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Year : 2023  |  Volume : 46  |  Issue : 5  |  Page : 1-36  

Theme 1. Foundation Topics on Radiation Protection Philosophy and Risk Estimates

Date of Web Publication07-Feb-2023

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Source of Support: None, Conflict of Interest: None

DOI: 10.4103/0972-0464.368740

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How to cite this article:
. Theme 1. Foundation Topics on Radiation Protection Philosophy and Risk Estimates. Radiat Prot Environ 2023;46, Suppl S1:1-36

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. Theme 1. Foundation Topics on Radiation Protection Philosophy and Risk Estimates. Radiat Prot Environ [serial online] 2023 [cited 2023 Mar 23];46, Suppl S1:1-36. Available from: https://www.rpe.org.in/text.asp?2023/46/5/1/368740

  Abstract - 11147: Measurement of natural radioactivity in sediments of Thamirabarani river, Tamil Nadu, India using gamma ray spectrometry Top

V. Thangam, A. Chandrasekaran

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

E-mail: [email protected]

Introduction: Natural radioactivity is all over the place in the earth's environment and it exists in various geological formations such as rocks, soils, sediments, sand and as well as in plants, water and air. Sediment plays an important role in aquatic radioecology which are formed by the broken rocks or organic materials during the movement of water. Radioactivity contents are also present in sediments because there is no change in the chemical composition of rocks when it is broken by the water. It is one of the most common raw materials for construction works. These contain naturally occurring radionuclides such as 238U, 232Th and their daughter products and 40K.

Objectives: The main objectives of the present work are,

  • To determine the activity concentration of the radionuclides 238U, 232Th and 40K
  • To calculated all the associated radiological parameters.

These were compared with the world average values to assess the hazards to humans as per international standards.

Materials and Methods: 15 sediment samples were collected as per standard protocols (IAEA) from Thamirabarani river, Tirunelveli, Tamilnadu, India. The sampling area covers a total length of 80 km. The sampling locations are shown in [Figure 1]. All the collected samples were prepared as per the procedure and stored into 250 ml PVC containers (0.5 kg approx.). These packed samples were kept aside for 30-40 days to bring 222Rn and its short-lived daughter products into equilibrium with 226Ra. Each prepared sample was then placed on top of a 3''×3'' NaI(Tl) detector and the spectra was recorded for 20000 seconds. The gamma ray photo peaks corresponding to 1461 keV for 40K, 1764 keV for 214Bi and 2614 keV for 208Tl were used for determining the activity concentrations of 40K, 226Ra and 232Th respectively.
Figure 1: Sampling location map

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Results: In the present study, the levels of mean activity concentrations of the radionuclides 226Ra, 232Th, 40K and other radiological parameters such as radium equivalent activity (Raeq), absorbed dose rate (DR), annual effective dose equivalent (AEDE), external and internal hazard indices (Hex, Hin) and activity utilization index (AUI) for sediment samples from Thamirabarani river were determined [Figure 2]. These results were compared with the International recommended values given by UNSCEAR 2000. The results indicate that the mean activity concentration of 226Ra and 232Th are similar to the world average values whereas that of 40K is higher than the world average value. The major parameters which imply the effects of long exposure are well within the recommended limits. It can be concluded that the sediments of Thamirabarani river do not pose any radiological risk to humans and also when used as construction materials.
Figure 2: Activity concentrations of the radionuclides

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Keywords: Annual effective dose equivalent, dose rate, gamma ray spectrometry, radioactivity, river sediments

  References Top

  1. International Atomic Energy Agency. Soil Sampling for Environmental Contaminants. Vienna: International Atomic Energy Agency; 2004.
  2. United Nations Scientific Committee on the Effect of Atomic Radiation. Sources and Effects of Ionizing Radiation. UNSCEAR 2000 Report to General Assembly, with Scientific Annexes. New York: United Nations, UNSCEAR; 2005.

  Abstract - 11182: Risks associated with beryllium exposure, comparison with radiological risk philosophy and response models Top

Munish Kumar1,2, Alok Srivastava1

1Industrial Hygiene and Safety Section, Health, Safety and Environment Group, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Safety of individuals from harmful effects of carcinogen chemicals like Beryllium (Be) & ionizing radiations is always a matter of concern. The concepts & philosophy for chemicals like Be with which stochastic as well as deterministic effects are associated may not be exactly similar when compared to radiological protection.[1] Present paper provides an analogy between Be & radiation safety based approaches and related risk models. Like radiation, for Be at workplaces, the associated health effects are - i) Stochastic like cancer (mainly lung) & ii) Deterministic. The deterministic effects are acute beryllium disease (ABD), chronic beryllium disease (CBD) & skin related complications like dermatitis, cyan, skin reddening etc. Generally ABD occurs at higher Be concentrations (> 100 μg/m3) whereas CBD results from chronic exposures may be after long delayed periods after Be inhalation. While recommending levels for hazardous chemicals like Be, the concentration up to which no or lowest adverse effects appear are popularly called no or lowest observed adverse effect levels (NOAEL or LOAEL) & forms the basis for adoption of daily limit/level i.e. threshold limit value (TLV) or permissible level of exposure (PLE). Also temporary emergency exposure limits (TEEL) for hazardous chemicals along with emergency response planning guidelines (ERPG) can be helpful in quantification of deterministic effects in accidental situations. Further in radiological protection, the protection philosophy is to minimize the likelihood of stochastic effects like (cancer & genetic effects) to lowest acceptable level & to rule out the occurrence of deterministic effects like skin reddening, cataract etc. As per US-EPA the cancer (lung) risk for Be inhalation of 1.00 μg/m3 is 2.4 x 10-3 (US-EPA, 1998) which for recently revised PLE of 0.20 μg/m3 is 4.80 x 10-4 by application of linear non-threshold (LNT) model & is of the same order of magnitude (10-4) as adopted by ICRP for radiation induced lung cancer. Further for other Be related health hazards like ABD, skin complications etc., the threshold dose response model as defined for radiation induced effects is applicable. For CBD, the typical response model is non-linear with a threshold as shown in [Figure 1] (EC Report, 2018). It is important to note that the risk numbers mentioned in this paper are only for understanding and not to predict any cancer/ CBD cases as is recommended by International Commission on Radiological Protection (ICRP) as well as United Nations Scientific Committee on Effects of Atomic Radiation (UNSCEAR). From the discussion, it can be seen that as far as carcinogenicity from Be or radiation is concerned, the typical risk factors are comparable. However, protection levels and approaches are not exactly similar as the protection strategies for radiation induced effects and beryllium are different although both (radiation and Be) are known to exhibit stochastic as well as deterministic effects. Further in case of Be, the main concern is due to inhalation whereas in case of radiation, different exposure pathways become important depending upon the radionuclide.
Figure 1: Typical response (threshold) model for CBD (EC Report, 2018)

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Keywords: Beryllium toxicity, chronic beryllium disease, stochastic and deterministic effects and risk models

  References Top

  1. Lochard J. Proceedings IAEA International Conference Geneva. 2002. p. 143-52.
  2. U.S. Environmental Protection Agency, Report. Washington, D.C; 2002.
  3. European Commission – Final Report for Beryllium; 2018. p. 17.

  Abstract - 11209: Dose rate estimation of accidental discharge with 134Cs and 137Cs from NPP Top

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

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

E-mail: [email protected]

Dose rate measurements play an important role in post-assessment of nuclear accident. An air borne activity in a discharge plume used to estimate dose rate through which environmental impact and scale of accident severity can be decided. It is observed from past data that air borne activity discharged during nuclear accident is < 30% of total inventory present in reactor core IAEA series. 134Cs isotope formed as fission product (FP) as well as activation product (AP). But yield of 134Cs as FP is negligible. It is produced in reactor after neutron capture by 133Cs as AP. 137Cs is purely generated as FP Walker et al. The saturated activity ratio of 134Cs:137Cs for one year is evaluated for various reactors with different type of fuel [Figure 1], Bell et al. In this work, the total dose rate and fractional dose rate estimation Oza et al from Cs-137 FPs & Cs-134 APs of reactor plume is proposed. 134Cs:137Cs ratio estimated from plume, generated at time of event could be helpful during nuclear emergency situations to discriminate between nuclear reactor accident and nuclear weapon test. The dose rate estimated at grid point of Radiation Early Warning System (REWs) installed at nuclear site. The grid points are installed Area Gamma Monitors (AGM) used to monitored radioactivity discharge or movement in terms of dose rate (nGy/h) or count rate (cps). A special case of 100 MW thermal power capacity PHWR reactors is simulated for accidental release of prominent FP 137Cs and AP 134Cs. The activity concentration and dispersion in the discharge plume is function of metrological condition and stack parameters Shirvaikar. To cover all conditions, activity concentration estimation is carried out with Gaussian Plume Model (GPM) for extreme conditions (i.e. stable and unstable) and neutral condition. The receptor point is taken exactly above the AGM grid point in the downwind direction. The estimated activity concentration assumed to be uniformly distributed in homogeneous air medium NSRD report. In this work stack height is considered to be 100 m. The source term at release point and activity concentration and dilution factor at downwind distance of interest i.e. receptor point is tabulated in [Table 1].0. The dose rates are evaluated using deterministic code Subbaiah and Sarngapani[3] & simulation code. The assumptions made for evaluations are viz, cylindrical shape radioactive plume with homogenous air medium and uniformed radioactivity concentration. The diameter of cylinder deduced from horizontal бy(m) & vertical dispersion coefficient бz(m) of plume. The length of cylinder is taken 100m on either side of receptor point i.e. AGM location in REWS grid point. The axis of cylindrical passed through the point of released. The contributions from ground reflection and inversion aloft effect are negligible [Figure 2] Ref 3. The dose points and corresponding dose rates evaluated using Monte Carlo Simulation technique are tabulated in [Table 2].0 and [Table 3].0. The deterministic code showed 20% higher dose rate values as compared to simulation dose rate values.
Figure 1: 134Cs:137Cs Ratio in one Year for various type reactor and fuel

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Figure 2: Approximate cylindrical plume passed over AGM of REWs

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Table 1.0: Activity concentration and duration in 1.0 h

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Table 2.0: Dose rate in downwind direction x=250 m at different elevations

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Table 3.0: Dose rate in downwind direction x=1000 m at different elevations

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Keywords: Activity discharge, activity ratio, burn up, dose rate, gaussian plume

  References Top

  1. Nuclear Accident Knowledge Taxonomy. IAEA Series NG-T6.8; 2016.
  2. Shivalkar V. Guide Book for Nuclear Parameter in Nuclear and Ancillary Chemical Installations. BARC Report; 1973.
  3. Subbaiah KV, Sarngapani R. GUI2QAD_3D A graphical user interface for QAD-CGPIC programm. Ann Nucl Energy 2001;33:22-9.

  Abstract - 11242: Mortality analyses in the updated a mine cohort of uranium miners (1958–2018) Top

Wu Xiaoyan, Zhan Jingming, Xue Xiangming, Liu Zhanqi, Wang Jianming1, Zhang Yanna, Zhang Jingyun, Yang Kai, Yangxue, Ma Yuefeng, Gu Xiaona, Li Youchen

Department of Radiological and Environmental Medicine, China Institute of Radiation Protection, Taiyuan, 1HuaXiang Community Management Committee, Chenzhou, China

E-mail: [email protected]

Objectives: In order to analyze mortality risks in the extended follow-up of uranium miners from A mine and to examine the relationship between lung cancer to occupational exposure to radon.

Methods: The cohort included 5738 unranium miners employed in A Mine and followed up from 1958 to 2018. The essential information and dose data of uranium miners were recorded in personal files. The cause of death of miners were mainly confirmed by consulting the red headed notice of occupational disease treatment (especially lung cancer cases and silicosis cases), hospital death notice, medical certificate of hospital death, cremation certificate, death certificate of neighborhood committee, etc. in personal files. For a small number of cases with unknown cause of death, the community management committee confirms the cause of death to the relatives of the deceased. The standardized death ratio(SMR) was calculated by using the age specific disease specific death rate of male residents in small and medium-sized cities in China from 1987 to 2018. At the same time, Poisson regression analysis was used to calculate the odds ratio (OR) and ERR/WLM of lung cancer induced by radon.

Results: Over the period of follow-up, 213879 person-years at risk were accrued and 36.2% of the subjects had died. The average age of survivors, the exposure duration of the miner and the first exposed time were 65.96 ± 13.16 years, 14.05 ± 8.45 years and 24.44 ± 5.93 years respectively. The cumulative exposure of radon and its daughters was 19.74wlm (GM) and 69.68wlm (AA).Except for death from external causes (SMR=1.19,95%CI:1.03~1.36), lung cancer (SMR=1.20 95%CI:1.03~1.38) and nasopharyngeal cancer (SMR =1.80,95% CI:1.10~2.78), the mortality of other diseases were not higher than that of residents in small and medium-sized cities in China; the risk of lung cancer increases with the radon cumulative exposure (p<0.01); The relative risk coefficient (ERR/100WLM ) of radon induced lung cancer was 1.73 (95%CI 0.36-3.11), and the absolute risk coefficient(EOR/100WLM) of radon induced lung cancer was 1.215 (0.060-2.429) /100WLM.

Conclusion: After 60 years of operation in A mine, the mortality of respiratory cancers such as lung cancer and nasopharyngeal cancer in miners were still higher than that in small and medium-sized urban residents, which may be related to the deposition of radon and its daughters in the respiratory system after inhalation; The value of the ERR/100WLM and the EOR/WLM obtained in the study was within the range of that reported by ICRP115 about low exposure and low exposure rate. However, 79 workers still died of unknown causes in this survey, which may have a certain impact on the results. With the improvement of China's disease registration system and the increase in the number of miners surveyed in the later stage, the above problems will be improved to a certain extent. In short, through this study, it was believed that in the decades after the first exposure, lung cancer and nasopharyngeal cancer was still higher than that of residents in small and medium-sized cities. Long term updating of these cohorts will likely assess the continuing impact of past exposure and play an important role in the autonomous research of ERR/100WLM.

Keywords: Mortality, occupational epidemiological cohort, the cumulative exposure doses of radon, uranium miner

  Reference Top

  1. Tirmarche M, Harrison JD, Laurier D, Paquet F, Blanchardon E, Marsh JW, et al. ICRP Publication 115. Lung cancer risk from radon and progeny and statement on radon. Ann ICRP 2010;40:1-64.

