work, efficiencies of an over-square p-type LOAX (low energy coaxial) HPGe detector were estimated for point sources at 1 cm & 19 cm in the energy range 60-3000 keV using the Monte Carlo FLUKA code. Efficiency curves were fitted using fifth-order polynomial logarithmic functions.

Materials and Methods: Experiment: Measurements were performed with point reference sources of 22Na, 137Cs, 133Ba & 241Am of activities (1.257±0.075), (44.5±2.5), (5.76±0.34), (407.4±24.3) kBq using a p-type coaxial HPGe detector of 8.45 cm dia and 3.03 cm thickness with Be entrance window of thickness 0.08 mm having 45% relative efficiency. In order to obtain a counting statistics error of less than 5% and minimize dead time and coincidence summing, the low active sources were kept at a distance of 1 & 3 cm, and the high active sources were kept at distances of 19 and 30 cm from the Be window.

Simulations: In order to estimate the low energy (<100 keV) efficiency exact modelling of the different components of the detector such as a dead layer, protective grove, bulletization, etc., is essential. A numerical model of the detector with all the above components optimized was constructed using the FLUKA code[1] and efficiencies were obtained using DETECT card for different energies.

Results and Discussion: Validation: The detector model was validated using experimental point source efficiencies. The efficiency of 60Co was corrected for coincidence. The agreement between the measured and simulated efficiencies is more than 80% [Figure 1], validating the FLUKA model of the detector.

Efficiency Curve up to 3 MeV: The simulated point source efficiencies were fitted using a fifth-order polynomial logarithmic function using equation 1 as shown in [Figure 2].


The efficiencies reduced by a factor of 2.17-2.27 throughout the energy range for 19 cm compared to 1 cm. The air attenuation is negligible in this energy range. So, this decrease is due to the solid angle subtended.

Conclusions: A computational model of the detector was constructed in FLUKA & was validated. Efficiency curves were established at two distances for point sources using the theoretically estimated values, using which efficiency value of in-between energies can be obtained. The fitted equation gives activity with 95% agreement for 133Ba using 302.852 keV, thus validating the fitted curve for 19 cm distance. The validated model can be applied to estimate efficiency when there is no standard source equal in terms of energy, geometry, density, and composition to the sample and can be applied to the human subject as well.{Figure 104}{Figure 105}

Keywords: Efficiency, HPGe detector, Monte Carlo simulation


Ferrari A, Sala P, Fasso A, Ranft J. Fluka: A Multi-Particle Transport Code. 2005.

 Abstract - 55103: Identifying process-based subsequence in the environmental dose rate time series of a nuclear facility using unsupervised machine learning algorithms

Anirudh Chandra, Shashank Saindane, S. Murali

Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

During normal operation of nuclear reactors, gaseous effluents that get generated, are released through the stack, within authorized limits. Among them is the radioisotope of argon gas – 41Ar,[1] which is typically generated in all reactor apertures where air gets irradiated with neutrons. Its dispersion around the reactor causes nearby gamma ray monitoring instruments to register higher dose rates which hampers the surveillance at the facilities and the site. This study focuses on the problem of autonomously identifying the process-based subsequences in the midst of 41Ar interference in the real-time gamma dose rate data recorded by area gamma monitors (AGMs) within a site housing a research reactor and an associated radioisotope extraction facility, as a part of the radiopharmaceutical supply chain. A 9-month dataset of recorded environmental dose rate data from two AGMs at a radioisotope extraction facility was used in this study. One AGM was located at the front and the other was located at the back of the facility. The data sampling rate of the AGM was 2 minutes and the entire time series was subjected to the workflow shown in [Figure 1]. An outlier detection technique based on median absolute deviation (MAD) was used to identify outliers in the time series and segments of such outliers were aggregated using a rule-base to form sub-sequences. For each of the sub-sequence, several features were created to increase the dimensionality of the dataset, such as – mean, standard deviation, peak value, rise time, fall time etc. Using these features, two unsupervised clustering algorithms[2] – k-means clustering and Hierarchical clustering– were used to identify those sub-sequences that corresponded to the actual process being monitored (k=1) such as – loading of radiopharmaceutical into vans, movement of source etc. and those that were non-process (k=2), i.e., 41Ar. A total of 404 sub-sequences (process + non-process) were identified and subject-matter experts (SMEs) were asked to label them as process or not, using the Delphi technique. The performance of each of the unsupervised algorithm on these sub-sequences is shown in [Table 1], against the results from the SME based evaluation. The identified processes are shown in [Figure 2]. The objective was to use the output of the unsupervised models to aid the SMEs in validating their original labels and identify any new sub-sequences.

