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 Table of Contents 
Year : 2023  |  Volume : 46  |  Issue : 5  |  Page : 405-417  

Theme 7. Existing Exposures

Date of Web Publication7-Feb-2023

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Source of Support: None, Conflict of Interest: None

DOI: 10.4103/0972-0464.368746

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How to cite this article:
. Theme 7. Existing Exposures. Radiat Prot Environ 2023;46, Suppl S1:405-17

How to cite this URL:
. Theme 7. Existing Exposures. Radiat Prot Environ [serial online] 2023 [cited 2023 Mar 23];46, Suppl S1:405-17. Available from: https://www.rpe.org.in/text.asp?2023/46/5/405/368746

  Abstract - 71158: Measurement of radon removal amount of activated carbon Top

S. R. Kim1, G. S. Kim1, Y. J. Ahn1,2, T. H. Kim1,2

1Radiation Safety Section, Korea Institute of Radiological and Medical Sciences, Seoul, 2Department of Medical Physics, Korea University Sejong Campus, Sejong-Si, Republic of Korea

E-mail: [email protected]

The Korea Institute of Radiological and Medical Sciences has an experimental plan using 226Ra as a liquid target. The 226Ra produces a progeny, 222Rn; however, the production process could cause the 222Rn to leak into the laboratory. To reduce the internal exposure of laboratory workers, a facility to remove any leaked 222Rn is required. This study seeks to evaluate and quantify the effectiveness of activated carbon in reducing 222Rn, a gaseous radioactive material.

The radon source was produced as follows:

  • Stones were collected near an abandoned uranium mine in Boeun-gun, Chungcheongbuk-do (36.44641° N, 127.60244° S).
  • The stones were crushed and dried at 105 °C for 4 h.
  • The powdered stone chippings were sealed in a plastic chamber connecting a 20 ℓ plastic chamber and stored for 3–5 days, where the radon maintained a concentration of 2,000–2,500 Bq/m3.

Radon was detected using a RAD7 (Durridge, U.S.A) radon detector, which has a built-in air pump with a flow rate of 1 ℓ/min, and the provided desiccant was used to maintain a relative humidity of 4% or less in all experiments. Filters were made and tested using a coconut-based activated carbon manufactured by NAC Co., Ltd. In cases where the activated carbon used was 5 g or less, an experimental filter was made using a 10-cc plastic tube, and in cases where the activated carbon exceeded 5 g, the experimental filter was made using a 25-cc plastic tube. The experiment was conducted by filling a sealed chamber with radon gas as shown in [Figure 1]a and, subsequently, attempting to remove the gas using the experimental filters as shown in [Figure 1]b. The change in the radon concentration and the amount of radon removed were measured, as shown in [Figure 2]a and [Figure 2]b, respectively. As it was a practical experiment, the decay rate of radon, which is significantly smaller than the removal rate by the filter, was not considered. For quantitative comparison, the amount of radon removed per unit weight of activated carbon was obtained, as shown in the secondary axis of [Figure 2]b, using the amount of radon removed over 5 hours suggested removal times for practical purposes. The experimental results dependent on the flow rate, the particle size of the activated carbon, reuse of activated carbon, and whether Triethylenediamine (TEDA) was attached are shown in [Figure 3].

From the experiment it can be shown that the experimental filter has the following characteristics:

  • Radon removed by the filters was 2 kBq per kg of activated carbon, regardless of the size of the particles or whether TEDA was attached.
  • The filters can be reused after 29 days, which results in a similar performance to first-time use. However, because this result was based on only two experiments, supplementation through additional experiments is necessary.

This study was supported by a grant of the Korea Institute of Radiological and Medical Sciences (KIRAMS), funded by Ministry of Science and ICT (MSIT), Republic of Korea (No. 50422-2022 and No. 50606-2022).
Figure 1: Experiment Schematic

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Figure 2: Radon concentration in the chamber and amount of radon removed

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Figure 3: Amount of radon removed under various conditions

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Keywords: Active carbon, filtration, radon, radon removal

  Abstract - 71175: Influence of thoron daughters on air monitoring program of CORAL reprocessing facility Top

A. Dhanasekaran, K. Jothi, Abhishek Maurya, Sunil, 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]

Compact Reprocessing of Advanced fuels in Lead cells (CORAL) is situated at Kalpakkam and reprocesses the spent fuels discharged from the Fast Breeder Test Reactor (FBTR). The facility has a well-defined air monitoring program with Gross α Continuous Air Monitors (CAM), Alpha spectrometry-based CAM and size-selective Spot Air Samplers to detect the very meager level of artificial radioactivity releases. The natural radon and thoron daughter product concentration in ambient air significantly influences the artificial radioactivity's Minimum Detectable Concentration (MDC). CORAL is situated in the Monazite-rich coastal area and shares a common area with a facility where Thorium Rods from Cirus and Dhruva were reprocessed. Thus the thoron daughters, mainly the ThB (212Pb) concentration, has a more significant influence than the radon daughter concentration. A study is conducted to quantify the ThB concentration and its impact on the air monitoring program of CORAL. The air samples are collected from Emergency Assembly Area (EAA), CORAL Operating Area (COA), Active Analytical Laboratory (AAL) and JRod Operating Area (JOA) using Staplex air samplers for 10 m. In each area, two air samples are collected, one with the Large Area filter paper (LA) and the other with the Size Selective (SS) sampler attachment and more than 50 such samplings are carried out for each area. The samples are counted in a pre-calibrated alpha counting system (ACS) for 10 m with the delay of 0.5, 5.0 and 168 h. The gross alpha concentrations estimated using the 0.5 h delay counting data are used to estimate the collection efficiency of natural aerosols in a size-selective sampling arrangement.