  Abstract - 11250: Application of occupational health risk assessment in a radioactive waste treatment facility Top

Yang Xue, Zhan Jing Ming, Wu Xiao Yan, Xue Xiang Xing, Gu Xiao Na

China Institute for Radiation Protection, Taiyuan, China

E-mail: [email protected]

Occupational health risk assessment methods includes EPA model, ICMM model, mom model, COSHH model, etc. which are mainly used for the risk assessment of chemical poisons, and there are few reports on the risk assessment of ionizing radiation. In 2021, China researched Classification of occupational disease hazards risk for work sites in nuclear fuel reprocessing plant (EJ/T20267-2021). This classification method is applicable to the comprehensive risk classification study of ionizing radiation, chemical toxicant, noise, high temperature, etc. of workers in workplaces. The evaluation principle of the classification method is to comprehensively consider the factors such as the annual effective dose of individuals, the inherent hazards of chemical poisons, the exposure level of occupational hazards, protective measures against occupational hazards, the number and the physical labor intensity of workers, and give certain weights to each factor, and calculate the risk value (T value) according to the calculation formula. According to the size of T value, post risk could divided into four levels, low risk (T≤1), general risk (1<T This classificatio<T This classification method This study used the classification method to evaluate the occupational health risk of a new radioactive waste treatment facility, in order to understand the occupational health risk level of the personnel in the facility. The processes involved in the facility include waste liquid transportation, waste liquid combustion, tail gas treatment, incineration ash treatment, etc. The main posts were waste liquid transportation and waste liquid treatment. The occupational hazard factors exposed to waste liquid transportation were mainly ionizing radiation and noise. The occupational hazard factors exposed to waste liquid treatment posts were mainly ionizing radiation, noise, calcium oxide, nitrogen oxide, sodium hydroxide, carbon monoxide, sulfur dioxide and occupational exposure heat stress. The higher risk of waste liquid treatment post was higher risk, and the risk of waste liquid transmission post was at the general risk level of lower middle level, indicating that there were potential health risks in both posts. The reasons for the high risk of waste liquid treatment posts were more exposure to occupational disease hazards, higher personal dose estimation results, and higher inherent risks of poisons. For this post, the enterprise should strengthen the radiation shielding, sealing and other protective measures, management measures, personal protective measures to reduce the personal exposure dose of personnel, urge the post personnel to correctly equip themselves with ionizing radiation personal protective equipment for protection, and carry out occupational health examination and regular detection of occupational disease hazards in the workplace on time. Due to the single occupational hazard factors, low level and good protection, the waste liquid transportation post was a general risk operation post due to the large number of contacts. For this post, the current radiation protection measures and operation methods such as radiation shielding should be maintained, personal protective equipment should be worn correctly, workplace hazard detection, employee physical examination and occupational health training should be carried out regularly, and workplace radiation monitoring and other actions should be carried out regularly.
Table 1 : Occupational hazard status and risk results of radioactive waste treatment facilities

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Keywords: Ionizing radiation, occupational hazards, radioactive waste liquid, risk assessment

  References Top

  1. Risk Assessment Guidance for Superfund Volume I: Human Health Evaluation Manual (Part F, Supplemental Guidance for Inhalation Risk Assessment) EPA-540-R-070-002OSWER 9285. Office of Superfund Remediation and Technology Innovation Environmental Protection Agency; 2009. p. 7-82.
  2. Ministry of Manpower Occupational Safety and Health Division. A Semi-Quantitative Method to Assess Occupational Exposure to Harmful Chemicals [EB/OL]; 2022. Available from: http://www.Mom.gov.sg/workplace-safety-health/.

  Abstract - 11254: Experimental study on radio-adaptive response of immune system induced by ionizing radiation Top

Jixia Han, Li Dong1, Zhongwen Wang2, Xiangming Xue, Xiaona Gu, Lixia Su, Xue Yang, Ruifeng Dong, Lianfeng Zhao, Jingming Zhan

Division of Radiology and Environmental Medicine, China Institute for Radiation Protection, Shanxi University, 1Institutes of Biomedical Sciences, Shanxi University, Taiyuan, 2China Institute of Atomic Energy, Beijing, China

E-mail: [email protected]

Radio-adaptive response (RAR) is transient phenomena, where cells primed with a low dose exhibit reduced DNA damage with a high challenging dose. The immune system is one of the main target organ of ionizing radiation. RAR of immune system has always been a hot spot in the field of radiation immunology.

Purpose: The objective of this study was to find out if low dose γ-ray can induce RAR of the immune system in vivo.

Materials and Methods: Nine to fourteen weeks old C57BL/6N mice were irradiated with 0.075 Gy priming dose (PD), only 2.5 Gy challenge dose (CD) and both priming and challenging dose with an interval of 6 h (P+CD). Mice were dissected at 18 h and 1 week after CD and radiation induced immunosuppression in immune organs and peripheral blood immune cells was monitored as the endpoint for RAR.

Results: In pre-conditioning where the priming dose was 0.075Gy, RAR was induced in thymus and peripheral blood lymphocytes after 2.5 Gy and persisted until 1 week. Moreover, similar phenomena were also indicated in both CD4 T cells and CD8 T cells. However, the same treatment conditions in spleen and peripheral blood monocytes indicated absence of RAR at 24 h following PD. On the contrary, 24 hours after PD, the spleen index and the number of monocytes in peripheral blood of mice after CD decreased compared with the control group, and the decrease was more obvious than that after 2.5 Gy irradiation alone. In addition, 1 week after γ-ray irradiation, the peripheral blood monocytes of the mice in each group tended to increase.

Conclusion: These results suggest that low dose induced RAR of the adaptive immune system, but whether it can induce RAR of the innate immune system needs further study to demonstrate. This study may provide new data for the biological significance of immune enhancement effet and RAR induced by low-dose radiation, and is of great significance for accurately evaluating the impact of low-dose radiation on human health and its clinical application.

Keywords: Immune system, ionizing radiation, radio-adaptive response

  Abstract - 11255: Introduction of expert certification with the Australasian radiation protection accreditation board Top

B. Rogers, C. Jeffries1, K. Gregory2, R. Akber3

South Eastern Sydney Local Health District, Sydney, 1Department of Health and Wellbeing, SA Medical Imaging, 2SA Radiation Pty Ltd., Adelaide, 3Safe Radiation, Brisbane, Australia

E-mail: [email protected]

Introduction: Professional accreditation in radiation protection had its official origin in Australia at the 1990 Conference, which was held jointly between Australasian Radiation Protection Society (ARPS) and Australasian College of Physical Scientists and Engineers in Medicine (ACPSEM). Over the next decade a working group was empanelled to develop an accreditation scheme consisting of various practical examinations and a bank of questions so that a theoretical exam could be completed and graded. In 1997, ARPS and ACPSEM completed the “Candidates Kit” for a Radiation Protection Advisor level of certification, and were joined by the Australian Institute of Occupational Hygienists (AIOH) as a sponsoring organisation.[1] In 2000, the Australasian Radiation Protection Accreditation Board (ARPAB) was incorporated and began certifying Radiation Safety Advisors (CRSA). Since establishment, very little changed about ARPAB aside from the doubling from two to four representatives from each sponsoring society.

Methods: In 2018, during Kent Gregory's term as ARPAB Chair, a new working group was empanelled to create an exam bank for an expert level of certification, along with designing the Candidates Kit for being a Certified Radiation Safety Expert (CRSE). This advanced level of certification is designed to comply with the 2016 International Radiation Protection Association (IRPA) Guidance on Certification of a Radiation Protection Expert.[2] The Working Group with Kent as convenor further consisted of Riaz Akber, Cameron Jeffries and Brent Rogers, and as with the original certification, the expert certification has been internally authenticated.

Conclusion: At the 2021 ARPS Conference, March 2022, an advanced level of certification was introduced by ARPAB. The Working Group has now become the Panel of Examiners. The steps to achieving the Expert level of certification will be discussed in this presentation. Information about certification can be found on the ARPAB website.[3]

Keywords: Australasian Radiation Protection Accreditation Board, certification, International Radiation Protection Association expert, radiation safety advisor, radiation safety guidance on certification of a radiation safety expert

  References Top

  1. Available from: https://www.arpab.org.au/how-to-get-accredited/.
  2. Available from: https://irpa.net/docs/IRPA%20Guidance%20on%20Certification%20of%20a%20RP%20Expert%20(2016).pdf.
  3. Available from: https://www.arpab.org.au/.

  Abstract - 11291: Performance evaluation of electronic personal dosimeters Top

R. K. Dwivedi, P. C. Dhakar, A. K. Sudhir, K. Girish Kumar1, Kumar Manu

Health Physics Unit, NPCIL, Rawatbhata, Rajasthan, 1HP Group, NPCIL HQ, Mumbai, Maharashtra, India

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

Introduction: Electronic Personal Dosimeters (EPDs) of two different make models are in use at RAPS-3&4. The detectors used in these EPDs are Silicon Diode (semiconductor detectors-based devices). EPDs are being widely used as operational dosimeters in nuclear power plants because of some inherent advantages as compared with passive dosimeters like TLDs. This study details the performance checks of Electronic Personal Dosimeters (EPDs) carried out at Standard Dosimetry Calibration facility established at RRS-3&4. The conversion coefficients from air kerma have been determined for Dosimeters belonging to two manufactures used in RRS-3&4. Primary standard ionization chamber was used to measure the personal dose equivalent, Hp (10) as standard dosimetric quantity of interest. Parameters tested include the variation of response with radiation energy, angle of incident and dose rate dependence. Dosimeter irradiation and measurements were performed at the Standard Dosimetry Calibration facility of RRS-3&4. Standardized Radiation sources 137Cs and 60Co were used for performance test. The air kerma rates (measured with primary Ionization Chamber by RSSD, BARC) were 4 μSv/h to 0.16 mSv/h at the 75 cm distances respectively, The desirable dose rate at different distances are derived accordingly for performance test.. The relative uncertainties associated to the reference quantities Hp(10) for the sources used in the comparisons are ± 10% for 137Cs source; ± 5% for 60 Co source. The angular response to photon radiation were tested at an angle of 0, 60, 90,120 and 180 degrees for both make of EPDs with respect to 60Co and 137Cs sources. Angular response check was carried out to estimate the variation in dose and dose rate registered in the EPDs for different angles of incidence. This has helped in identifying the deviation of dose registered in TLD and EPD. Performance which is the ratio between estimated dose (Hi) and Delivered Dose (H) for the point of test for EPDs were estimated in addition to testing the Dose linearity response check. ANSI 2009 (Tolerance Level L ≤0.30.) and Trumpet Curve criteria were followed for performance analysis of dosimetry system as External Quality Assurance programme of EPD.

Conclusion: The response of electronic dosimeters was found within limits of acceptable performance. On the basis of angular response study the wearing arrangement of one make of EPDs was standardized to avoid TLD and EPD discrepancy.
Figure 1: Trumpet curve

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Figure 2: Angular response for 60Co source

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Figure 3: Angular response for 137Cs source

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Keywords: Dosimeter, gamma ray, Hp (10), KERMA, trumpet curve

  References Top

  1. IAEA Safety Series No.16. Calibration of Radiation Protection Monitoring Instruments. IAEA Safety Series No.16;
  2. Robers PL, Holbrook KL. Guideline for the Calibration of Personnel Dosimeters.

  Abstract - 11292: Effectiveness evaluation of lead Apron in external exposure control during EMCCR of RAPS-3 Top

Prakash Chandra Dhakar, Rohit Kumar Dwivedi, B. K. Salvi, A. K. Sudhir, Kumar Manu

Health Physics Unit, RRS#3 and 4, NPCIL, Rawatbhata, Rajasthan, India

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

Introduction: En-mass coolant channel replacement of Rajasthan Atomic Power Station is planned from the year 2022. The major activity during this large-scale refurbishment will be carried out at high dose rate area in Fueling Machine Vaults (FMV). In order keep external dose as low as reasonably achievable to radiation workers, Lead Apron as individual shielding is to be used for protection of personnel against external radiation. Hence it is essential to evaluate the effectiveness of the Apron in complex radiation field such as encountered in reactor environment. In addition, the effect of positioning of different dosimeters at different locations/Geometry needs to be studied which may cause unintended differences between doses recorded by different dosimeters. The protection provided by Apron of specific design has been evaluated for EMCCR of RAPS-3. This paper discusses the Protection Factor (P.F) observed for the Lead Apron.: Materials and Methods

The experimental set up was done with a standard 137Cs source of activity 21.5 MBq and 60Co source of radiation field 0.157mSv/h at 75cm (0.27mSv/h at 75 cm as on Oct 2012) was centrally mounted on a stand about 1m above ground level. A phantom made of (PMMA) and a wooden frame (for free air dose measurement) were placed at a distance of 0.5m in diametrically opposite directions at a parallel height with respect to the source. The set up was installed in the center of a calibration room so that scattering effects are uniform. For measurement purpose CaSO4: Dy TLDs were used. A set of three TLDs and EPDs were placed outside the lead apron enclosing the phantom in free air for exposure with 137Cs source at distance of 30 and 50 cm. At the same time another set of three TLDs and EPDs were placed inside the lead apron enclosing the phantom. The whole set of experiments were repeated with 60Co source for the measured distance of 60 and 70 cm. The value of the evaluated dose with respect to exposure on Phantom/Free in Air is given in [Table 1].

Result and Discussion: The calculated percentage reduction of intensity for lead aprons at station was in 22.8 % for 137Cs isotope and 9.7% for 60Co isotope. The experimental result shows that the reduction was about 14-27.3 %.

Conclusion: The average energy of major isotopes observed have been taken 700 keV (nearly to 137Cs gamma energy) as experienced and observed in pervious EMCCR campaigns have been taken for convenient theoretical calculations. The study revealed the effective protection factor of Lead apron was 14-27.3% for 137Cs and 60Co. Enhancement in protection factor is feasible only by increasing the thickness of lead in the apron. This will consequently increase the weight of the Apron and cause inconvenience to the workers and more time consumption for executing the job. Achieving ALARA exposure requires the optimization of both Time and Shielding.
Table 1: Comparison of theoretical and observed protection factor

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Figure 1: Experimental set up for determination of effectiveness of lead apron

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Keywords: ALARA, EMCCR, lead apron

  References Top

  1. Introduction to Health Physics. Herman Cember;
  2. Supplier's Specification from M/s Shah Corporation.

  Abstract - 11332: Derived limits for radon and progeny in underground uranium mines and Indian perspective Top

V. N. Jha1, S. K. Jha1,2, Rajesh Kumar1, R. L. Patnaik1, S. K. sahoo1, M. S. Kulkarni1,2

1Health Physics Division, BARC, 2HBNI, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: New Dose Conversion Factor with revised risk estimates and their implications on uranium mining industry is widely discussed issue for the facility owner, radiation protection personnel and the regulators. It has been emphasized that the ICRP 137 (2017) will be adopting the dosimetric approach that can be applicable under different conditions including uranium mining with site specific modifications. Despite major revision in dose coefficient for radon progeny the international consensus is yet awaited due to scientific, technical and societal implications for the uranium mining sector. Present papers summarize such guidelines and briefly mention the Indian perspective of applicable derived limits of 222Rn and progeny.