The results were encouraging and will be used in generating a labelled dataset for supervised learning in order to develop a framework for reducing false alarms at portal monitors during radioactive source movement across sites.{Figure 106}{Figure 107}{Table 34}

Keywords: Machine learning, radiation monitoring, unsupervised clustering


Chatterjee MK, Divkar JK, Patil SS, Singh R, Pradeepkumar KS, Sharma DN. Study on enhanced atmospheric dispersion of 41Ar at the Trombay site. Radiat Prot Dosimetry 2013;155:483-96.Hastie T, Tibshirani R, Friedman J. The Elements of Statistical Learning – Data Mining, Inference and Prediction. New York: Springer-Verlag; 2017.

 Abstract - 55237: Finding time-efficient optimized pathway with minimum radiation exposure using deep reinforcement learning

Biswajit Sadhu1, Tanmay Sarkar1,2, S. Anand1,2, Kapil Deo Singh1, M. S. Kulkarni1,2

1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected], [email protected]

The guiding principle of radiation protection firmly stands on the optimization of radiation exposure to the human under the ALARA (As Low as Reasonably Achievable) concept. However, deducing an optimized pathway that satisfies ALARA often become challenging due the complex stochastic set up of routine radiological works, especially so in case of accidental scenarios.[1] The general pathfinding algorithms often requires deterministic information, therefore does not work well in an environment where source location and other attributing factors could be arbitrary. The recent advancement in deep reinforcement learning (DRL) shows promise on solving the optimal routing problem on dynamic unstructured environment.[2] In this work, we apply deep Q network (DQN), a subbranch within DRL to propose an algorithmic framework that designed on providing an optimized radiation exposure pathway in a time-efficient manner. Here, we use a toy model, 2D Grid World, which simulate the floor of a 10 × 10 unit2 room [Figure 1] with an entry and exit cell. In this grid, we place couple of radiation sources in specific grid cells. In DQN, an “agent” is allowed to understand the environment through incentives and penalties. DQN follows the formal framework of Markov Decision Process (MDP). Therefore, the agent's future only depends on the current state as a result agent maps its individual actions to learn an optimal policy leading to maximum incentives. Here, agent receives incentives for avoiding the exposure while getting closer to sources leads to penalties. Further, we penalize the agent for each move to introduce the time concept in the game. As such, this leverages us a gaming environment where the agent plays to find time-efficient path keeping exposure as low as possible. The agent can take any of the eight valid actions in the grid cell, namely left, right, down, up and four diagonal moves. In real scenario, the agent could be a human, robot, or drone. To find the best possible action at a given state in Grid World, we employ a deep neural network comprising fully connected layer and rectified linear unit (RELU) activation function. Further, we also utilized experience replay and target network to provide stability in the learning process. Among many environments in which we train the agents, we discuss here couple of them. In environment 1 [Figure 2]a, two sources are placed at (2,0) and (7,9). In this case, trained agent successfully finds an optimum route that traverse diagonally across the room. In environment 2 [Figure 2]b, where we place the radiation sources at (2,2) and (4,4) location, the trained agent delineates a different time bound route that avoid high exposure area. It may seem that following a path along the bottom and right walls in environment 2 may lead to minimum exposure but it will require more number of steps/time to exit the room. Rather, agent-derived path nicely balances both the time and exposure factor. Present study shows promising results from the preliminary investigations indicating the usefulness of DRL based architecture into the field of radiation protection. Optimization of hyper-parameters within the algorithmic architecture is presently under progress.{Figure 108}{Figure 109}

Keywords: Artificial intelligence, deep Q learning, deep reinforcement learning, radiation protection


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