Where GAC is the gross alpha concentrations of SS and LA. [Table 1] shows the CE of the size-selective sampling arrangement for natural aerosols. It is observed that for 0.5 h delay counting, the device has the CE 1.21 – 1.47% as average CE with relative standard deviation (RSD) <100. Any abnormal ratio observed in the counting would indicate the presence of artificial radioactivity. The ThB concentration is estimated with 5.0h delay counting data using the standard radioactive equilibrium equations. [Figure 1] shows the histogram of the ThB concentrations observed in different areas. The EAA always showed minimum ThB concentration, whereas the JOA showed the maximum. COA and AAL having the common area exhaust showed a resembling histogram. [Figure 2] shows the simulated eq. DACh values for the Max., Av. and Min. ThB concentration in the COA and AAL. The Max. ThB concentration can generate a background equivalent to ~30 -35 DACh of artificial radioactivity in CAM. The secondary axis of [Figure 2] shows ASCAM MDA and even with the Max. ThB concentration, the ASCAM can detect <1.5 DACh. A similar interference is unavoidable, while CAM filter papers delay counting. Even adopting a standard two-count method, the MDA for artificial alpha activity remains higher and varies with the average ThB concentration during the sampling period. It is concluded that the ASCAM is necessary for the high natural background area for early artificial radioactivity detection. The size-selective sampling reduces the ThB interference in the spot air sampling.
Table 1: CE observed in different areas of the facility

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Figure 1: ThB concentration histogram

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Figure 2: Simulated DACh for CAM and MDA for ASCAM

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Keywords: Air monitoring, centripeter, grab sampling, thoron, work level

  Abstract - 71403: Uncertainty analysis of etching characteristics for SSNTD based DRPS using Th-229 Source Top

R. Prajith1, R. P. Rout1 , R. Mishra1,2, S. Jalaluddin1, A. T. Khan1, B. K. Sapra1,2

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

E-mail: [email protected]

LR115 SSNTD based DRPS are used widely for assessment of inhalation dose due to the 222Rn progeny in indoor and occupational environments.[1] Chemical etching of the SSNTD films used in DRPS is an important procedure that affect the track density and thereby the inhalation dose assessment. The standard etching protocol of 2.5N NaOH at 60° C for 90minutes is followed to process the films to arrive at the activity concentration of progeny.[2] In the present study we have quantified the uncertainty due to small variation in etching temperature and time in DRPS. Exposures were carried out in a desiccator with 229Th Planchet Source having multiple alpha energies of 4.9, 5.8, 6.3, 7.1, and 8.4 meV which coincides with the energy of the alphas emitted in the 222Rn decay series. The distance between the source and the DRPS was theoretically calculated and kept fixed for all the exposures. Three sets of experiments were carried out as follows: (i) Variation of track density with etching temperature: Exposed DRPS were chemically etched at different temperatures at constant etching time of 90 mins. The variation of track density and bulk etch rate with respect to temperature are given in [Table 1]. It was observed that even for one degree variation in temperature, the variation in track density was around 12%. (ii) Variation of track density with etching time: Etching of exposed DRPS were carried out at different times at constant etching temperature of 60°C. The variation of track density and bulk etch rate with respect to etching time are given in [Table 2]. It was observed that variation of 5mins in etching time gave rise to only 2% variation in track density. (iii) Intermittent etching: In a bid to check the effect of a break in etching procedure owing to some emergency, two combinations of time break were considered. In one set, films were etched at 60 mins and then for 30 mins and in the second set the timings were reversed. It was observed that there was negligible effect on the track density due to intermittent etching. Hence it may be concluded that the track density is sensitive to even slight variation in etching temperature compared to etching time.
Table 1: Variation of track density and bulk etch rate with respect to temperature

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Table 2: Variation of track density and bulk etch rate with respect to etching time

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Table 3: Variation of track density with time during intermittent etching

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Keywords: Bulk etch rate, chemical etching, DRPS, track density

  References Top

  1. Mishra R, Rout R, Prajith R, Jalalluddin S, Sapra BK, Mayya YS. Innovative easy-to-use passive technique for 222Rn and 220Rn decay product detection. Radiat Prot Dosimetry 2016;171:181-6.
  2. Eappen KP, Mayya YS. Calibration factors for LR-115 (type-II) based radon thoron discriminating dosimeter. Radiat Meas 2004;38:5-17.

  Abstract - 71404: Estimation of dose conversion factor of thoron progeny using ICRP 130 HRT model Top

R. P. Rout1, R. Mishra1,2, A. T. Khan1, B. K. Sapra1,2

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

E-mail: [email protected]

It is important to measure effective dose to lungs due to inhalation of thoron (220Rn) progeny as it may leads to development of lung cancer in humans. Inhalation dose due to 220Rn progeny is generally estimated from its progeny concentration using the UNSCEAR proposed Dose Conversion Factor (DCF). In the present study, DCF of 220Rn progeny was estimated from the Human Respiratory Tract (HRT) model proposed by ICRP 130.[1] In the recent ICRP 130 model, the clearance model is revised in which the 5 regions of HRT (ET1, ET2, BB, bb and AI) are divided into 11 compartments in contrast to the 14 compartments proposed by ICRP 66.[1] The clearance rates by particle transport (mi) and absorption to blood (Si) from each of the compartments i has also been revised [Figure 1]. Taking into consideration the revised compartmental model, a code has been developed in Mathematica for calculation of equivalent and effective dose to different regions of HRT and also the DCF which is effective dose per unit concentration per unit exposure time. Assuming breathing rate (BR) to be 0.75 m3 h-1 for general public,[2] bimodal size distribution for thoron progeny with unattached and attached fraction size to be 1 nm 200 nm respectively, activity (Ai) deposited in each of the compartments i, was calculated using Eq 1.[3]