Discussion: Despite major revision in latest ICRP guidelines for estimation of 222Rn progeny dose for mine workers with new dose coefficient in place the industrial sector has highlighted some gap. Due to the difference in perceived risk between smokers and non-smokers categories of mine workers such gap seems prudent. ICRP 65 (1993) has recommended 4 WLM as Annual Limit averaged over five years with maximum of 10 WLM in a year. The Mines Safety and Health Administration (MSHA), US has set a maximum yearly radon exposure of 4 WLM for underground mining (30 CFR Part 57) with a maximum progeny concentration of 1 WL (3.7 kBq m-3 at 100% equilibrium). USNRC (10 CFR Part 30) and OSHA (29 CFR 1910.1090) for uranium processing facility also adopt a limit of 4 WLM and 1.1 kBq m-3 (assuming 100 % equilibrium). Further reduction below this level as Permissible Exposure Level (PEL) of 1 WLM is recommended (NIOSH, 1987). This PEL guideline assumes 100 % daughter equilibrium and worked out radon concentration of 8.3 pCi l-1 or 0.3 kBq m-3(NAP 2012). ACGIH (2011) recommended a TLV and BEI based annual exposure level of 4 WLM. Also an upper value for an individual worker's annual effective dose from radon of 10 mSv, which is related to the workplace action level for 222Rn of 1.5 kBq m−3 as specified by the ICRP 65, assuming the F factor of 0.4 and a DCF of 5 mSv per WLM. Derived Limits for Occupational Exposure provided by OSHA (1996) recommended the PEL of 3.7 kBq m-3EERn (equivalent to 1WL) for 40 h working week of 7 consecutive days and 1.11 kBq m-3EERn (30 pCi/l) averaged over a year. In ICRP publications emphasis has been given on the optimization process accounting the other radiation sources of the mines such as external radiation, ore dust inhalation etc. ICRP 126 has recommended a Reference Level of 10 mSv y-1 for workplace inaccessible to public, which can reflect an exposure of 2 WLM in existing system. Like other radionuclides, based on Biokinetics model data ICRP 137 (2017) has recommended the use of dose coefficient of 10 mSv WLM-1 without the adjustment for aerosol characteristics in uranium mines assuming the F factor as 0.2. Detailed review of recommendations on radon and progeny suggests that despite changes in the DCF there is broader consensus on Annual Exposure Limit of 4 WLM (averaged over five years) with 1 WLM = 5mSv. The Limit Derived for Indian Uranium mines as 0.3 WL (1 kBq m-3EERn) using the Annual Exposure Limit of 4 WLM presuming 2000 h working in a year and 170 h (standard) in a month is in line with the international recommendations. The Derived Limits applicable in Indian uranium thus can safely be used along with other restrictions already in place for ensuring the minimal exposure such as Annual Dose Limit of 20 mSv averaged for five years, maximum 30 mSv in a year and optimization criteria with [(External dose /Dose Limit) + (Radon progeny exposure in WLM / 4 WLM)] ≤ 1.

Conclusion: The Derived Limit for 222Rn and progeny of 1.0 kBq m-3 EERn applicable for Indian Uranium mines is in line with international recommendations. Revision in the light of new DCF can be considered only after a detailed scientific evaluation and justifiable additional benefits for Indian context.

Keywords: Dosimetry, internal dose, miners, optimization

  References Top

  1. ICRP. Protection Against Radon-222 at Home and at Work. ICRP Publication 65. Annals ICRP; 1993. p. 23.
  2. ICRP. Occupational Intakes of radionuclides: Part 3. ICRP Publication 137. Annals ICRP 2017. p. 46.
  3. NAP. Uranium Mining in Virgenia. 2012. p. 132-43.

  Abstract - 11341: Synthesis and TL characteristics of YAG:Gd,C Top

J. Misra, S. Bhattacharya1,2, L. Shaiju1, R. R. Bihari, V. Sathian, K. P. Muthe1, P. Chaudhury

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

E-mail: [email protected]

Yttrium Aluminium Garnet (YAG) when doped with: Gd, La, Ce etc. is a very interesting phosphor material Kottaisamy et al.[1] The dopant Ce3+ undergo 4f-5d transitions. These are well known phosphors as well as are promising material for high energy gamma dosimetry. Kulkarni et.al.,[2] have studied the Thermoluminesence (TL) and Optically Stimulated Luminescence (OSL) characteristics of YAG:C. Kulkarni et al.[2] had synthesised YAG:C using an electron gun under vacuum environment, and the TL & OSL response studies were carried out using 60Co gamma and 90Sr/90Y beta exposure. Similar efforts were made by Xin et al.,[3] using ball mill technique to synthesize the YAG:C. YAG:C powders as milled for 24h were subjected to TGT in reducing atmosphere at 1827°C to grow YAG:C crystals and studied the TL response for beta exposure. TL response of single crystal Al2O3 and YAG to neutrons from reactor environment were studied.[4] In this work we report the TL response of YAG:C synthesized by High Energy Ball Milling (HEBM) followed by sintering. The synthesis process was carried out using high purity Y2O3 (99.999%), Al2O3 (99.997%), Gd2O3 (99.999%), and graphite powder (>99.99%) as starting material. The molar ratio of Y2O3 and Al2O3 were 3:5. YAG powders have been synthesized using ball mill (agate balls and vials) for 4h followed by sintering at 1500°C in ambient condition. The synthesized material was characterized for phase purity using a bench top X-ray diffraction (XRD); [Figure 1]. XRD of prepared ball milled powder shows that phase formation takes place only after sintering at 1500°C. After phase purity was confirmed with the help of XRD, YAG; YAG:C1.0 and YAG:Gd1.0 were characterized for their TL response at 10, 50 and 100mGy of 137Cs gamma. TL response, for gamma dose of 50 mGy is shown in [Figure 2]. It can be seen from [Figure 2] that the response of Gd doped YAG is 1.8 times of undoped YAG. TL sensitivities of the phosphors have been compared with commercial available LiF standard TL material. The dosimetry grade YAG synthesis with doping concentration within ranges of 1-10000 ppm and other characterizations are under progress for neutron sensitivity studies. YAG:Gd shows single peak at 209.5°C for gamma. This indicates high potential of YAG:Gd for gamma dosimetry.
Figure 1: XRD pattern of YAG

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Figure 2: TL response of undoped and doped YAG

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Keywords: Gamma and neutron dosimetry, Thermoluminesence, YAG phosphors, YAG:Gd,C

  References Top

  1. Kottaisamy M, et al. Mater Res Bull 2008;43:1657-63.
  2. Kulkarni MS, et al. Rad Meas 2008;43:492-6.
  3. Xin BY, et al. J App Phys 2009;106:033105.
  4. Izerrouken M, et al. Nucl Instrum Methods Phys Res B 2014;326:90-4.

  Abstract - 11342: Synthesis of crack free xerogel basic network Top

J. Misra, V. Sathian, P. Chaudhury

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

E-mail: [email protected]

In 1958, Young achieved the tracks in insulating solids due to radiation interaction. Many experiments have been performed on various polymers due to the neutron interaction of hydrogen and carbon. Polymers are macromolecules of carbon and hydrogen bonds. Since then, many polycarbonates have been tested for neutron tracks [1] and CR-39 PADC (Polly-allyl-diglycol carbonate) is one of them. CR-39 shows good response between neutron energy 100keV to 10MeV. Its response is very promising between 1-2MeV neutron energy which corresponds to Nuclear reactors and fuel reprocessing facilities for personnel neutron dosimetry.[2] Since xerogel is a polymer material so it is attractive candidate for interaction study by neutron and gamma radiation. Synthesis of xerogels has been done using resorcinol and formaldehyde at ambient temperature.[3] In this work, we report the preparation of crack free resorcinol-formaldehyde (RF) network. The synthesis process was carried out using AR grade of Formaldehyde (37% in aq. solution, 36.5-38% stab. with 10-15% Methanol) and Resorcinol [C6H4(OH)2 , Purity: 99%]. 80g of Resorcinol is dissolved in 800 ml of (CH3)2CO and Formaldehyde of 160 ml is added to dissolve resorcinol solution. The mixture is subjected to continuously stir for 30 minute to synthesized homogeneous solution. 20 ml of 1.5M hydrochloric acid is added to the mixture to enhance the rate of gelation. HCl acts as a catalyst in this process. After half an hour, the solution is transferred into the mould. The solvent was evaporated by a porous cover at ambient conditions to get the three dimensional mould structure xerogel as shown in [Figure 1]. But cracks appears on the surface of formed xerogel is shown in [Figure 1]. Therefore modification in the above method is necessary to synthesize the crack free xerogel. SynthesIs of crack free xerogels has been carried out by the adjustment of temperature parameter for the above described method as shown in [Figure 2]. Now the xerogel surfaces are free from cracks as shown in [Figure 2]. The temperature of formation is found to be 80°C. The material is a light weight and kept inside the ordinary water for more than three years. It does not dissolve and decompose in the presence of water. Addition of graphite dopant in the xerogel network experiments have been carried out using chlorosulphonic acid. RF network accommodates graphite dopant very easily. Material characterizations of xerogels are under progress. The smooth crack free monolith of any size and shape could be prepared using appropriate mould and amount of the xerogel resin along with single and multiple dopants. This shows the potential applications of doped xerogel in the primary study of radiation interaction, damage and dosimetry.
Figure 1: Xerogel (RF-network

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Figure 2: Crack free Xerogel (RF-network)

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Keywords: Neutron and gamma interaction, neutron dosimetry, resorcinol-formaldehyde, xerogel

  References Top

  1. Sohrabi M, Morgan KZ. A new polycarbonate fast neutron personnel dosimeter. Am Ind Hyg Assoc J 1978;39:438-47.
  2. Pal RP, Jayalakshmi V, Sathian D, Chaurasiya G. IEEE Trans Nucl Sci 2009;56:3774-8.
  3. Pekala RW. J Mater Sci 1989;24:3221-7.

  Abstract - 11364: Effect of ionizing radiation on cellular stress, immunity and lifespan of common fruit fly Top

N. P. I. Das, V. Subramanian, B. Venkatraman

Aerosol Transport and Biodiversity Section, Radiological and Environmental Safety Division, Safety, Quality and Resource Management Group, IGCAR, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Low dose ionizing radiation (LDIR) has a wide range of biological effects including hormesis, adaptive response, and hypersensitivity that has an effect on lifespan. Based on certain field observations and few experimental studies LDIR is often considered to have beneficial effect on organisms. Nevertheless mechanism and effect of LDIR on various aspects of life is less understood. Towards this, as a beginning, a cumulative dose starting from 50-500 mGy is given to common fruit fly Drosophila melanogaster, which is an excellent non-human model to study effect of LDIR. In the next step, these doses will be apportioned as a multiple fractions and the effect will be studied. The present study intends to identify changes of lifespan and expression stress-sensitive genes and infection resistance in Drosophila melanogaster, exposed to said doses of γ-irradiation at various stages of development. Drosophila melanogaster (canton-S) strains, collected from National Drosophila stock, University of Mysore, was used for LDIR exposure. The control- and experimental flies were maintained at 25°C and a 12 hour light regime on a jaggery-yeast medium. In order to evaluate the effect of radiation, flies were exposed to 50 mGy, 100 mGy and 500 mGy. Flies were exposed at first instar larval stage and adult stage in duplicate. To check the immunity of flies against bacterial infection the flies were exposed to Pseudomonoas aeruginosa. Overnight culture of P. aeruginosa was prepared and was adjusted to OD600 = 0.008. The bacteria was administered to the exposed adult flies via fine needles. Their survival was observed for 7 days. Similarly locomotory behavior was observed by counting the number of flights made by the flies. To understand the antioxidant defense system expression Superoxide dismutase (SOD) activity was studied. After bacterial infection it was observed that flies exposed to 100 mGy and 500 mGy survived 6 days and 7 days respectively. On the other hand control and 50 mGy exposed flies survived for a lesser period of 4 days. This implies that an exposure to LDIR may increase the innate immunity of the flies. It was observed that flies exposed to 500 mGy survived more than 40 days where as control flies survived for 38 days. The 50 mGy and 100 mGy flies survived 40 and 41 days respectively. In general it was observed that the lifespan of flies increased marginally after exposure to low level radiation. The locomotory behavior of the flies were compared after exposure. The number of flights per minutes was counted for the flies in control and exposed tubes. The Flies exposed to 500 mGy showed locomotory behavior reduced 3 fold in compared to control. Superoxide dismutase (SOD) activity was studies using cayman SOD assay kit in Teccan infinite 200 pro microplate reader. Results were expressed in Unit/mg protein. Following irradiation SOD showed changes ranging from 1.86, 2.12 and 1.62 folds in larval stages. Similar trend of expression was also observed in adult flies. This suggest that LDIR may responsible for higher expression of SOD. The above findings shows the probable positive effect of LDIR on cellular stress management and immunity and lifespan of Common fruit fly Drosophila.
Figure 1: SOD activity in exposed Drosophila fruit flies larvae and adults

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Keywords: Hormesis, immunity, low dose ionizing radiation

  References Top

  1. Moskalev A, Shaposhnikov M, Turysheva E. Life span alteration after irradiation in Drosophila melanogaster strains with mutations of Hsf and Hsps. Biogerontology 2009;10:3-11.
  2. Parkes TL, Elia AJ, Dickinson D, Hilliker AJ, Phillips JP, Boulianne GL. Extension of Drosophila lifespan by overexpression of human SOD1 in motorneurons. Nat Genet 1998;19:171-4.

  Abstract - 11425: Integrating buoyant gas diffusion scheme and chemical conversion scheme in FLEXPART-WRF for chemical gas dispersion and chemical emergency impact assessment Top

Shanu Karmakar1, P. T. Rakesh1, C. V.Srinivas1,2, S. Chandrasekaran1, S. S. Raja Shekhar3, B. Venkatraman1,2

1Indira Gandhi Centre for Atomic Research, 2HomiBhaba National Institute, IGCAR, Kalpakkam, Tamil Nadu, 3National Remote Sensing Centre, ISRO, Hyderabad, Telangana, India

E-mail: [email protected]

In this work the effects of buoyancy and chemical conversion in atmospheric transport and impact assessment of chemical/toxic gas (e.g. NH3, Cl, LPG etc.) releases during chemical industrial emergencies are studied by incorporating buoyant gas diffusion and pseudo first-order chemical conversion scheme in the Lagrangian particle dispersion model FLEXPART-WRF.[1] Buoyancy scheme is incorporated in the z component of trajectory equation in FLEXPART-WRF[2],[3] as,

B is density difference, τ is Lagrangian time scale, N = Brunt-Väisälä frequency. dW is Wiener process with mean zero and variance dt. Air entrainment into the gas is modeled using entrainment coefficient

Where and are the mean position and standard deviation of ith particle in k, l, m cell at time t. For the dispersion becomes passive. The pseudo first-order chemical conversion is implemented based on the equation,; For anhydrous Ammonia to Ammonium Hydroxide μ = 7.5 10-5s-1. Simulation is run for a fictitious 1341 tons of liquid Ammonia release from Factories & Boilers (FAB), Kerala site with 1 km WRF Domain-3 resolution [Figure 1] and 250 m FLEXPART-WRF dispersion grid for 1100 IST 2nd Nov, 2016 and 0200 IST 3rd Nov, 2016 to test Boundary Layer (BL) effects on dispersion. WRF simulation shows local land-sea breeze circulation, deep convective BL (700 m-800 m) and shallow stable BL (150 m-200 m).

The results [Figure 3] indicate that

  • Daytime simulated ground level concentration (GLC) of NH3 decreases significantly due to strong buoyancy effects within convective BL.
  • Nighttime GLC of NH3 reduces significantly due to chemical conversion in shallow stratified BL.
  • Overall nighttime GLC is higher than daytime due to near-ground stratification of stable BL.

Centerline GLC falls to (1 hr Acute Exposure Guideline Level, www.epa.gov) AEGL1 (30-159 ppm) from AEGL3 (≥1100 ppm) at 0.5 km and 2 km for daytime and nighttime cases respectively [Figure 4]. Both schemes are successfully incorporated in FLEXPART-WRF model and valid for any buoyant chemical species for impact assessment study.

Authors thank Director, IGCAR for support.
Figure 1: Study domain

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Figure 2: Simulated wind flow and boundary layer height at Kerala

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Figure 3: GLC (ppm) at (a) 1100 IST 2nd and (b) 0200 IST November 3, 2016

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Figure 4: Downwind centreline GLC (ppm)

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Keywords: Boundary layer, buoyancy, chemical conversion, chemical emergency, entrainment, stability

  References Top

  1. Brioude J, et al. Geosci Model Dev 2013;6:1889-904.
  2. Van Dop H. Atmos Environ 1992;26:1335-46.
  3. Gopalakrishnan SG, Sharan M. Atmos Environ 1997;31:3369-82.

  Abstract - 11427: Development of a software APDCAL for annual public dose assessment from routine releases of radionuclides during normal operation of nuclear facilities Top

Shanu Karmakar, P. T. Rakesh, R. Deepu, C. V. Srinivas, S. Chandrasekaran, B. Venkatraman

Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

In this work a fortran and php based software named Annual Public Dose Calculation or APDCAL is developed to calculate the annual effective dose in the public domain around a nuclear facility. The software is developed based on the methodology prescribed by the regulatory authority which includes the calculation of public doses from all the relevant exposure pathways from all the relevant radionuclides. The doses are calculated both for infants and adults through all pathways such as inhalation, immersion, cloud exposure, ingestion and ground deposition.

In order to compute the dose due to various pathways first the ground level concentrations are computed using sector averaged Gaussian plume model for 16 wind sectors. The user has to provide site specific meteorological data containing variables such as wind speed, wind direction and stability for any number of years of data for statistically meaningful dilution factor information. With the above data and user inputs of site name, release height, site boundary distance, gamma energy etc the software creates the namelist.dat file. The dilution factor is then used for the computation of inhalation, immersion, ground deposition and ingestion doses due to food crop, milk and meat for adults and infants. The dose conversion coefficients for 120 relevant radionuclides, dietary habit and other parameters for all exposure pathways and adult/infant age groups are provided as look up table for the software. To compute the total dose, plume shine dose is also computed by triple integration of the Gaussian plume equation. Finally, total dose is computed both by immersion and plume shine method and maximum among the two is considered as the total dose and sector on which it is computed is the effected sector. The methodology is based on the revised ECPDA document. A few stages in the software are shown in subsequent figures.

The software computes the wind rose and triple joint frequency distribution after taking user input [Figure 2] the user need to input the radionuclides of interest and release quantities either manually or by uploading the data in the prescribed format. After uploading the files, the software runs in stages as shown in [Figure 3].

After the successful execution of each module an information.txt file is generated containing all doses through all pathways, sector of maximum impact, TJFD etc. The Software calculations are tested with Hukkoo et al.[1] manual estimates for a fictitious 1 TBq/d release from PHWR-MAPS (Madras Atomic Power Station) at Kalpakkam for 2015-2019 meteorological observation. Except for the cloud shine dose the software outputs are found to vary <1% compared to the manual estimates. The cloud shine dose is ~2.5% lesser compared to the manual estimate due to scheme of triple integration used for cloud shine dose in the software. Overall the software is efficient tool for annual public dose assessment compared to manual calculation.
Figure 1: Start page of APDCAL

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Figure 2: Page for user to provide NPP information

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Figure 3: Stages in dose computation

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Authors thank Director, IGCAR for providing support.

Keywords: Cloud shine dose, dilution factor, immersion dose, public dose assessment, software

  References Top

  1. Hukkoo RK, Bapat VN, Shirvaikar VV. BARC Report 1412; 1988.
  2. IAEA SRS-19. International Atomic Energy Agency, Vienna; 2001.
  3. AERB ECPDA Revised Report-2; 2020.

  Abstract - 11464: Comparative study of natural radioactivity in marble slurry and cement  Top

D. Meena, S. K. Gupta, S. Degra, N. K. Bawalia, V. P. Yadav

Department of Physics, University of Rajasthan, Jaipur, Rajasthan, India

E-mail: [email protected]

All living species have been constantly exposed to an incessant radiation flux on the surface of the earth. An average of 90% of the total radiation dose of a person comes from both primordial radionuclides and cosmogenic origin natural radionuclides. Out of which 80% is contributed by primordial radionuclides. There were some radionuclides present at the time of formation of the earth and hence are also present in earth crust which is called primordial nuclides. The main source of natural radioactivity is primordial radionuclides and the most common primordial radionuclides are 40K, 232Th and 238U series radionuclides.[1]

Slurry and Cement samples were collected and prepared by following standard method by Sharma et al.[2] for measurement of natural radioactivity. Gamma-ray spectroscopy is used to estimate the natural radioactivity of (250 gm) slurry and cement samples. It is done by using a Canberra Coaxial HPGe semiconductor detector based setup detail of detection device given Meena et al.[3]
Table 1 : Calculated activity of slurry and cement

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Table 2: Calculated different radioactivity parameters

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Details of detector calibration of energy and efficiency are given in the previous study by Meena et al.[4] which was done using five (137Cs, 133Ba, 60Co, 57Co and 22Na) standard gamma ray sources. Absolute efficiency calibration of HPGe detector is shown in [Figure 1].
Figure 1: Absolute efficiency calibration of HPGe detector

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Detector was shielded with 10 cm lead and 2 mm inner layer of copper sheet to minimize the background and X-ray from lead respectively as mentioned Al-Sewaidan.[5]

The activity concentration of slurry and cement samples was measured using the following formula Sharma et al.,[2] Nemlioglu et al.[6]

Where; N, η, Iɼ, t and m are net counts, measured photo peak efficiency, gamma-ray intensity, data acquisition time, and weight of the sample respectively.

In the present study, the activity concentration of radionuclides 226Ra, 232Th and 40K are obtained (6.43 ± 0.68, 5.09 ± 0.72 and 528.78±21.78 Bq/kg ) and (85.83±2.56, 123.50±3.63 and 558.75±22.45 Bq/kg ) for Slurry and Cement respectively. The Cement sample has very high activity concentrations for all radionuclides compare to Slurry. Similarly, other hazard indices are also obtained high for Cement shown in [Table 2].

Keywords: Absorbed dose, activity concentration, gamma spectroscopy, HPGe detector, natural radioactivity

  References Top

  1. UNSCEAR. Report to the General Assembly with Annexes. United Nations, New York; 2000.
  2. Sharma BA, et al. Radiat Prot Environ 2017;40.
  3. Meena D, et al. Adv Sci Eng Med 2019;11.
  4. Meena D, et al. Int J Radiat Res 2019;17.
  5. Al-Sewaidan HA. J King Saud Univ Sci 2019;31.
  6. Nemlioglu S, Sezgin N, Ozdogan Cumali B. Adv Toxi Const B Mat 2022.

  Abstract - 11601: A profile study on electromagnetic fields radiation from 5G radio base station in Kuala Lumpur Top

R. Tukimin1,2, S. Z. Sapuan2, S. W. Yunoh1, N. A. Zainal1

1Radiation Health and Safety Division, Malaysian Nuclear Agency, Bangi, Selangor, 2Faculty of Electrical and Electronic Engineering, Universiti Tun Hussein Onn Malaysia, Johor, Malaysia

E-mail: [email protected]

In recent years, the numbers of Electromagnetic Fields (EMF) sources have increased, especially in urban areas due to proliferation of the radiocommunication infrastructure base station for wireless mobile communications. This is due to the development of wireless communication technology and its exponential growth and the advent of 5G technology. This scenario has triggered public fear especially on the long-term health effects for those living or working in the vicinity of 5G base station. Malaysian government and telecommunication industry are improving the infrastructure and quality of communication technology services to ensure the optimum coverage and better connectivity, while the public remain concerned on the continuous radiofrequency (RF-EMF) emission exposure. Public questioned on being overexposed and living adjacent to the base station. Due to the concern, a profile study of RF-EMF radiation level was carried out around the 5G base station in Kuala Lumpur when the 5G network were started to be deployed. The objectives of this study is to assess and measure the 5G RF EMF radiation exposure level in various location in Kuala Lumpur and compare with the permissible exposure limit (PEL) to human as stated in Malaysian Communication and Multimedia Commission (MCMC) Mandatory Standard and International Commission for Non-Ionising Radiation Protection (ICNIRP) guidelines.[1],[3] The measurement was conducted accordance to Malaysia Technical Standard Forum Berhad (MTFSB).[2] Frequency Selective Radiation Monitor (SRM) has been used as a Spectrum Analyzer together with an isotropic electric field probe (420MHz – 6GHz) and 5G decoding module. Then, an extrapolation factor for downlink signal has been used to evaluate the maximum emission. Study at five locations around Kuala Lumpur has been conducted at 5G frequency (FR1) of 700 MHz and 3.45 GHz. The frequency central (FC) provided by Network Service Provider has been obtained to decode the 5G signal.

[Figure 1] shows the FC signal scanning from the SRM system. The Emax will be calculated as equation 1 based on the FC data and extrapolation factor obtained from the NSP.
Figure 1: 5G EMF signal captured between 3.4 and 3.419 GHz

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Emax = Emeas + 10 log (K) (1)

Where Emeas is the maximum electric field from the measurement at FC and K is the extrapolation factor. The results (Emax) of RF-EMF emits by 5G base station are comply to the permissible exposure limits for public. The RF-EMF level is only 3.51% of the PEL stipulated by MCMC and ICNIRP guidelines.[1],[2],[3] The data obtained from the study is very important for current and future reference when 5G network is fully deployed for optimum coverage in Kuala Lumpur, Malaysia.

Keywords: 5G, EMF radiation, permissible exposure

  References Top

  1. MCMC. MCMC Mandatory Standards for Electromagnetic Field Emission from Radiocommunications Infrastructure, Determination No. 5 of; 2021.
  2. MCMC MTFSB TC G032. Prediction and Measurement on EMF Exposure from Base Station; 2021.
  3. ICNIRP Guideline. Guidelines for limiting exposure to electromagnetic fields (100kHz to 300 GHz). 2020;118:483-524.
  4. ITU-T K.61. Guidance on Measurement and Numerical Prediction of Electromagnetic Fields for Compliance with Human Exposure Limits for Telecommunication Installations; January, 2018.

  Abstract - 11628: Effects of neutron and γ-ray combined irradiation on the transcriptional profile of human peripheral blood Top

Y. Y. Yuan, D. J. Chai, R. F. Zhang, J. C. Dong, Z. X. Zhang, X. H. Dang

Department of Radiation and Environmental Medicine, China Institute for Radiation Protection, Taiyuan, China

E-mail: [email protected]

Neutron is a kind of high LET ray. Compared with low-LET rays, the radiation damage caused by neutrons is more complicated and difficult to repair, which will cause more serious health effects.[1] In actual scenarios, neutron radiation such as clinical radiotherapy, nuclear power plant leakage, space flight, high-altitude flight, and atomic bomb explosion is often accompanied by gamma rays.[2] Because the research on the mixing effect of neutrons and gamma rays is restricted by irradiation conditions and other factors, the mixing effect of neutrons and gamma rays is still unclear. The study of biological effects is the key to solve a series of radiation protection problems such as radiation health impact assessment. Therefore, it is necessary to study the mixing effect of neutron and gamma rays. We studied the effects of γ, neutron and neutron+γ-ray combined irradiation on the transcription spectrum in human peripheral blood of three healthy adult men. Samples were irradiated with 1.42 Gy 2.5-MeV neutron irradiation, 0.71 Gy neutron+0.71 Gy 137Cs γ irradiation, and 1.42 Gy 137Cs γ irradiation. Transcriptome sequencing identified 303, 244, and 197 differentially expressed genes, of which 56 were shared [Figure 1]a. GO analysis of common differentially expressed genes showed that the three groups were co-enriched for 289 GO terms, of which 119 were significantly different (p < 0.05), including 48 second-order GO terms, which were related to biological adhesion, biological regulation, immune system processes, metabolic processes, positive and negative regulation of biological processes, response to stimulation, protein activity, etc. [Figure 1]b. Compared with control, there were 108 pathways in common among three groups of which 14 were significant (p < 0.05). KEGG enrichment results showed that the signal pathways with the largest number of activated genes were mainly cytokine receptor interaction pathway and p53 signaling pathway. The 56 genes were analyzed by protein-protein interaction networks [Figure 1]c, and the core genes (AEN, DDB2, PCNA, FDXR, MDM2, POLN and BAX) were verified by fluorescence quantitative polymerase chain reaction (qPCR) in human peripheral blood [Figure 1]d. Additionally, irradiation of AHH-1 human lymphocytes with a 252Cf neutron source at 0, 0.14, 0.35, and 0.71 Gy, fluorescence qPCR revealed a dose-response relationship for BAX, DDB2, and FDXR at dose ranges of 0–0.71 Gy [Figure 1]e, with R2 of 0.803, 0.999, and 0.999 [Table 1], respectively. Thus, three irradiation methods can induce changes in the genomic transcription spectrum of human peripheral blood lymphocytes. BAX, DDB2, and FDXR are expected to be molecular targets of neutron injury.
Figure 1: Experimental data. (A) Co-expressed gene (B) GO enrichment map of differentially expressed genes (C) protein-protein interaction networks (D) Verification results of core genes (E) Dose–response relationship of BAX, DDB2, and FDXR

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Table 1: BAX, DDB2 and FDXR gene expression fitting equations

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Keywords: 252Cf, combined irradiation, neutron, transcriptome, γ-ray

  References Top

  1. Mukherjee S, Grilj V, Broustas CG, et al. Radiat Res 2019;192:189-99.
  2. Dietze G, Bartlett DT, Cool DA, et al. Assessment of radiation exposure of astronauts in space, ICRP Publication 123. 2013.

  Abstract - 11630: Potential of FLASH-RT for clinical treatment of glioma Top

Lin Li, Yayi Yuan, Yahui Zuo

China Institute for Radiation Protection, Taiyuan, China

E-mail: [email protected]

Recent preclinical evidence suggests that ionizing radiation at ultra-high dose rates, also known as FLASH radiation therapy (FLASH-RT), can selectively reduce radiation damage to normal tissues while remaining as effective as conventional radiation therapy (CONV-RT) in killing tumours. In this review article, we explore the advantages of FLASH-RT in protecting normal tissues and its applicability in the treatment of gliomas. We also briefly discuss the challenges encountered with the use of FLASH-RT. Radiation therapy mainly involves irradiating tumor tissue with radiation to induce DNA damage and kill the tumor. However, although image guidance is used to precisely plan the irradiation target area during irradiation, the damage to normal tissues caused by radiation cannot be eliminated. As an unconventional radiotherapy technique, FLASH radiotherapy (FLASH-RT) has a single dose rate of ≥40Gy/s, which is much higher than the dose rate of conventional dose-rate radiotherapy (CONV-RT) currently used in clinical practice (≤0.1 Gy/s). The current study found that although there was no significant difference in long-term tumor control between FLASH-RT and CONV-RT, FLASH-RT demonstrated a surprising ability to protect normal tissue. So far, clinical trials of FLASH-RT have been conducted in cutaneous lymphoma and tumor bone metastases, both with more favorable results. Glioma is the most common primary intracranial malignant tumor. Currently, the treatment for glioma is mainly surgical resection, with adjuvant radiotherapy and chemotherapy 2-6 weeks after surgery. Although the above treatment methods have efficacy, the post-treatment complications should not be ignored. Among them, the early complications after radiation therapy mainly include vasodilation, blood-brain barrier damage, and edema leading to intracranial hypertension. Late complications include white matter encephalopathy, radiation brain necrosis, and cerebrovascular lesions, significantly reducing the quality of patients' survival after treatment. In contrast, FLASH-RT is an ideal solution for glioma. Preclinical studies in mice and rats showed no significant difference in long-term tumor control between FLASH-RT and CONV-RT for gliomas. However, FLASH-RT demonstrated outstanding CNS protection, including protection of cognitive and memory functions, protection of brain vasculature, blood-brain barrier, and preservation of information processing, learning, memory, emotion, and social skills in mice, significantly improving the quality of patient survival. However, more preclinical studies are needed for FLASH-RT as a new treatment. (1) The physiological mechanism of the effect of FLASH-RT on glioma is still unclear. The current research on FLASH-RT for glioma mainly focuses on tumor control effects. However, the mechanism of FLASH-RT in glioma needs an in-depth study. (2) The role of physical parameters in the FLASH effect still needs to be clarified. Initially, an average dose rate of ≥40 Gy/s was thought to be responsible for triggering the protective effect of FLASH. However, it was found that other physical parameters, including radiation sources (e.g., electrons, protons, heavy ions, and photons), single-pulse dose, pulse frequency, and total exposure time, can also affect the FLASH effect. Therefore, it is still unclear how the various physical parameters affect the FLASH effect. (3) The effect of split irradiation on the tumor-controlling effect of FLASH-RT and the protective effect of normal tissues remains unclear. Although animal experiments have shown that the tumor tissue control effect of FLASH-RT is similar to that of CONV-RT, most of the current experiments use a single irradiation. (4) To determine the area of tumor tissue irradiation prior to treatment versus patient positioning during treatment, more precise imaging equipment is needed to deliver high doses of radiation to the head in a short time. Some issues remain to be resolved in translating FLASH-RT from laboratory studies to clinical applications. However, the advantages of FLASH-RT in reducing normal tissue damage have attracted the interest of experts and scholars. As a result, it is believed that the academic community will soon witness a research boom in FLASH-RT. This work was supported by Shanxi Provincial Youth Research Fund (No. 20210302124283).