Where, is the nominal intake activity and DE is the deposition efficiency in each compartment. Subsequently, the remaining activity after clearance was calculated using the following equation,

Where, A'i is the deposited activity (Ai) in each compartment i due to inhalation and also it Includes particles cleared to the compartment i from another compartment j. λR is the radioactive decay constant. Committed Equivalent dose (HT) and effective dose (E) to lungs are then calculated using the following expressions,

Where, S and T represent the source and target, is the total number of nuclear transformations of radionuclide j in S over a period of 50 years following the intake, SEE is the specific effective energy of the alpha, beta, gamma radiations emitted by the 220Rn progeny, is the tissue weighting factor of lung. Using all the above equations in the code, DCF for an adult male was found to be ~36 nSv h-1 per Bq m-3 which was observed to be in good agreement with that determined from the commercially available LUDEP code and also the UNSCEAR proposed value [Table 1]. With BR of 1.2 m3 h-1 and attached size of 100 nm, DCF was found to be 92 nSv h-1 per Bq m-3, very close to the ICRP 137 recommended value.
Table 1: Estimated values of dose conversion factor of 220Rn progeny

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Figure 1: Revised compartmental model for clearance of particles from HRT by ICRP 130

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Keywords: Compartmental model, DCF, HRT, Thoron

  References Top

  1. ICRP Publication 130. Occupational Intakes of Radionuclides: Part-1. Ann ICRP; 2015. p. 44.
  2. Porstendorfer J. Radon: Measurements related to dose. Environ Int 1996;22:S563-83.
  3. Rout RP. Thesis, HBNI; 2021.
  4. UNSCEAR. Sources and Effects of ionizing Radiation. Annexure B; 2000. p. 107-8.

  Abstract - 71457: Radon and thoron measurements at glass fiber reinforced gypsum–based model buildings at Kalpakkam DAE complex Top

N. Chitra, H. Krishnan1, S. N. Bramha, K. S. Briteena, S. Chandrasekaran, C. V. Srinivas, B. Venkatraman

Environmental Assessment Division, 1Radiological and Environmental safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Glass fiber reinforced gypsum (GFRG) wall, a new composite wall product known as Rapidwall/ Gypcrete in the industry, is made essentially of gypsum plaster, reinforced with chopped glass fibers .Phosphogypsum is classified as TENORM (Technologically Enhanced Naturally Occurring Radioactive Material), which is a solid waste containing heavy metals and naturally occurring radioactive elements from the rock matrix. Characterization studies[1] have been done to quantify the radionuclide contents of Phosphogypsum. The results indicate that that 226Ra and 232Th are found with concentrations 100-500 Bq kg-1 and 6-25 Bq kg-1 respectively. So, radon and thoron emanations from GFRG cannot be ruled out. The present radiation monitoring study is carried out in model GFRG buildings at Edaiyur and Kunnathur army base situated in the Kalpakkam DAE complex. The study includes measurements in two phases of three month duration each. During Phase -1 (Oct- Dec, 2021), the buildings were occupied, with typical indoor ventilation conditions- a combination of both air conditioned and natural ventilation and in Phase -2 (Jan-Mar, 2022) there was no ventilation as well as no occupancy. Pin-hole dosimeters and (Direct thoron progeny sensor) DTPS/ (Direct Radon Progeny Sensor) DRPS) badges designed and developed by BARC were used for the measurement of radon/thoron gases and their progenies. 30 locations were chosen within the rooms of these two buildings and 2 pin hole dosimeters along with 2 DTPS/DRPS badges were deployed in each location for three months at least 2 ft away from walls. The annual effective inhalation doses are estimated using the dose conversion factors (DCF) reported by UNSCEAR (2000). Annual eff. dose(radon) = EERC × DCF × occupancy factor x No of hours in a year. One room (Edaiyur Dormitory) was chosen for further analysis. The radon contribution is from the walls and the ceiling. For no ventilation condition the wall flux was calculated by using the measured indoor radon concentration. With this flux, the ventilation rate of Phase -1 scenario was calculated.

Results and Discussion: The radon and thoron concentrations in Phase-1 range from 9± 1 to 54 ± 4 Bq m-3 and 15 ± 1 to 186 ± 6 Bq m-3 respectively. The radon and thoron concentrations in Phase-2 range from 21 ± 1 to 449 ± 34 Bq m-3 and 22± 4 to 188 ± 24 Bq m-3 respectively. The annual effective dose calculated from the measured Equilibrium equivqlent radon concentration and thoron concentration (EERC and EETC) values in each room for Phase 1 and 2 are given in [Table 1] and [Table 2]. The annual effective doses due to thoron progeny for the two phases are similar in values whereas the doses due to radon progeny for the two phases show extreme variation. Thoron progeny with higher deposition rate are independent of ventilation rate. The source (GFRG) radon flux was found to be 1.39 ± 0.03 Bq m-2. The ventilation rate for the Phase -1 exposure condition was found to be 0.12 h-1.
Figure 1: Comparison of EERC values. EERC: Equilibrium equivqlent radon concentration

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Keywords: Glass fiber reinforced gypsum, radon, thoron, ventilation rate
Table 1: Progeny concentrations and annual inhalation dose in Phase-1

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Table 2: Progeny concentrations and annual inhalation dose in Phase-2

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  Reference Top

  1. Gezer F, Turhan Ş, Uğur FA, Gören E, Kurt MZ, Ufuktepe Y. Natural radionuclide content of disposed phosphogypsum as TENORM produced from phosphorus fertilizer industry in Turkey. Ann Nucl Energy 2012;50:33-7.