Keywords: conventional radiotherapy, FLASH radiotherapy, glioma, nervous system

  References Top

  1. Bourhis J, Sozzi WJ, Jorge PG, Gaide O, Bailat C, Duclos F, et al. Treatment of a first patient with FLASH-radiotherapy. Radiother Oncol 2019;139:18-22.
  2. Mascia AE, Daugherty EC, Zhang Y, Lee E, Xiao Z, Sertorio M, et al. Proton FLASH radiotherapy for the treatment of symptomatic bone metastases: The FAST-01 nonrandomized Trial. JAMA Oncol 2022;e225843.
  3. Liljedahl E, Konradsson E, Gustafsson E, Jonsson KF, Olofsson JK, Ceberg C, et al. Long-term anti-tumor effects following both conventional radiotherapy and FLASH in fully immunocompetent animals with glioblastoma. Sci Rep 2022;12:12285.

  Abstract - 12308: Reductive transformation of U(VI) to U(IV) in soil system Top

Sabyasachi Rout, S. S. Wagh, Sonali Yadav, Vandana Pulhani, A. V. Kumar

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

E-mail: [email protected]

Uranium is a radioactive metal that is prevalent in the environment as a result of natural and human-induced activities. The largest sources of subsurface U contamination are mining for nuclear power generation and weapons manufacturing, despite the fact that modest amounts of U are found in all crustal material. As the oxidation state of U changes, its mobility in the environment changes as well. In light of the fact that the sorption of U(VI) is highly dependent on environmental factors, extensive studies have been carried out to immobilize U(VI) by reducing it to U(IV) in uraninite (UO2(s)), which is significantly less soluble than U(VI) species. A study was conducted in natural soil system to investigate the redox transformation of U(VI) with ageing. For the study, soil (inside a Lysimetric tank) was amended with 20 L of 200 mg/L U(VI). Redox transformation of the U has been monitored for a period of three years. For a better understanding, a detailed characterization of soil was carried out before and after the amendment of soil (spiking) in terms of pH, particle size, Fe, Mn, organic matter, CEC, mineralogy, and total U content. Speciation study was carried out using X-ray absorption near edge spectra (XANES) and the carbonate leaching technique. Investigation of soil properties shows that, there is no significant change in soil properties over the study period except for U, which exhibits a decreasing trend till the end of the 2nd year. The small variations in estimated values of parameters in different samples may be attributed to measurement uncertainty. U was not detected in the background sample. The XANES spectra were recorded for all the post-amended soil samples. Unfortunately, spectra were recorded for the 1st and 6th month samples only [Figure 1]. In other samples, due to low U content, the spectra were not interpretable. Comparison of samples with standard data of U(IV) and U(VI) revealed that, U exists as U(VI) in both the samples with a minute deviation from U(VI) standard XANES spectra at energy 17187eV, which corresponds to distortion from octahedral geometry and can be viewed as binding of U(VI) to the soil sorbent. The carbonate leaching data presented as bar plot in [Figure 2] shows that, U remained as U(VI) till the end of '1st year of contamination, then gradually transformed to U(IV) and at the end of the 3rd year about 95% of total U transformed to U(IV). This transformation may be attributed to multiple factors: biogeochemical induced oxidation of Fe(II) to Fe(III) followed by reduction of U(VI) to U(IV),[1] occlusion of U-Fe (oxides/hydroxide) complex on amorphous silica[2] etc. Furthermore, in soil system, various biotic and abiotic reducing agents are available; those are capable of reducing U(VI) to U(IV). The presence of so many potential reducing agents in soil system makes it difficult to deduce the nature of redox transition pathways. Studies that link molecular-scale U speciation to pore-scale mineralogy and macroscale behavior over defined time intervals are needed to better understand redox transition mechanisms and biogeochemical processes. Molecular scale investigation is limited since the U level in aged soil is below the detection level of XAS.
Figure 1: XANES spectra of Soil samples with standard U(IV) and U(VI) XANES

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Figure 2: % of U(IV) and U(VI) distribution in soil samples generated by carbonate leaching

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Keywords: Reduction, soil, uranium, X-ray absorption near edge spectra

  References Top

  1. Fredrickson JK, Zachara JM, Kennedy DW, Duff MC, Gorby YA, Shu-mei WL, et al. Geochim Cosmochim Acta 2000;64:3085-98.
  2. Rout S, Kumar A, Ravi PM, Tripathi RM. J Hazard Mater 216;317:457-65.

  Abstract - 12309: Radiological assessment of burrowing mammals (rat) living in uranium mill tailings pond, Jaduguda Top

N. K. Sethy1,2, S. K. Jha2,3, V. N. Jha1,2, S. Singh1, G. P. Verma2, S. K. Sahoo2, M. S. Kulkarni2,3

1Health Physics Division, BARC, Jaduguda, Jharkhand, 2Health Physics Division, BARC, 3Homi Bhabha National Institute, Anushaktinagar, Mumbai, Maharashtra, India

E-mail: [email protected]

The use of absorbed dose in biota as a tool is a recent development in the discourse of environmental radiation protection. ICRP[1] have recommended the separate protection of biotic and abiotic component of the environment in the different radiological situations. Along with general public, the non-human biota which includes plants and animals needs adequate protection from the exposure to ionizing radiation arising from industrial activities. Uranium industry during processing of the ore discharges the radioactive waste at a designated site called tailings pond. Non-operational tailings pond with complex ecosystem gradually becomes habitats supplanting rich biodiversity. Rats surviving in the tailings pond area exposed to U series radionuclide & receive dose. Dose rate to the rats living in the uranium tailings pond is a measure of radiological risk to inhabited organisms. Exposure to radionuclide may be internal due to uptake/ ingestion or external due to gamma radiation. Internal exposure of terrestrial animals results due to ingestion & inhalation of contaminated food, soil or dust.[2] In this study accumulation of uranium in the rat tissue has been studied and corresponding dose rate was estimated.

Live traps were placed during night time at various locations of the tailings pond to catch rats. Twenty rat samples were collected, dried and wet digested using analytical grade reagents. The complete whole-body sample was processed and no different organ were digested separately. Uranium concentration were measured by UV-fluorimetry. A suitable aliquot of sample was evaporated & fused with a mixture of NaF (15%) and Na2CO3 (85%) at 700 °C in a platinum disc. The fused sample was measured for florescence intensity (565 nm) in a ECIL make uv-fluorimeter (365 nm).[3] Quality assurance of measurement were ensured by analysis of certified standard reference materials.[4] The activity concentration of uranium in the rat's tissue has been observed in the range of 0.056 to 0.21 Bq kg-1 with a median of 0.09 Bq kg-1 and stdev of 0.041 Bq kg-1. One outlier with the activity concentration of 0.88 Bq kg-1 has also been found. The estimated internal dose rate to rat's varied from 1.37 x 10-4 to 2.16 x 10-5 μGy h-1 with the median dose rate of 2.3 x 10-4 μGy h-1 by methodology adopted from Blaylock et al.[5] Using ICRP[1] approach, the internal dose to rats varied from 1.57 x 10-4 to 2.47 x 10-3 μGy h-1 with a median of 2.6 x 10-4 μGy h-1 (5.7 x 10-3 μGy d-1). The dose rate in rat samples is provided in [Figure 1]. The median activity concentration of uranium in the mixed tailings from the said area was estimated to be 1335 Bq kg-1. Using the median uranium concentration of 0.087 Bq kg-1 in rat the corresponding Bio-concentration Factor (BCF) of uranium has been worked out as 6.5 x 10-5. The dose rate to rats living in the uranium tailings pond area is insignificant and well below the dose threshold provided by USDOE of 1000 μGy d-1.
Figure 1: Dose rate to burrowing mammals (rat)

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The transfer of uranium to the tissue of rat is very negligible and hence the corresponding dose rate is well below the dose threshold. This may be due to the fact that the tailings contain trace level of uranium & subsequent tissue transfer is very insignificant.

Keywords: Dose, rats, tailings pond, uranium

  References Top

  1. Annals of ICRP-136. Dose Coefficients for Non-human Biota. 2017.
  2. IAEA. Technical Report Series 172. 1976.
  3. Eappan KP, Markose PM. Amine extraction for uranium estimation. Bull Radiat Prot 1986;9:83-6.
  4. IAEA. Tecdoc-1350. 2003.
  5. Blaylock BG, Trabalka JR. In: Lett JT, Alder H, editors. Advances in Radiation Biology. 1978. p. 103-52.

  Abstract - 12478: Activity concentration of 137Cs, 40K, 226Ra, 228Ra, 210Po in marine biota of Kalpakkam coast Top

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

Radiological and Environmental Safety Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

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

Assessment of natural and fallout radionuclides in the environment, their effects on living organisms and their safety has been one of the most important issues in radiological protection. The average activity concentration of the world ocean is 13.6 Bq/kg, which is mainly from naturally occurring radionuclides. Living organisms and plants actively accumulate them and magnify it to higher trophic strata. Intertidal marine biota, especially bivalves, crustaceans and seaweeds have been used broadly as first-order biological indicators for various contaminants, owing to their exposure and feeding habits. The results will form the baseline information on natural and fallout radionuclides for an array of native marine species. 20 different fish species, 4 crab, 2 prawn, 2 bivalve, 2 gastropod and 5 sea weed species were collected and analysed for 210Po natural. However, only 8 fish, 3 crab, 1 prawn and 5 seaweed species were analysed for gamma emitting radionuclides (137Cs, 40K, 238U, 232Th). Upon collection samples were cleaned and fresh weight was taken. The edible soft tissues were removed and lyophilized. The dried samples were homogenized and taken for analysis. For 210Po radiochemical separation was done after digesting with HNO3 in a microwave digester. Further, the samples were brought to HCl background before the electro chemical deposition of 210Po in the silver discs. The discs were counted in an alpha spectrometer. For gamma counting (137Cs, 40K, 226Ra, 228Ra) the samples were transferred to a bottle of standard geometry, sealed and kept for 28 days to ensure radioactive equilibrium between 226Ra, 228Ra and its daughter products. 210Po in the edible fishes ranged between 1.4-16.7 Bq/kg fw. Caranx (Parai meen) registered less activity conc. of 1.4 whereas Sardinella had the highest. Crabs had higher 210Po and were ranged between 23.6 and 96.7 Bq/kg fw. [Figure 1]. Between all the environmental matrices studied, Crabs had higher 210Po content followed by bivalves, prawn and gastropods. Prawn has 26.8 and the seaweeds had a range of 2.1-8.2 Bq/kg fw of 210Po. Among the intertidal organisms bivalves, (the filter feeders) had higher 210Po in the tissue as compared to other biota and sediment [Figure 1]. From the gamma emitting radionuclides, 137Cs was not detectable in any of the collected samples. 40K ranged between 27.4 (Ulva) and 311.3 (Chaeotomorpha) Bq/Kg fw. 226Ra ranged between 0.8 (Tuna fish) and 3.8 (Chaeotomorpha) Bq/Kg. Similarly, 228Ra varied between 1.4 (crab) and 31.0 (Chaeotomorpha) Bq/Kg.
Table 1: Gamma emitting radionuclides in different biota

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Figure 1: Mean–SD 210Po activity in different marine biota, sediment and water (Sediment in Bq/kg dw, and water in mBq/L

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Keywords: Activity, marine biota Kalpakkam, radionuclide

  Reference Top

  1. Saiyad Musthafa M, Arunachalam KD, Raiyaan GI. Env Chem Ecctox 2019;1:43-8.

  Abstract - 12482: Distribution of 210Po in soil and different parts of rice plant grown in the experimental field at Kalpakkam Top

S. Panigrahi, S. Gokul, S. N. Bramha, S. Chandrasekaran, C. V. Srinivas, B. Venkatraman

Radiological and Environmental Safety Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India

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

Rice (Oryza sativa) is a staple food in India and the soil-to-plant transfer factor of radionuclides present in the environment is an important parameter for estimating the internal radiation dose from food ingestion. In general, transfer factors show a large degree of variation dependent upon several factors such as soil type, species of plants, and other environmental conditions. In order to assess more precisely and realistically the internal radiation exposure to the public around nuclear facilities, site-specific parameters should be taken into account in the area concerned. Radionuclide transfer factors have been studied extensively for various components of food chains common to European and American countries, but studies on the transfer of radionuclides to rice are very limited. The IAEA has reviewed (IAEA TRS-472, 2010 and IAEA TECDOC, 2009) data on transfer factors for different foods and have compiled the available data on transfer factors (TF) for rice. Detailed studies on the evaluation of radionuclide transfer factors for rice grown in India are scarce. There are very few studies carried out on the west coast of India, but absolutely no reported values are there for the East coast. This paper presents soil to rice transfer factors for 210Po in rice grown in natural field conditions on the East Coast of India. An experimental field [Figure 1] has been developed within Kalpakkam DAE Campus for the radionuclides transfer factor studies in various crops. Different parts of the rice plant (root, stem, leaf, different stages of growth of grains) and soil were analyzed in triplicate (within 15 days of collection) to determine the soil-to rice/rice plant transfer factors. Upon collection, the root samples were thoroughly cleaned with tap water and then with MQ water to remove all the attached soil particles. At least 4 samples from each sector of the field [Figure 1] for soil and different plant parts were studied and the mean value is taken. Activity concentrations of 210Po in soil and different plant parts were analyzed in an alpha spectrometer after auto-deposition of the acid-digested samples.[3] The activity of 210Po in soil was 8.5 – 16.67 Bq/Kg [Figure 1]. 210Po in rice was recorded as 2.32 ± 0.7 (range:1.4-2.9) Bq/kg in dw whereas, in wet weight (ww) it was observed as 1.76 ± 0.5 (range:1.2-2.3) Bq/kg. TF for 210Po in rice grain and parts are presented in Table-1, for dw and ww. It was observed that the root has higher activity and was followed by stem leaf and grains [Table 1]. Interestingly, the early stage of the grain was having higher activity as compared to the fully grown grains. The observed value of 210Po in rice was comparable with other reported values from India. TF of 210Po was less in the grains and was followed by the leaf, stem, and root. In dw measurements, it varied between 0.18 – 1.47, whereas, in ww measurements, it varied between 0.13 – 0.43. The parts of the rice plant that are more in contact with the soil have higher TF. Transpiration of water from the soil plays a significant role in the accumulation of radionuclides. The milking grains are having more water content and have higher activity. While the grown-up grains have less water content and less 210Po content. This forms the first information on 210Po TF in soil-rice from Kalpakkam.
Figure 1: Radionuclide transfer factor experimental field at DAE site, IGCAR (appx. 30 m × 40 m)

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Table 1: Activity concentration and site.specific transfer factor of 210Po in the soil.rice compartment at Kalpakkam

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Keywords: 210Po, Kalpakkam, rice, transfer factor

  References Top

  1. IAEA. Technical Report Series (TRS) No. 472. Vienna: International Atomic Energy Agency; 2010.
  2. IAEA. TECDOC Series No. 1616. Vienna: International Atomic Energy Agency; 2009.
  3. Panigrahi, et al. Marine Pollution Bulletin. Part B. Vol. 173. 2021. p. 113147.

  Abstract - 12571: Radioecological study of domestic hen and egg samples collected near the Kudankulam Nuclear Power Project  Top

Jeben Benjamin1, R.P. Praveen Pole1,2, S. Godwin Wesley1, Thomas George3

1Department of Zoology, Scott Christian College (Autonomous), Nagercoil, 2Department of Zoology, V. O. Chidambaram College, Tuticorin, 3Environmental Survey Laboratory, KKNPP, Kudankulam, Tamil Nadu, India

E-mail: [email protected]

The domestic hen, Gallus gallus domesticus, is a free-ranging omnivorous bird reared for its meat and eggs. Pentreath[1] has advocated certain vital parameters for considering 'reference organisms' in radioecological studies and these birds and their eggs have been found to fulfill them. Natural radionuclides such as 226Ra, 228Ra, 210Po, 210Pb and 40K are very important radioecologically, as they contribute significantly to natural background radiation dose to biota. The objectives of the study were: (i) to estimate the activity concentrations of 226Ra, 228Ra, 210Po, 210Pb and 40K in the birds and their eggs; and (ii) to calculate the radiation dose due to these radionuclides to the birds and the eggs. Hens (n=6; 3-4 hens per sample) and eggs (n=6; 50-60 eggs per sample) were collected from villages within 30 km radial distance of the Kudakulam Nuclear Power Project. The study was conducted between 2009 and 2012, i.e. before the start of the operation of the Power Project. The hens were killed and the eggs were cracked and dried in a hot air oven separately until a constant weight was achieved. The dry samples were homogenized and subjected to both dry and wet digestion for estimating radionuclides. In the dry method, samples were made to ash, hermetically sealed in vials and analysed for 226Ra, 228Ra and 40K in HPGe; another portion (10 g) of the dried sample was wet-digested using HNO3, HCl and H2O2 and 210Po electrodeposited onto a silver disc. Then the disc was counted for alpha particles. The solution which had been plated for 210Po was stored for a minimum of 6 months and again plated for 210Po (ingrown from grandparent, 210Pb). The 210Pb activity of the sample was thus calculated. Dose conversion coefficients (DCCs) of Ulanovsky and Prohl[2] were used for calculating dose. The internal dose rate was a product of the activity concentration of a radionuclide in the organism and the internal DCC. Similarly the external dose rate was a product of the activity concentration of a radionuclide in the substrate (soil) and the external DCC. The total dose is the sum of internal and external doses.In both the hen and egg samples, 40K recorded the highest activity concentration. But transfer factor values revealed that 210Pb was taken up more by the hen followed by 210Po. Both radium nuclides were found to have lower concentrations in the both the samples. The significant finding of this study is that the external dose rate is higher than or equivalent to the internal dose rate. Yet, the total dose rate is almost 3 orders of magnitude lesser than the US DoE (2002)- recommended dose rate (reference dose) of 1 mGy day-1 for terrestrial animals. Hence the animals are safe radiologically. This study was supported by the Department of Atomic Energy-Board of Research in Nuclear Sciences (Sanction No. 2008/36/57/BRNS/3019).
Table 1: Activity concentration (Bq kg-1 f.w.), transfer factors and dose rates (mGy day-1)

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Keywords: Hen and eggs, natural radionuclides, radioecology, reference organism

  References Top

  1. Pentreath RJ. A system for radiological protection of the environment: Some initial thoughts and ideas. J Radiol Prot 1999;19:117-28.
  2. Ulanovsky A, Pröhl G. Tables of dose conversion coefficients for estimating internal and external radiation exposures to terrestrial and aquatic biota. Radiat Environ Biophys 2008;47:195-203.
  3. United States Department of Energy (US DoE). A Graded Approach for Evaluating Radiation Dose to Aquatic and Terrestrial Biota. United States Department of Energy-Std-1153. 2002. p. 58.

  Abstract - 12572: Natural radiation dose to the marine mollusc Sepia pharaonis and assessment of radiological risk Top

R. P. Praveen Pole1,2, Jeben Benjamin1, S. Godwin Wesley1, B. Vijayakumar3

1Department of Zoology, Scott Christian College (Autonomous), Nagercoil, 3Environmental Survey Laboratory, Kudankulam Nuclear Power Project, Radhapuram, 2Department of Zoology, V. O. Chidambaram College, Tuticorin, Tamil Nadu, India

E-mail: [email protected]

Radioecological assessment of marine environment receives attention as radionuclides contribute as much to marine pollution as heavy metals, plastics, etc. All marine biota are exposed to natural radiation and unavoidably receives radiation dose. Studying the biogeochemistry of natural radionuclides of 238U and 232Th origin and the primordial radionuclide 40K in the marine environment is important since organisms concentrate radionuclides in their body disproportionately to their body size. Different factors like feeding behaviour and metabolism of the organisms determine the uptake of radionulides. In this scenario, the cuttle fish, Sepia pharaonis (a reference organism) was considered to estimate the activity, concentration factor, radiation dose and associated risk due to radionuclides such as 226Ra, 228Ra, 210Po, 210Pb and 40K. The present study was carried out along the southeastern coast of India (around Kudankulam Nuclear Power Project) between June 2010 and May 2012. Sepia samples (six samples; n=3 for each collection) were obtained from fish-landing centres along the coast. After completing pre-concentration, the ash samples were packed in vials and left undisturbed for 1 month to allow the attainment of secular equilibrium between the parent and daughter radionuclides of the uranium and thorium series. Gamma spectrometry (HPGe) was employed for estimating 40K, 226Ra and 228Ra. After wet digestion of the sample, 210Po was plated onto a silver disc by chemical deposition; the disc was alpha-counted. The solution left after plating was kept undisturbed for more than 6 months, allowing 210Po in-growth from 210Pb; after this period, 210Po was again plated and the 210Pb activity was calculated. The internal and external radiation dose was calculated by using the formula:

where Dj is the dose to the organism j; DCCi is the dose conversion coefficient and Ci,o is the activity of the radionuclide i in the organism. DCCs were taken from Ulanovsky and Pröhl.[1] As seen from [Table 1], there is a vast difference in the activity concentration of these radionuclides. This may be due to the reasons like speciation of the radionuclide in the medium and the feeding pattern of the cuttlefish. The concentration factor values show that 210Po is taken up by the cuttle fish in higher concentrations from the medium. The total radiation dose (sum of internal and external dose) to this molluscan species was in the range 4.8×10-4 to 1.31×10-3mGy day-1. The radiological risk due to all these radionuclides was calculated and it was far below the reference dose (10 mGy day-1) for aquatic organisms.[2] This work was funded by the Board of Research in Nuclear Sciences (BRNS), Department of Atomic Energy (DAE), Government of India under the sanction No. 2008/36/57-BRNS.
Table 1: Radiological data of Sepia pharaonis

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Keywords: Dose conversion coefficient, natural radiation, radiation dose, reference organism, screening index

  References Top

  1. Ulanovsky A, Pröhl G. Radiat Environ Biophys 2008;47:195-203.
  2. United States Department of Energy (US DoE, 2002). A Graded Approach for Evaluating Radiation Dose to Aquatic and Terrestrial Biota. USDoE-std-1153. Module 3 Methods Derivation. United States Department of Energy; 2002. p. 58.

  Abstract - 12576: Site specific assessment of radio activity in waste disposal facility at Orissa Sand Complex Top

Annapurna Rout1, A. Sahu1, P. Prusty1, R. Patra1, 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]

Orissa Sand Complex (OSCOM), Chhatrapur, Odisha is the largest division of IREL for production of heavy minerals. The monazite heavy mineral is the major source of commercial rare earths. The Rare Earth Extraction Plant (REEP) is engaged in commercial Processing of monazite to produce rare earth chloride. During the monazite processing use of various chemicals lead to generation of solid, liquid as well as gaseous waste. In order to deactivate the rare earth chloride solution co-precipitants such as barium chloride and magnesium sulphate are used. Finally these chemicals come out with loaded radio activity and produces radioactive solid waste. Proper management and disposal of these radioactive wastes is a major concern owing to its direct impact on occupational and environmental radiation safety. All administrative and operational activities involved in the handling, pre-treatment, treatment, conditioning, transportation, storage and disposal of radioactive waste are regulated by Atomic Energy Regulatory Board (AERB) under the Atomic Energy (Safe Disposal of Radioactive Wastes) Rules, 1987.[1] Therefore monitoring the waste disposal facility is an important task that need to be carried out routinely.

At REEP, based on chemical composition there are four types of solid wastes. They are i) insoluble muck ii) lead barium slurry iii) iron carbonate cake and iv) ETP slurry. These solid wastes are routinely monitored with respect to their alpha, beta and Ra228 activity. There are engineered reinforced cement concrete trenches for safe and long term disposal of solid radioactive waste. But considering the possibilities of spillage, breech in the concrete, extreme meteorological conditions, geomorphology of the site and its proximity to water bodies there is a need of routine radio activity analysis of surrounding soil and water.

In this regard, routine sampling of soil and water samples were carried out surrounding the solid waste trenches. Basically alpha beta dual counters and High purity Germanium detectors are used for radio activity studies. Here the results are discussed in [Table 1]. The gross α activity of the soil varies from 0.09 to 0.4 Bq l-1 depending on the waste disposal site. But as a whole it was observed that the radio activity of soil surrounding the SWT area were comparable with the background activity and also these area are well known for its high background. In addition to this spillage of solid waste during transportation and disposal contributes the maximum activity, which used to be taken care by cleaning of that area or soil decontamination. There was no distinct level of increase in activity that will indicate the presence of reasonable amount of radio nuclides. Similarly water samples collected from bore wells in these area were also monitored and as per to [Table 2] it was observed that the activity and uranium content were in corroboration with background activity level.
Table 1: Radioactivity analysis of soil over a period of 6 months, soil from nearby area of solid waste trench site

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Table 2: Radioactivity analysis of groundwater

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The above studies were in well agreement that disposal of waste in properly designed trenches are very much effective as it controls contamination as well as migration of radio nuclides to soil and water bodies.[2] In future this study can probe the physical durability of the concrete walls (trench wall) that stands as a barrier between solid waste and underground soil layer. The above routine analysis were conducted to ensure the environmental radiation safety at waste disposal site.

Keywords: High background area, monazite, solid waste trench

  References Top

  1. AERB Safety Guide No. AERB/NF/SG/RW-5, August 2007.
  2. Barakos G, Mischo H, Gutzmer J. Rare Earth Industry Technological, Economic and Environmental implications. 2016. p. 121-38.

  Abstract - 12618: Distribution of tritium in wild plants after short-term exposure of tritiated water vapor Top

C. A. O. Shaofei, L. I. Jianguo, Yuan Han

China Institute for Radiation Protection, Taiyuan, China

E-mail: [email protected]

Tritium is a radioactive isotope of hydrogen, and its physical and chemical properties and migration behavior in the environment are similar to hydrogen. Tritium mainly exists in the form of tritiated water (HTO) in the environmental media. Plants absorb tritium mainly through the following two ways:[1] leaves absorb it from the air, and roots absorb it from the soil. Part of the HTO entering the plant body forms organically bound tritium (OBT) through photosynthesis, while the remaining part returns to the atmosphere or remains in the plant in the form of tissue free water tritium (TFWT).[2] In order to explore the distribution of tritium in typical wild plants in Northwest China,the accumulation behavior of tritiated water (HTO) vapor in two wild plants (Tamarix ramosissima Ledeb.,and Alhagi sparsifolia Shap.) after short-term release was studied in a sealed environmental chamber. The experimental plants were cultured in pot (inner diameter×height: 25.5 cm×29.5 cm).In order to meet the release conditions of the source term and realize the real-time monitoring of environmental parameters such as temperature and humidity, the short-term release experiment of HTO was conducted in a closed environmental chamber. The results show that the distribution of tritium after short-term HTO release varies with plant species and organs. For the same plant species, the activity concentrations of TFWT and OBT in different organs are as follows: leaf > stem> root. The activity concentrations of TFWT in different parts of the three experimental plants show a decreasing trend with the growth of plants, especially in leaves. The activity concentrations of OBT in the leaves of two experimental plants decrease significantly with the growth of plants, and the decreasing range is from large to small, while the activity concentrations of OBT in the stems increase first and then decrease.
Figure 1: Environmental chamber for HTO short-term exposure

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Figure 2: Changes of tissue free water tritium, organically bound tritium activity concentrations with time in different parts of plants

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Keywords: Short-term exposure, tritiated water, tritium concentrations, wild plants

  References Top

  1. Boyer C, Vichot L, Fromm M, et al. Environ Exp Bot 2009;67:34-51.
  2. Choi YH, Lim KM, Lee WY, Diabaté S, Strack S. Environ Radioact 2002;58:67-85.

  Abstract - 13361: Effect of the aerosol particle size distribution on the radiological impact assessment of radiological dispersal device detonations Top

B. Sreekanth, S. Anand1, M. K. Sharma2, Probal Chaudhury, B. K. Sapra2

Radiation Safety Systems Division, Bhabha Atomic Research Centre, 1Health Physics Division, Bhabha Atomic Research Centre, 2Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Radiological Dispersal Device (RDD), commonly known as a 'dirty bomb', combines chemical explosive and radioactive material. The detonation of RDD generates a radioactive aerosol cloud, and the meteorological conditions disperse, transport and contaminate air and surface over a large area in the downwind direction. The contamination exposes the public for a prolonged period and delivers radiation doses. Radiological Impact Assessment (RIA) of RDD detonations is a major concern for the radiation emergency preparedness and response planning authorities worldwide. Therefore, dose assessment is crucial in RIA for effectively planning and implementing the required countermeasures. It is generally achieved by estimating contamination levels using atmospheric dispersion modelling. The aerosol cloud depletes through deposition processes, which is highly sensitive to the constituent particulate size as the deposition velocity (vd) varies a few orders of magnitude for the complete range of the particulate size. However, most studies in the literature considered the particulate size distribution as monodisperse, whereas a vast amount of the literature proved it as polydisperse. Di Lemma et al.[1] gave the particulate size distributions for vital radioactive materials concerned with RDD. Therefore, it is interesting to study the quantitative effect of the polydispersity on the depletion of cloud and the resultant implications on the dose estimation.