  Abstract - 71484: Radon concentration measurements in water bodies at Kalpakkam DAE complex Top

N. Chitra, S. Chandrasekaran, 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]

Radon concentration in water cannot be neglected owing to its relatively higher solubility compared to the other noble gases. Nevertheless, the agitation of water due to human domestic activities can lead to increased liberation of radon to the air medium. Radon concentration in ground water depends on the parent material (rock and mineral composition) of the geographical terrain. Generally surface waters namely lakes and ponds have low radon concentration owing to the aerodynamic effect by winds facilitating escape of radon into the atmosphere. In the present study, the radon concentration in samples collected from water bodies within Kalpakkam DAE complex, were estimated. The water samples were collected from open wells, borewells, lakes, ponds and reservoirs. Kalpakkam receives rain fall mainly during the north east (NE) monsoon (Oct-Dec). Hence, the study was carried out in two phases-Pre-NE monsoon (September) and Post –NE monsoon (December). Water samples were collected in 1l PVC bottles and brought to the lab without shaking so that turbulence in the samples is prevented. This minimizes the radon loss from the samples due to its escape into air. The radon measurement in the samples is carried out by using RAD H2O water accessory of RAD7 radon online monitor. RAD H2O method employs a closed loop aeration scheme whereby the air volume and water volume are constant. The air re-circulates through the water and continuously extracts the radon until a state of equilibrium develops. The radon concentration in the water samples collected from the water bodies located within Kalpakkam DAE complex is given in [Table 1]. Phase-1 (September -2021) 11 samples showed below detectable limit values for radon activity (MDA-114 Bq m-3). In the remaining samples, the radon activity concentration ranged from 200 ± 95 to 3115 ± 708 Bq m-3( 0.200 – 3.115 Bq l-1). Phase-2 (December-2021) 9 samples showed below detectable limit values for radon activity (MDA-114 Bq m-3). In the remaining samples, the radon activity concentration ranged from 130 ± 50 to 445 ± 78 Bq m-3 ( 0.13 – 0.44 Bq l-1). From the above, it is seen that radon levels in all the analyzed samples are below the prescribed limits. The radon concentrations in Phase -2 samples of post-monsoon are considerably lower compared to the pre-monsoon. This could be due to the increased rain water collection in the surface water bodies as well as borewells during the NE monsoon and resultant dilution.
Table 1: Radon concentration in water bodies for phase -1 and 2 measurements

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Keywords: Closed loop aeration, radon, Water bodies

  Reference Top

  1. Singh B, Kant K, Garg M, Singh A, Sahoo BK, Sapra BK. India J Radioanal Nucl Chem 2019;319:907-16.

  Abstract - 71523: Spatial distribution of 222Rn progeny in a walk-in type chamber: A comparison of experimental and calculated results Top

A. P. Vijith, Rosaline Mishra1,2, N. Karunakara

Centre for Advanced Research in Environmental Radioactivity, Mangalore University, Mangalagangotri, Karnataka, 1Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: The deposition and somatic transfer process of 222Rn progeny due to inhalation of radioactive aerosols contribute a major fraction to the inhalation dose. The attachment and deposition mechanism of 222Rn progeny atoms on the aerosol are sensitive to various environmental parameters such as temperature, humidity, ventilation rate etc. A calibration chamber enables us to conduct experiments to understand the dependence of 222Rn progeny attachment on aerosols on environmental parameters. In this work we studied the spatial homogeneity of 222Rn and progeny concentrations in the 22,000 L walk-in type 222Rn calibration chamber at the Centre for Advanced Research in Environmental Radioactivity (CARER), Mangalore University [Figure 1]. The experimental results were compared with the calculated values.[1]

Materials and Methods: The flow mode integrated samplers based on direct 222Rn progeny sensors[2] were placed at various monitoring positions inside the walk-in type calibration chamber to study the spatial distribution of 222Rn progeny at a steady 222Rn gas concentration of 3000±250 Bq m-3 by adopting semi-dynamic injection algorithm of soil-gas.[3] The exposure conditions maintained during the experiment is given in [Table 1]. The sampling duration was 30 minutes for every 3 hours interval. Then the detectors were processed following the established method.[2] The experimentally determined progeny concentrations using flow mode integrated sampling were compared with the values calculated from the knowledge on the 222Rn concentration and equilibrium factor.[1] Scintillation cell-based 222Rn monitor (Smart RnDuo, AQTEK SYSTEMS, India) and ionization chamber-based monitor (AlphaGuard PQ2000PRO) with a detection range of 0.008 - 50000 kBq m-3 and 0.002 – 2000 kBq m-3 respectively were used for the 222Rn measurements.

Results and Discussion: The experimentally obtained progeny concentrations were compared with the calculated values [Figure 2]. The average of deviation between the two set of values was ~11% and that is well within the uncertainty of measurements.
Figure 1: 222Rn chamber at CARER, Mangalore University

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Figure 2: Comparison of theory and experimental data

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Table 1: Exposure conditions maintained during the experiment

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  References Top

  1. Sapra BK, Sahoo BK, Mishra R. Handbook on Radon Transport Model and Measurement Methods RP and AD Group. Mumbai: BARC; 2010.
  2. Mishra R, Sapra BK, Mayya YS. Nucl Instrum Methods Phys Res B 2009;267:3574-9.
  3. Trilochana S, Somashekarappa HM, Sudeep Kumara K, Mayya YS, Karunakara N. J Radioanal Nucl Chem 2020;323:507-13.