The current paper presents the effect of the polydispersity in RIA for RDD scenarios using a case study of RDD (10 kg TNT and 4x105 TBq of CsCl). The simulations have been carried out using the HOTSPOT health physics code.[2] HOTSPOT divides the particle size distribution into two bins: respirable (diameter <10 μm) and non respirable (diameter >10 μm). The user has to provide the respective mass fractions. HOTSPOT can be used for monodisperse approximation by setting the mass fraction of non-respirable as zero. Based on the Di Lemma et al.[1] results, the total airborne mass/activity can be divided into multiple fractions, and each fraction has a different particle size and respective deposition velocity. Thus, the polydisperse cloud is treated as a combination of multiple monodisperse constituent clouds of different particulate size such that the contamination levels caused by a polydisperse cloud is the summation of contamination levels caused by respective monodisperse clouds. To emphasize the effect, for a given airborne activity, the simulations have been carried out for three different approximations, they are: i) the particles are monodisperse and respirable (vd=0.3 cm/s), ii) the particles are monodisperse and non-respirable (vd=40 cm/s), and iii) the particles are polydisperse and size-dependent deposition velocity is used. Unstable metrological condition (Pasquill Gliffod category –A) in a city terrain is considered for the simulation.

The results [Figure 1] show that the two extreme monodisperse approximations showed significant deviation from the realistic polydisperse approach. These approximations can lead to substantial inefficiency in the implementation of countermeasures; for example, based on polydisperse approach the downwind distance for implementation of the temporary relocation (dose 50 mSv in the first week, ICRP 96) is ~58% higher than respirable monodisperse approximation and ~30% lower than the non-respirable monodisperse approximation. In conclusion, the study clearly shows the effect of the polydisperse nature of the aerosol in estimating the dose levels and hence recommends the current approach for a more RIA.
Figure 1: Comparison of the ground shine dose rate for monodisperse and polydisperse approaches for RDD. (10kg TNT and 4x105 TBq CsCl). RDD: Radiological Dispersal Device

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Keywords: Aerosol, deposition, dispersion, polydisperse, radiological impact

  References Top

  1. Di Lemma FG, et al. JNST 2016;53:391-401.
  2. Homann SG. UCRL-MA-106315. Livermore, Ca: LLNL; 2011.

  Abstract - 14556: IRPA task group for communications and engagement with the public on radiation and risk Top

H. Yoshida, Task Group Members1

Task Group Chair, IRPA executive Council, Cyclotron and Radioisotope Center, Tohoku University, Sendai, Japan, 1IRPA Task Group on Public Understanding

E-mail: [email protected]

Enhancing public understanding of radiation and risk is highlighted by experiences from past emergencies, including the accident at TEPCO's Fukushima Daiichi Nuclear Power Plant in 2011 and the following post-disaster recovery, as one of the most important challenges, and this challenge is common across almost all public interfaces regarding radiation and risk. There is a growing need and interest for the IRPA Associate Societies (AS) to enhance their programmes in this important area. This is a key but challenging activity which needs further support. To this end IRPA has been continuing a Task Group (TG) activity for Public Understanding on radiation and risk since 2013. It has become very clear that public understanding, trust and consent are central to ensuring effective and proportionate radiation protection without unduly limiting the safe use of medical, scientific and industrial radiological practices for the benefit of mankind. IRPA strongly believes that all radiation protection professionals and the radiation protection societies have a duty to engage with the public. In order to enthuse all of us in our profession to become more active public advocates for radiation protection and to provide information, experiences and techniques to help us to become more effective and comfortable in this challenging task, IRPA published 'Practical Guidance for Engagement with the Public on Radiation and Risk' on the IRPA website in October 2020 [Figure 1],[1] which was made based on the outcome of a series of workshops in various regions of the world and consultations of the AS. The guidance document provides information on a range of specific situations as follows; medical exposures, non-ionising radiation, radon-related issues, radioactive waste management, radiological and nuclear emergencies, malevolent use of a radiation source, and post-accident and long-term recovery situations. In each situation, practical tips for effective communication and specific support reference material are introduced. Public engagement can include providing scientific knowledge and stimulating interest in radiation, as well as the engagement with relevant media, politicians and other key influencers. The guidance document aims to provide help in addressing all eventualities. Following these activities, TG activity for new term (-2024) has started with the objectives of encouraging and supporting the IRPA AS in the development of effective means of enhancing public understanding of radiation risk through the sharing of good practice, ideas and resource material.
Figure 1: IRPA Practical Guidance for Engagement with the Public on Radiation and Risk[1]

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Key tasks are as follows:

  • To keep collecting good practices, needs and views.
  • To promote the work of IRPA in this area to our colleagues.
  • To encourage the AS to organize internal meetings-workshops to share the IRPA Guidance, to assist the AS and individual professionals to better understand the challenges of communication, and to keep collecting good practices, needs and views.
  • To consider emerging issues, e.g., how we cope with the spread of information through social media, Twitter, Facebook, and other social networking sites especially misinformation and rumours through social media? These might lead fears about radiation and harmful rumours.

In this presentation, recent TG activities for the key tasks will be reported.

Keywords: Communications, engagement, radiation, risk, the public

  Reference Top

  1. IRPA; 2020. Available from: https://www.irpa.net/members/IRPA%20Guidance%20Public%20Engagement.pdf.

  Abstract - 15359: Use of high-resolution satellite data for quick estimation of population near nuclear power plants for use during nuclear emergencies Top

D. G. Mishra, Manish K. Mishra, Vandana Pulhani, A. Vinod Kumar

EMAD, Health Safety and Environment Group, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

India follows a decadal census (started in 1881) which is carried out by visit and verification of each individual in each household. It is an exhaustive and expensive exercise consuming immense time, human and infra resources. Population data from decadal censuses along with calculated growth-rate can be used for calculating population for a given year. Remote-sensing satellites providing high-resolution imagery with clearly detectable population settlements can be used for quick estimation of the population (independently of the censuses) depending upon several geospatial parameters. Such estimates work well for requirements which do not need an accurate measurement of population e.g., emergency preparedness, evacuation, prophylaxis, dose-evaluation, etc. The European space agency (ESA) has recently made public a world LULC (Land-use and Land-cover) data, at 10-meter spatial resolution, [Figure 1], which provides a detailed raster image (singe band) for tree/forest cover, human settlements (built-up), sparse vegetation, permanent water bodies and so on.[2] This raster data is provided in a Coordinate Reference System (CRS) of WGS 84-EPSG 4326, datum. The CRS was then projected from a degree-based system to EPSG-3857 (Pseudo-Mercator) for measurement in meters. After the transformation, the raster image was clipped using mask layer (e.g., 30 km area around NAPS, in which population estimation is required). The clipped raster was then vectorized as polygons in QGIS (ver. 3.24) using raster processing tool, into vector format (QGIS, 2018).
Figure 1: Land-use land-cover image from European Space Agency and processed image of settlement (in red) for population calculation

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The resultant polygons were then collected based on similar 'digital number' (DN) derived from the original raster, and were categorized accordingly. The 'field calculator' was then used to determine the total area covered by the 'built-up' portion of the image. The total area was divided by the average area one household (HH) typically requires (around 45 sq. meters in rural areas), resulting in the total number of households. The resultant household number is then multiplied with individual per household (6 individual per HH is Indian average, but it varies within states and districts). This gives us a workable estimate of the population on the date at which satellite imagery is taken. We have used this model for successfully predicting population near Narora Atomic Power Plant. We validated the data with censuses 2001 and 2011 and average growth-rate. A minor accumulative area from roads, workshops, commercial locations, factories etc. gets inadvertently added up into the built-up area which can be removed by using symmetrical difference tool. This model can be developed to predict population to a high accuracy based on imagery or point clouds with a spatial resolution of 1 m from commercial satellite or to a few centimeters from LiDAR surveys, respectively. The steps can be summarized and developed into a graphical modeler which can be used to automate following all the steps given below:

Raster Image → Warp Projection → Clip by Mask Layer → Vectorize → Collect Geometries → Calculate Area

Keywords: Census, dose, population

  References Top

  1. QGIS. QGIS, Supported Data Formats. QGIS; 2018. Available from: https://docs.qgis.org/2.8/en/docs/user_manual/working_with_vector/supported_data.html.
  2. Zanaga D, Van De Kerchove R, et al. ESA WorldCover 10 m 2020 v100. 2021. [Doi; 10.5281/zenodo.5571936].

  Abstract - 15476: Precise detection of neutron source position using multi position sensitive detector orthogonal assembly Top

Shraddha S. Desai1, Mala N Rao1,2

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

E-mail: [email protected]

Neutron flux mapping and area monitoring are carried out using He3 filled efficient neutron detectors. Point detectors are used for portable monitoring, due to the advantage of the compact pulse processing and data acquisition. Whereas position sensitive detector (PSD) presents a faster and simultaneous data about neutron flux monitoring at nuclear facilities and ports in public area. A method for accurate source location is presented here by unfolding the position sensing data from two orthogonally arranged PSDs. This instrument facilitates the time dependent data to explore the source movement and the direction of the transport. Decoding the data from position output is made convenient for portable area monitor.

Principle of Operation: PSDs are useful to record the variation in neutron flux distribution over the sensitive length (Desai 2013).[1],[2] Intensity distribution recorded in two orthogonal PSD [Figure 1] indicates the direction of the incident neutron from a source (Desai 2021).[3] Further the position data over the sensitive length of the PSD is analysed for the source location from the reference point of the monitoring instrument.
Figure 1: (a) Schematic of the orthogonal PSDs and a point source C in 2nd quadrant. (b) Position spectrum of a PSD1 with uniform background and a point source. PSD: Position sensitive detector

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Experimental Method: Two PSDs 1 and 2 are chosen with identical efficiency and position resolution. PSD1 and PSD2 are mounted with the orthogonal geometry as shown in [Figure 1]a. Arrangement is made at perfect right angle and origin matching to geometrical as well as electronics centre. PSD data for the source position detection:

  • Position output from each of the 1 m long PSDs is recorded for equal times simultaneously
  • PSD length 1000 mm corresponds to 256 channels position output. Neutron flux is integrated over 4 mm of length of PSD
  • The data is in the form of neutron intensity listed with 1-256 channels
  • Neutron intensity is integrated over 1-128 and 129-256 channels from PSD 1 and 2 and flux mapping is done using integral value.

Source Position Analysis: These four values represent integrated neutron flux from +x, -x, +y and -y axis forming four quadrants of the position reference. Higher intensities of these data sets are indicative of the quadrant of the source location. Neutron source shown in [Figure 1]a is in quadrant 2 as the intensity in -x and +y section is higher compared to the pairs of data indicating quadrant with higher integral intensity. Position data array is processed with the ratio of y/x for each channel value in the quadrant of interest. Value of tan-1(y/x) =θ is averaged over the data to get direction of the source. Pu Be Extended source 30 cm long Placed at 2 m away parallel to PSD1 and a point source Am-Be at position (30 cm, 30 cm) from origin of the multiPSD assembly. [Figure 1]b shows the neutron intensity resulting from both the flooded and point sources. The hump at channel ~75 overriding the flooded uniform intensity is seen due to source position closer than PSD length and it is directly detected using the peak positions. y/x values other than peak area are used for the location of the flooded source. Spectra appears as a ramp when source is placed away from the multiPSD assembly.

Conclusion: Two PSDs in orthogonal geometry indicate quadrant of the neutron source and ratio of intensities from respective channels gives estimation of polar coordinates of the source (r, θ). It is useful to identify the location of unknown source position and time dependent variation in the integral flux indicates source movement.

Keywords: Area monitoring, gas filled proportional counter, neutron detection, neutron flux

  References Top

  1. Desai SS. PhD Thesis, University of Mumbai; 2006.
  2. Desai SS. J Phys 2014;528:012037.
  3. Desai SS, et al. IARPNC-2020 paper no 246; 2020.

  Abstract - 15610: Nuclear accident consequence assessment technology research and software development Top

Lyu Minghua, Yao Rentai, Niu Yanjing, Zhang Junfang, Zhao Duoxin, Huang Sha, Guo Huan, Li Mingye

Department of Nuclear Environmental Science, China Institute for Radiation Protection, Taiyuan, China

E-mail: [email protected]

Nuclear emergency response is the last barrier to the nuclear security depth of defense system. It is a means to mitigate the consequences of serious accidents, natural disasters and terrorist events in nuclear facilities and activities. Nuclear accident consequence assessment technology and software system are quite important technical means for decision-making during nuclear emergency response. In view of the complex, fast and multi-scale characteristics of the migration process of radioactive pollutants in the atmosphere, CIRP has developed a national nuclear accident consequence assessment and decision support system (NACADOS) in order to meet the needs of radiation consequence assessment and response of emergencies in China and abroad, and improve the ability of nuclear emergency response. By studying multi-scale weather forecasting methods, airborne radioactive material migration and diffusion simulation technology and comprehensive evaluation of system effectiveness, NACADOS includes multi-scale model chain that suitable for China's national conditions.

  • By studying the dynamic down-scaling method and the coupling technology between different modules of different meteorological modules, a multi-scale meteorological prediction module is established to realize the multi-scale meteorological prediction simulation function at any point and at any time in the world, and provide a grid meteorological field for the subsequent multi-scale diffusion/dose calculation.
  • Considering the actual emission scene, including the simultaneous emission superposition effect of multiple release sources at multiple reactor sites, and the effects of long -term continuous release, the diffusion module applicable to different scales and environmental conditions has been developed, including the trajectory/particle module applicable to large scales Lagrangian trajectory model, random walk particle model and CFD module considering the influence of complex terrain and meteorological conditions. Meet the simulation of multi -source, variable sources, and mobile sources, and realize the multi -scale migration and diffusion simulation of any global and any time.
  • Through comparison with on-site monitoring data, excellent tracing experiment results (such as ETEX, Urban2000, etc.), and excellent consequence evaluation system (JRODOS, RASCAL, etc.) simulation results, the system's function, performance and effectiveness are tested in all aspects from the effectiveness verification of physical modules, system testing, and engineering software testing. The results show that the system calculation results are true and accurate, and the system operation is stable and reliable.
Figure 1: Urban dispersion models of NACADOS

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Figure 2: Left: UNSCEAR 2020 report (radiation level and impact caused by Fukushima Daiichi nuclear power plant accident since UNSCEAR 2013 report) Right: Cs-137 ground sediment concentration of NACADOS

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The NACADOS can obviously reflect the processes in the UNSCEAR report. The coincidence factor between the system simulation results and the detection data is significantly improved, with the coincidence degree reaching 80% within 10 times, 68% within 5 times, and 58% within 3.5 times, 33% within 2 times.

Keywords: Migration diffusion, multi-scale, nuclear accident assessment, validation

  Reference Top

  1. Nasstrom JS, Sugiyama G, Baskett R, et al. The National Atmospheric Release Advisory Center (NARAC) modeling and decision support system for radiological and nuclear emergency pre-paredness and response. Int J Emerg Manag 2005.