  Abstract - 71599: Radon survey in a healthcare facility Top

C. Jeffries

Medical Physics and Radiation Safety, South Australia Medical Imaging, Adelaide, Australia

E-mail: [email protected]

Radon is a naturally occurring radioactive gas that occurs everywhere. Elevated concentrations of radon gas are associated with increased health risks and generally occur in workplaces or areas with elevated concentrations of uranium or thorium, such as in mines. Recently, there has been an increased recognition that radon is potentially a wider health risk and the International Atomic Energy Agency (IAEA) has established exposure standards through its Basic Safety Standards.[1] These require consideration of exposure to 222Rn in all workplaces, including those that would not normally be considered. In Australia, the reference levels for radon concentration are 200 Bq.m-3 for dwellings and 1,000 Bq.m-3 for workplaces. Note that the workplace reference level equates to approximately 20 mSv/y for full time (2,000 h/y) exposure. The work described in this paper was undertaken at a metropolitan teaching hospital that is a major trauma centre and offers a full range of healthcare services. The nuclear medicine department undertakes radioiodine (131I) therapy treatments for patients. Iodine contaminated waste resulting from therapy is held in storage to allow for radioactive decay (t1/2 = 8 days) before disposal. The radioactive waste is stored either in the nuclear medicine department or in a dedicated radioactive waste store. The waste store is located on the lower level of the hospital and is partially open to bare earth within a fully enclosed area. In 2020, the radioactive waste store was inspected by the Radiation Safety Officer (RSO), with a focus on the potential for elevated airborne radon concentrations due to poor air quality and limited ventilation. Continuous radon monitoring was implemented in June 2020 using an FT Lab Radon Eye (Model RD200, Serial No. RE21703060023) which logs radon concentrations hourly. The mean, geometric mean, minimum and maximum radon concentration were 145.7, 137.6, 44.0 and 381 Bq.m-3 respectively. Radon monitoring was continued in the waste store using the Radon Eye due to the elevated radon concentrations. Disruption to healthcare services during the Covid-19 pandemic allowed contaminated waste to be managed entirely within nuclear medicine hot-lab. This continued until August 2022 when radioiodine therapy activity increased, resulting in the need to manage the radioiodine waste in the waste store. Monitoring data was retrieved from the Radon Eye in early August and analysed. A summary of radon concentrations by calendar month is shown in [Figure 1]. The maximum radon concentration was 519 Bq.m-3. Exposure to radon in the waste store can be effectively controlled by restricting. Nuclear Medicine staff require access to place new contaminated material, or to remove decayed waste. The maximum exposure time is twenty hours per year. An interim control allows access to the waste store when the radon concentration is less than 400 Bq.m-3. Approximately 1% of results exceeded this level. The radiation dose is approximately 0.07 mSv per year in the maximum exposure case (20h, 400 Bq.m-3) based on a dose conversion factor of 9 nSv/Bq.h.m-3.[2] Radiation dose will be lower due to the use of PPE (i.e. masks) for Covid-19. Elevated radon concentrations have been unexpectedly detected in hospital. An interim control has been introduced to radon exposure. Permanent mitigation measures will be investigated in the future.
Figure 1: Summary by month showing minimum, first quartile, median, third quartile and maximum radon concentration in radioactive waste store

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Keywords: Exposure controls, healthcare, nuclear medicine, occupational exposure, radon

  References Top

  1. International Atomic Energy Agency. Radiation Protection and Safety of Radiation Sources: International Basic Safety Standards. Vienna: International Atomic Energy Agency; 2014.
  2. UNSCEAR. Sources, Effects and Risks of Ionizing Radiation, United Nations Scientific Committee on the Effects of Atomic Radiation, New York; 2019.

  Abstract - 72507: Emergency preparedness for BARC, Visakhapatnam: Environmental gamma dose rate assessment Top

P. Padma Savitri, J. Sudhakar, R. Balaramkumar, B. Ramesh, A. Vinod Kumar

Environmental Monitoring and Assessment Division, BARC, Visakhapatnam, Andhra Pradesh, India

E-mail: [email protected]

Environmental dose rate recording and analysis is essential before commissioning of any nuclear facility. The baseline radiation levels has very important role in the assessment of the radiological impact during the operation of plant and any emergency situation. This paper presents the analysis of gamma dose rate data collected regularly in and around up to 30km radius of BARC, Visakhapatnam during 2016-2021 covering all sectors by using different state of the art monitoring instruments. As a part of emergency preparedness radiation mapping is carried up to around 100km from project site covering various locations in the Visakhapatnam and Vizianagaram districts and near beach areas in a mobile monitoring vehicle. The external radiation levels are measured regularly using Thermo scientific Micro Rem tissue equivalent survey meter in around 20 locations within BARC and around 100 locations outside BARC up to 30km radius. Cumulative radiation levels are monitored using Environmental TLD's deployed in different locations. Dose rate is monitored continuously by Indian Environmental Radiation Monitoring Network (IERMON) system and Area gamma Monitors (AGM) located at fixed locations within and outside site. Radiation levels analyzed over a period of six years is presented in [Table 1] and [Table 2]. Radiation levels over the last six years within BARC is in the range 0.08-0.7 μGy/hr with the average 0.26±0.14 and outside upto 30km in the range 0.03-0.6 μGy/h. with the average 0.19±0.13. During the same period within site dose levels recorded by IERMON and AGM's are in the range 0.09 to 0.14 μGyh-1, 0.04 to 0.6 μGyh-1 and outside site 0.04 to 0.5 μGyh-1, 0.08 to 0.3 μGyh-1 respectively. This analysis indicates over the last six years radiation levels recorded in different systems are in the same range within statistical variation. Over the years little increase in the radiation levels at some locations could be due to i) increase in construction activities due to expansion within site and ii) increase in the area of survey and number of locations outside the site. The observed levels in the present study are in the comparible range 0.05-0.25 μGy/h with an average value 0.1 ±0.03 reported for Visakhapatnam vide,[1] and for India 0.02 to 1.1 μGy/h.[2] Maximum dose rate 2.0 μGy/h at a distance of 100 km from the site is observed in Garbham village of Vizianagaram district. Radiation levels 0.08-15 μGy/h was recorded in special survey conducted at beach areas of Visakhapatnam. High values are not included in [Table 2]. The higher gamma radiation level regions indicated likely hood of localized pockets of terrestrial radionuclides. Detailed investigation has to be done in those high background locations. The extensive monitoring carried out in and around BARC, Visakhapatnam and the analysis of data collected gives the radiation exposure to public due to natural background radiation. Based on the baseline studies carried out and the state of the art systems currently available for emergency management, prediction and assessment of Gamma dose rate in the public domain following any accidental releases is possible. This study will help in strengthening the emergency preparedness programme.
Table 1: Dose rate (μGy/h) within Bhabha Atomic Research Centre (V)

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Table 2: Dose rate (μGy/h) outside Bhabha Atomic Research Centre (V)

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  References Top

  1. Vinod Kumar A, et al. Assessment of Environmental Gamma Radiation and Radon Levels in and around the New BARC Campus. Visakhapatnam: BARC/2010/I/003.
  2. Nambi, et al. Natural Background Radiation and Population Dose Distribution in India. Health Physics Division, BARC, Report; 1986.

  Abstract - 73268: Investigation of LET spectra at low earth orbit by Monte Carlo simulations Top

Sandipan Dawn1,2, A. K. Bakshi1,2, B. K. Sapra1,2

1RP&AD, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

With advancement of the new technologies, human space activities are increasing manifold, and India is gearing up to undertake such missions (Gaganyaan mission, ISRO). For such human activities, the most preferred region is Low Earth Orbit (400 km a.s.l). Astronauts at LEO are exposed to a significant radiation field due to the Galactic cosmic rays (GCR) resulting from weaker geomagnetic fields. GCR consists of 87% proton, 12% 4He, and 1% heavy high energy ions typically up to atomic number of 28 with peak energies around 1 GeV/u.[1] Aluminium, carbon fibre reinforced polymer, etc. are used as the spacecraft wall materials which are not much effective for radiation shielding. Under the present study, PHITS Monte Carlo simulations have been carried out to investigate LET spectra, radiation quality factor (Q̄), absorbed dose and dose equivalent inside a typical spacecraft with aluminium as shielding material, at LEO. The following points have been considered in the simulations:

· Charged particles from atomic number 1 to 28 have been considered with actual flux weightage.

· Concentric spheres of 2.54 cm Aluminium and 15 cm water has been considered as irradiation geometry.

· The complete geometry is uniformly irradiated by the free space GCR spectra from all directions.

· Lineal energy spectra (y*d(y)) have been scored for different charged particles inside the phantom.

· Absorbed dose, dose equivalent, Q̄ have been derived from Lineal energy spectra.

Table 1 represents the total absorbed dose and dose equivalents due to all the charged particles at the centre of the water phantom. In addition, some of the particles which are contributing significantly to the total dose equivalent are also shown in [Table 1] separately. [Figure 1] shows the differential particle energy distribution inside the phantom. Except neutron, they have peak energies ~ 1GeV/u. It can be noted that neutrons have been produced by the interaction of the GCR particle with wall and phantom materials. [Figure 2] represents the cumulative y*d(y )spectra along with un-normalised spectra for different ions. It can be clearly seen that the cumulative spectra is dominated by low LET particles below 10 keV/μ which is mainly contributed by proton and alpha. Calculated average Q̄ value in the present study is found to be 2.57 which matches well with the TEPC measured value at ISS2.
Figure 1: Particle spectra due to GCR inside the space craft at LEO

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Figure 2: Lineal energy spectra due to GCR inside space craft at LEO. GCR: Galactic cosmic rays

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Table 1: Details of total absorbed dose and dose equivalents due to different charged particles at the centre of the water phantom inside space craft

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Keywords: Cosmic ray, LEO, LET, lineal energy spectra

  References Top

  1. Beringer J. Particle Data Group. Phys.Rev D 2012;86:010001.
  2. Zhou D, et al. Radiation measured for ISS-Expedition 12 with different dosimeters. NIM A 2007;580:1283-9.