  Abstract - 16223: An overview of mathematical models to assess the dose conversion factor of radionuclides for non-human species, exposed to ionising radiation Top

Ajay Kumar

BRNS Secretariat, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

The heterogeneity and diverseness of non-human species (flora and fauna) is a specific challenge in developing and applying dosimetric models to assess the effects of exposure to ionizing radiations. In the recent past years, numerous studies have been carried out, on the estimation of dose conversion factors of radionuclides for non-human species, exposed to ionising radiation. The concept of radiation protection of non-human species was first formulated in International Commission on Radiological Protection (ICRP) Publication 26[1] and reproduced with minor changes in ICRP Publication 60[2] followed by ICRP Publication 91.[3] Overall, the radiation protection to the non-human species was based on human, assuming that if man is adequately protected, then other living things are also likely to be sufficiently protected. However, these policies of ICRP was related only to the routine practice and never extended to accidental situations. Following the Chernobyl nuclear accident in 1986, the concept was entirely changed, when non-human species were found to be highly exposed than man and the radiation standards set for man, could not provide adequate protection for non-human species. Consequently, the ICRP Publication 96[4] considered the release of radionuclides in the environment, under two conditions: normal and accidental and accepted the environmental radionuclides concentration for non-human species, as the Reference Animals and Plants (RAPs). The radiation safety standards for human are based on the principle of maximum possible restrictions of stochastic effects. However, for non-human species, no consensus has been achieved with respect to their radiological doses as well as impact of radiation effects on non-human species. Subsequently, the Commission has considered the most common endpoints (early mortality, morbidity, reduced reproduction success and deleterious heredity effects), in non-human species while assessing the radiation effects.[3],[4] Many authors have attempted to derive the dose conversion factors for non-human biota using general screening model.[5],[6],[7] This model has derived biota concentration limits for various radionuclides in environmental media using dose conversion factors and Critical Dose Value (CDVb) for biota. The critical dose values, set for terrestrial plant, terrestrial animal and aquatic animal were 3.65 Gy/y, 3.65 Gy/y and 0.37 Gy/y respectively. However, dose conversion factors were estimated, based on screening levels. In continuation, the ICRP publication 108[8] has identified a small set of 12 Reference Animals and Plants (RAPs) and developed their derived consideration reference dose levels in terms of mGy/day. Similarly, ICRP publication 124[9] and ICRP publication 136[10] have also concurred the above recommendations and updated the similar data of dose conversion factors for RAPs in terms of μGy/h/Bq/kg, averaged over the mass of the organism. Due to large data gaps in dose-effect studies on non-human species, the available current models for dose conversion factors, bioaccumulation factors, dose calculation methods etc.) of radionuclides are inadequate and hence an extensive study is required to cover more organisms and radionclides.

  References Top

  1. ICRP. Annals of ICRP. ICRP Publication 26; 1977. p. 1.
  2. ICRP. Annals of ICRP. ICRP Publication 60; 1991. p. 21.
  3. ICRP. Annals of ICRP. ICRP Publication 91; 2003a. p. 33.
  4. ICRP. Annals of ICRP. ICRP Publication 96; 2005. p. 35.
  5. Amiro BD. J Environ Radioact 1997;35:37-51.
  6. Higley et al. J Environ Radioact 2001;66:41-59.
  7. DOE (US Department of Energy). US Department of Energy, Final Technical Standard No. DOE-STD-1153-2002. Washington, D.C: 2002.
  8. ICRP. Annals of ICRP. ICRP Publication 108; 2008b. p. 38.
  9. ICRP., Annals of ICRP. ICRP Publication 124; 2014a. p. 43.
  10. ICRP. Annals of ICRP. ICRP Publication 136; 2017. p. 46.

  Abstract - 16356: Studies on radiological dose to aquatic weed at Kakrapar atomic power station site using ERICA integrated approach Top

A. K. Patra, S. S. Wagh, I. V. Saradhi1, A. Vinod Kumar1

Environmental Survey Laboratory (ESS, EMAD, BARC), Surat, Gujarat, 1Environmental Monitoring and Assessment Division, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Low level radioactive liquid waste which contains mainly 3H, 134+137Cs, 90Sr, 60Co etc is generated during operation and maintenance of PHWRs. The low level liquid waste once discharged into the aquatic system, undergoes dilution/dispersion in the lake. Subsequently accumulation by different aquatic matrices such as sediment, weeds and fishes takes place. Assumption that if controls are applied to sufficiently protect humans, then the native non-human biota are likely to be adequately protected,[1] is not supported by scientific data. Hence, it is necessary to prove exclusively that the nonhuman biota is sufficiently protected from ionizing radiation. Internationally, few studies on nonhuman biota are available,[2],[3] however only limited data are available for Indian conditions. An attempt was made to evaluate the radiological dose to aquatic weed using ERICA Integrated Approach. This study was carried out at Kakrapar site where two PHWRs having capacity of each 220 MWe are operating since the year 1994. In Moticher lake, large volume of water is always available for further dilution of discharged liquid waste into the lake. Silt and weed samples were collected at different distances (10-100 m) from the liquid waste discharge point of KAPS at Moticher Lake. Hydrilla verticillata is the predominantly available species. The representative aquatic plant (weed) was identified as Hydrilla verticillata in the aquatic ecosystem of KAPS. 23 nos. of silt and weed samples were processed and packed in a suitable geometry and counted by HPGe detector of 100 % Relative Efficiency. Site-specific distribution coefficient (Kd), sediment, water and weed activity were measured for 134Cs, 137Cs and 60Co. Radiological dose to aquatic weed was evaluated using ERICA Integrated Approach. Based on the physical observation, the occupancy factor of the weed in the habitat is assigned as 0.15 on water surface; 0.75 in water and 0.1 in sediment surface. It is assumed that the Hydrilla verticillata is equivalent to vascular plant mentioned in the ERICA tool and the Dose Conversion Coefficient (DCC) of vascular plant is used for the dose rate evaluation. Using the input data, external and internal dose rate is calculated using ERICA tool. The observed total dose rate per organism is found to be 9.4E-5 – 1.2E-3 μGy/h and compared with other locations as mentioned in [Table 1]. The total dose rate per organism is well within the screening dose rate level (10 μGy/h).[3]
Table 1: Comparison of radiological dose rate for nonhuman biota with other data

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Keywords: Aquatic weed, ERICA, Kakrapar, radiological dose

  References Top

  1. ICRP. Ann ICRP 1990;21:1-3.
  2. FASSET, 2004. Available from: http://www.fasset.org.
  3. ERICA, 2006. Available from: http://www.erica-project.org.
  4. Wesley SG. (2009-2013), Final Project Report Submitted to BRNS, DAE, Mumbai. Sanction No.: 2008/36/57/BRNS/3019 Dated 06 March, 2009.
  5. Zinger I, Copplestone D, Howard BJ. J Environ Radioact 2008;99:1510-8.

  Abstract - 16365: Review on reference animals and plants for assessment of radiological risk to non-human biota from India biogeographic prospective Top

N. P. I. Das, V. Subramanian, B. Venkatraman

Aerosol Transport and Biodiversity Section, Radiological and Environmental Safety Division, Safety, Quality and Resource Management Group, IGCAR, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

With the development of our understanding on radiation protection of man and environment, the necessity was felt to demonstrate the protection of non-human species in natural environment against deleterious radiation effect (ICRP, 2009). The ICRP reference animals and plants (RAPs) are conceptual and numerical representations for estimation of radiation effect in apparent contaminated environment. ICRP publication 108[1] defines a list of specific animals and plants to be used for environment protection assessments. However, this list is observed to be more suitable to temperate countries, may be due to large set of studies in this regard are from developed countries from temperate climate. In this context certain gaps are observed in the ICRP-RAPs list, while representing Indian biogeographic condition. This study aimed to verify the applicability of ICRP's list of reference animals and plants to perform environmental assessments in India, with a focus on the flora and fauna surrounding the major nuclear facilities of India based on literature survey. Based on the distribution pattern of terrestrial organisms the world landmass is divided into 8 biogeographic realms [Figure 1]. India falls in Indomalayan realm, which has different floral and faunal distribution from Nearctic (North America) and Palearctic (Eurasia) region. A preliminary literature survey focusing the region surrounding Indian nuclear establishments was done. Indian nuclear establishments are situated in tropical climatic region mostly in coastal belts with few exceptions. The selected organisms were also cross checked for its distribution in Indomalayan region. The comparison of ICRP RAPs list with biota from the Indian region is presented on [Table 1]. It can be observed that from the 12 categories defined by ICRP, 5 are not adequately representing local biota around Indian Nuclear establishments. Accordingly flora and fauna are suggested as reference biota for Indian region in the remark section. Similarly the ICRP list doesn't include the reptiles, of which Wall lizard, turtle, snake can serve as representative biota. The ICRP reference list, although a noble step towards radiation protection of non-human biota, it doesn't truly represent the biota of worldwide nuclear sites. Hence a secondary list of reference biota need to be prepared based on the 8 biogeographic realms referring to the local nuclear sites.
Figure 1: Major biogeographic realms of world

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Table 1: Comparative status of International Commission on Radiological Protection reference biota in relation to their distribution in surroundings of Indian nuclear sites

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Keywords: Biogeography, fauna, flora, radiation protection

  References Top

  1. ICRP. Environmental Protection: The Concept and Use of Reference Animals and Plants Publication 108. Annals ICRP; 2008. p. 1-242.
  2. Rochedo ER, Wasserman MA, Barata GC. Radioecological risk assessment in tropical climate countries. Radioprotection 2011;46:785-8.

  Abstract - 16382: Assessment of radiation dose to aquatic non-human biota around Kaiga site using ERICA Top

Sanyam Jain, R. M. Joshi, J. P. James, M. S. Vishnu, I. V. Saradhi1, A Vinod Kumar1

Environmental Survey Laboratory, ESS, EMAD, HS and EG, BARC, Kaiga, Karnataka, 2EMAD, HS and EG, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

In this study, the radiation dose to non-human biota due to naturally occurring radionuclides (238U and 232Th) and anthropogenic radionuclide (137Cs) in freshwater sediments of Kadra reservoir around Kaiga site, is estimated using ERICA 2.0 (Environmental Risk from Ionising Contaminants: Assessment and Management) assessment tool (www.erica.tool.com). ERICA Tool is a dosimetric model that enables calculations of internal and external absorbed dose-rates to non-human biota covering a wide range of body masses and habitats for all radionuclides of interest. During the present work, radiological dose to benthic fish, Insect larvae, vascular plant, zooplankton, phytoplankton, mollusc-bivalve, amphibian, crustacean, and reptiles of freshwater ecosystem is estimated. Tier-2 in ERICA 2.0 was used to calculate the dose-rates to the organisms selected in this work. Inputs to the dose assessment model include radionuclide concentration, distribution coefficients (Kd), concentration ratios (CR), and occupancy factors. The activity concentration of the radionuclides, 137Cs, 238U and 232Th, in the reference organisms and water was calculated by multiplying the activity concentration in bottom sediments of Kadra reservoir by CR and Kd values. Wherever, site-specific values were not available, default values were used for the inputs. The modelled activity concentrations of 137Cs, 238U, and 232Th in reference organisms of the freshwater environment in the assessment are shown in [Table 1]. It was observed that the maximum activity concentration of 137Cs was found in reptiles followed by fishes. Whereas, in case of 232Th, vascular plant shows the maximum activity, and a comparatively lower concentration of 238U was observed in all the organisms. The corresponding total dose-rates (internal + external) to reservoir biota from 137Cs, 238U and 232Th were below the ERICA screening level of 10 μGy.h-1, as shown in [Figure 1]. These low dose-rates suggest that a more detailed dose assessment is not warranted. The maximum dose-rate was recorded for vascular plants (0.117 μGy.h-1) and minimum of 0.0003 μGy.h-1 was recorded for amphibians. This is mainly because of their habitat, occupancy factors, concentration factors and corresponding dose conversion coefficients.

The expected value of risk quotient,
Figure 1: Total dose-rates calculated using ERICA 2.0

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Table 1: Activity concentration in the reference organisms calculated using Environmental Risk from Ionising Contaminants: Assessment and Management 2.0

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investigated cases were <1, and therefore, the risk to aquatic non-human biota can be considered negligible. The estimated dose-rates are only a minuscule percentage of DOE limits (DOE-STD-1153).

Keywords: Biota, dose, ERICA, Kaiga, nonhuman

  References Top

  1. ERICA. Assessment Tool 2.0, 2021. Available from: www.erica.tool.com.
  2. DOE-STD-1153. US DOE, Washington; 2019.

  Abstract - 16573: Concentration ratio of 90Sr and dose to earthworm from tropical region Top

Vandana Pulhani, Manish Mishra, Moushumi D. Chaudhury, A. Vinod Kumar

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

E-mail: [email protected]

Non-human biota dose assessment is essential to verify that the environment is being, and will be protected, from the effect of ionizing radiation arising from existing, planned and emergency exposure situations. During operational phase, radiation dose to biota and assurance of environmental protection by planned, existing exposure situations is assessed by equilibrium transfer parameters. The transfer of radionuclides to wildlife (non-human biota) is normally quantified using an equilibrium concentration ratio (CRwo-media). Concentration ratio for a radionuclide 'i' in terrestrial biota is denoted by (CRi)

CRwo-media values for terrestrial biota have mostly been measured for low radionuclide activity, nearly at background level concentrations in soil.[1] However, for accidental situations, quantification is needed for terrestrial organisms in highly contaminated soils. The purpose of the present study was to investigate the effect of radionuclide activity concentration in soil, on the transfer of the long-lived fission product environmental contaminant 90Sr and 137Cs to a functionally important wildlife group earthworm (annelids - Eisenia fetida). Earthworm species have a high potential for exposure to radionuclides, due to their burrowing habit and ingestion of soil from their surroundings, which serves as their primary food. Due to their characteristics, of soil intake there is good probability of retention and exchange of radioactivity in its body tissue. The paper describes the study on whole body tissue concentration of strontium in earthworm and concentration ratio with respect to activity in soil and dose imparted. An experimentally spiked, controlled, contaminated area was selected for the study. Earthworms and soil in their habitat were collected. The earthworms were washed free of externally attached soil in running water using an ultrasonic bath and tapped dry with filter paper. Each earthworm was dissected and gut soil collected separately by water jet washing. The samples were suitably processed, counted by gamma spectrometry in suitable geometry and subjected to radiochemical separation for 90Sr analysis. The activity concentration of 90Sr and 137Cs[2] in earthworm tissue, gut soil material and soil is given in [Table 1]. CRwo-media values for 90Sr obtained are on the higher side of the range reported by D K Keoum et al., 2013. Environmental Risks from Ionizing Contaminants: assessment and management (ERICA) tool was used to assess the radiological risk. [Table 2] and [Table 3] give the Dose imparted to earthworm and risk quotients from 90Sr, 137Cs and their progeny as per ERICA Tool with an uncertainty factor of 5 (99% confidence level). The most important pathway of exposure to earthworm was ingested soil in the gut. The total dose rate per organism is well within the screening dose rate level of 40 μGy/h.[3] It is appropriate to assess the dose to biota in a contaminated area rather than lower contamination area to avoid uncertainties in estimation of the dose.
Table 1: Activity in soil and earthworm

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Table 2: Dose rate assessment on earthworm

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Table 3: Risk quotients to earthworm as per Environmental Risks from Ionizing Contaminants: Assessment and management tool (@99% confidence level)

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Keywords: Biota, ERICA, transfer factor

  References Top

  1. Keum DK, et al. J Environ Radioact 2013;126:427-33.
  2. Pulhani V, et al. 4th International Conferences on Radioecology & Environmental Radioactivity; 2017. p. 672-3.
  3. ERICA; 2006. Available from: www.erica-project.org.


  [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], [Figure 16], [Figure 17], [Figure 18], [Figure 19], [Figure 20], [Figure 21], [Figure 22], [Figure 23], [Figure 24], [Figure 25], [Figure 26], [Figure 27], [Figure 28], [Figure 29], [Figure 30], [Figure 31], [Figure 32], [Figure 33], [Figure 34], [Figure 35], [Figure 36], [Figure 37], [Figure 38], [Figure 39]

  [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], [Table 15], [Table 16], [Table 17], [Table 18], [Table 19], [Table 20]


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