  Abstract - 73561: Evaluation of terrestrial and cosmic components of external exposure in Rawatbhata Rajasthan site environment Top

S. K. Goyal, A. K. Gocher, S. N. Tiwari, I. V. Saradhi1, A. Vinod Kumar1

Environmental Survey Laboratory, EMAD, BARC, Rawatbhata, Rajasthan, 1Environmental Monitoring and Assessment Division, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Quantification of natural and anthropogenic radioactivity present in the environmental matrices around NPP Site is essential for evaluation of environmental impact of releases from the station. Primordial radioactivity of 40K, 238U and 232Th and their radioactive decay products present in the soil and the cosmic radiation from outer space are main sources of external exposure to humans. External environmental dose is directly measured by installing TLDs at different locations around NPP site. In the present study, an attempt is made to estimate the external radiation dose due to natural radioactivity present in the soil of Rawatbhata environment. Cosmic ray component is estimated using correlation study between estimated external exposure due to natural radioactivity and measured environmental TLD dose. The concentration of 238U, 232Th and 40K in the soil samples was determined employing gamma spectrometry using standard methodology. The counting efficiencies for different gamma energies and sample geometries were obtained by counting certified reference materials from IAEA (RGU-I and RGTh-I). Thirty-five soil samples were collected from seven different locations around RAPS site. The homogenized sample is filled, sealed and stored for a period of one month for equilibration of daughter products of uranium and thorium and further counted in HPGe detector for 60, 000 sec. for estimation of uranium, thorium and K-40. The external doses at 1 m due to the presence of 1 Bq/kg each of U, Th and K-40 in soil is estimated using standard methodology as per UNSCEAR 2000.

Geometric mean radioactivity of 238U, 232Th and 40K and corresponding external dose at different locations of Rawatbhata are given in [Table 1]. For evaluation of cosmic ray component, the external doses recorded by environmental TLDs are plotted against evaluated external doses at respective locations. The best fit curve obtained is given in [Figure 1]. Regression analysis between the dose by natural radioactivity of soil on X-axis and corresponding mean annual exposure of TLD on Y-axis is carried out and obtained best fit curve is given as Y = X + 0.37 The intercept value is used to determine the cosmic ray component which comes out to be 0.37mGy/y and is comparable to the reported value for Indian environment.[1] In addition, the ambient gamma radiation levels recorded at 1-meter height using portable radiation survey meters (Ludlum, model 19 micro R-meter) along with the navigation details. Terrestrial and cosmic radiation levels were also computed using the radiation survey data as per standard methodology. This value agrees well with cosmic component evaluated by previous method.
Figure 1: Comparision of TLD's and calculated NRA present in soil

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Table 1: Natural radioactivity content in Rawatbhata soils and corresponding dose

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Keywords Cosmic dose, daughter product, external dose, K-40, Natural radioactivity, Th-232, U-238

  References Top

  1. Krishnamoorthy TM, et al. Environmental Radiation Exposure: Natural and Man Made. Vol. 15. INCAS Bulletin; 1999. p. 25-35.
  2. Ramkumar S, Dole MU, et al. Evaluation of Terrestrial and Cosmic Components in KAPS Region. NSE; 2001.
  3. Verma GP, Sahoo SK, et al. Spatial Distribution of Terrestrial and Cosmic Radiation around Upcoming fuel Fabrication Facility at Rawatbhata, Rajasthan and Dose Assessment. NUCAR; 2021.

  Abstract - 74162: Estimation of natural radioactivity in granite used as building materials in Kanyakumari District, Tamil Nadu, India Top

F. S. Karolin Mary, G. Shanthi

Department of Physics, Women's Christian College, Nagercoil, Tamil Nadu, India

E-mail: [email protected]

Introduction: Granite is a natural source of radiation, like most natural stones. Granites are used as building materials, decorating and expensive materials due to its hardness, elegant look and availability of different colors. Therefore, it is important to measure the concentration of radionuclides in granite before selecting for use in interior decorations in house, as the inhabitants spend about 80% of their time indoors.[1] The objective of the study are (i) To determine the specific radioactivity concentrations of 238U, 232Th and 40K in granites used as building materials in Kanyakumari District (ii) To assess the possible radiological risks to human health and to compare the results with the UNSEAR data.

Materials and Methods: A total of 10 granite samples were collected from several local suppliers in Kanyakumari District. The collected samples were crushed and milled to fine powder, homogenized and filled in 100 ml bottle for gamma ray spectrometry. The concentration of natural radionuclides in the samples were determined using a 3“×3” NaI (Tl) gamma ray spectrometric system coupled with a 1 k multichannel analyzer. The system is calibrated using Co-60, Cs-137 and Monazite. IAEA reference materials (RGU-1, RGTh-1 and RGK-1) were also used for checking the calibration of the system.

Results and Discussions: The activity concentration of granite samples were estimated [Figure 1]. 238U range varies from ≤5.05 (MDA) Bqkg-1 to 103.89 Bqkg-1 with geometric mean of 10.04 Bqkg-1, which is lower than the world average value of 50 Bqkg-1; 232Th varies from ≤5.04 (MDA) Bqkg-1 to 634.28 Bqkg-1 with geometric mean of 101.54 Bqkg-1 which is higher than the world average value of 50 Bqkg-1, 40K concentration varies from ≤17.3 (MDA) Bqkg-1 to 1243.59 Bqkg-1 with geometric mean of 222.3 Bqkg-1 which is lower than the world average value of 500 Bqkg-1 (UNSCEAR, 2008).[2] The radium equivalent activity (Raeq) values ranged from 21.63 to 1007.83 Bqkg-1 with geometric mean of 206.03 Bqkg-1. The radium equivalent activity is above the recommended limit of 370 Bqkg-1 for three samples. The estimated absorbed dose rate ranged from 10.45 to 437.3 nGyh-1. The calculated geometric mean of absorbed dose was 91.93 nGyh-1 which is higher than the population-weighted value 84 nGyh-1. The indoor annual effective dose values ranged from 0.05 to 2.15 mSvy-1, with geometric mean of 0.45 mSvy-1 and the outdoor annual effective dose values ranged from 0.01 to 0.54 mSvy-1 with geometric mean of 0.14 mSvy-1, which are within the annual effective dose equivalent limit of 1 mSvy-1.[3] The geometric mean of internal and external hazard index are 0.56 and 0.60 respectively. The external hazard index values are higher than unity for three samples and the internal hazard index values are higher than unity for four samples respectively.

Conclusion: Granite rock could contain naturally occurring radioactive elements such as uranium, thorium and potassium. Some granite contains more of these than others, depending on the chemical composition and the formation of the molten rock. In the present study granite samples investigated are within the recommended safe limit, except for some samples.
Figure 1: 238U, 232Th and 40K Activity concentrations of granite samples

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Keywords: Absorbed dose, effective dose, hazard indices, radium equivalent activity, specific radioactivity

  References Top

  1. Al-Zahrani JH. J Radiat Res Appl Sci 2017;10:241-5.
  2. UNSCEAR. United Nations & Scientific Committee on the Effects of Atomic Radiation, Report of the United Nations Scientific Committee on the Effects of Atomic Radiation: Fifty-Sixth Session. United Nations Publications; 2008.
  3. Cherry SR, et al. Physcics in Nuclear Medicine. 4th ed., Ch. 23. Radiation Safety and Health Physics; 2012.

  Abstract - 74187: The IRPA Norm Task Group – Recent and future activities Top

J. Hondros, R. Gellermann1

JRHC Enterprises, Adelaide, Australia, 1Nuclear-Control and Consulting GmbH, Braunschweig, Germany

E-mail: [email protected]

In 2019, IRPA formed a task group dedicated to radiation protection issues associated with naturally occurring radioactive materials (NORM) in industry. The task group is made up of recognized radiation protection practitioners from a number of countries and provides a unique practical perspective on management and regulation of NORM. The task group meets regularly and discusses existing and emerging NORM related issues. An objective of the task group as defined by the IRPA Executive in original terms of reference, was to develop best practice guidance. This has evolved and the primary focus of the task group is to develop a practitioner's handbook on good practices in NORM with the intention is to release this at the IRPA Congress in Orlando in 2024. Other activities of the task group include the regular participation in international conferences and forums. This paper will provide an overview and update on the work and activities of the task group including a description of an upcoming publication related to good practice in industries with NORM.

Keywords: Case studies, IRPA, NORM

  Abstract - 74280: Natural high background radiation areas of western coast of Kerala and Tamil Nadu – A brief report Top

K. Sreekumar, S. K. Jha, Jaison T. John, S. Ajeshkumar, R. Sujata, M. S. Kulkarni

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

E-mail: [email protected]

Indian Peninsular coast is known for its rich deposits of heavy minerals including Ilmenite, Rutile, Garnet, Zircon, Sillimanite and Monazite. Monazite is a radioactive mineral containing Thorium (8-9%, 320 Bq/g) and Uranium (0.3%,36Bq/g). Decay of thorium and uranium results in emission of alpha, beta particles and a number of gamma photons. Though monazite is present in very negligible proportion (PPM) throughout this region, higher concentrations of the order 0.5-5% are observed at coastal belts of south India. United Nations Scientific Committee on the Effects of Atomic Radiation (UNSCEAR) reports global average ionizing radiation exposure received from natural sources as 2.4mSv y-1. However many populations living in different parts of the globe receive radiation doses at higher levels. This report gives a brief account of ambient radiation levels, monazite content in sand/soil at selected locations of high monazite regions of Kerala and Tamilnadu and verification of UNSCEAR 2000 formula correlating the radiation field with the monazite content in sand/soil.

Materials and Methods: Radiation at 1m height in the HBRA around residential areas and in beaches were recorded using portable radiation surveymeter linked with Global Position System. Indoor survey was conducted whenever possible. The study involved measurement of ambient dose rate, collection of sand/soil samples from various locations and estimation of monazite content from 232Th levels in the soil using gamma spectrometric analysis. The radiation survey of environment was done using portable survey meters including EnV Rad Log survey meter, ECIL make GM survey meter, ICX Identifinder portable spectrometer and RDS -31 radiation survey meter. Higher radiation levels were observed at three regions, mainly Chavara – Neendakara region of Kollam, kerala, Karimanal Village of Thiruvananthapuram, Kerala and Muttom _Midalam region of Kanyakumari, Tamil nadu. Since most of the population under study has varying daily occupancy hours within residential premises (between 14 to 23 hour), their external dose received annually was estimated assuming indoor occupancy factor of 0.8 and outdoor occupancy factor of 0.2.

The correlation between monazite content in the soil and dose rate at 1m height was made. The average value obtained 1.45±0.014 μSv/h per Monazite percentage is in good agreement with theoretically predicted value. Hence UNSCEAR 2000 formula for calculating AIR DOSE from natural radionuclide is verified for thorium rich deposits.
Figure 1: EnV RaD LoG Survey meter with mobile readout

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Table 1: Estimated effective external dose at natural high background radiation areas

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Keywords: Dose rate, monazite, thorium, uranium

  References Top

  1. UNSCEAR. Sources and effects of ionizing radiation. Report to the General Assembly. New York: United Nations Scientific Committee on the Effects of Atomic Radiation; 2000.
  2. ICRP. Recommendations of the International Commission on Radiation Protection. Vol. 60. Tarrytown, NY: Elsevier Science; Publication; 1990.


  [Figure 1], [Figure 2], [Figure 3], [Figure 4], [Figure 5], [Figure 6], [Figure 7], [Figure 8], [Figure 9], [Figure 10], [Figure 11], [Figure 12], [Figure 13], [Figure 14], [Figure 15]

  [Table 1], [Table 2], [Table 3], [Table 4], [Table 5], [Table 6], [Table 7], [Table 8], [Table 9], [Table 10], [Table 11], [Table 12], [Table 13], [Table 14]


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