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ABSTRACT
Year : 2023  |  Volume : 46  |  Issue : 5  |  Page : 120-162  

Theme 3. Radiation safety and protection in medical and industrial sectors


Date of Web Publication07-Feb-2023

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Source of Support: None, Conflict of Interest: None


DOI: 10.4103/0972-0464.368742

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How to cite this article:
. Theme 3. Radiation safety and protection in medical and industrial sectors. Radiat Prot Environ 2023;46, Suppl S1:120-62

How to cite this URL:
. Theme 3. Radiation safety and protection in medical and industrial sectors. Radiat Prot Environ [serial online] 2023 [cited 2023 Jun 2];46, Suppl S1:120-62. Available from: https://www.rpe.org.in/text.asp?2023/46/5/120/368742




  Abstract - 31105: Radiation protection in fabrication, clinical deployment and safe management of I-125 and P-32 brachytherapy sources Top


Sanjay Kumar Saxena, Yogendra Kumar

Radiopharmaceuticals Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

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

Introduction: Brachytherapy is a modality in which highly conformal radiation doses are delivered to malignant lesions by placing sealed radiation sources in close proximity of diseased site. The main goal is to achieve high therapeutic gain by imparting tumorocidal doses to the desired site and minimising radiation dose to nearby healthy tissues and critical organs. In 'Low Dose Brachytherapy', radiation dose is delivered at lower rate (Gy/h) over a prolonged time. Low energy x-rays/gamma photons emanating radionuclides such as I-125, Pd-103, Cs-131 in sealed form are used either as temporary implant or permanent implant for treatment of eye and prostate cancer respectively. In a similar way, high energy beta emitters such as P-32, Y-90 etc. are used as sealed patches for the treatment of skin cancer and keloids. The paper briefly describes radiological protection strategies adopted during fabrication, quality assurance, transportation, clinical use and management of disused sources in India.

Radiation Safety in Preparation of Sources: Fabrication of I-125 seeds for eye/prostate cancer was carried under optimised conditions by adsorption of 18 -150 MBq of I-125 on palladised silver wires and their Nd: YAG laser encapsulation in 50 micron thick titanium capsules.[1] Entire operation was carried out inside an air tight glove box with a 1.5-2 inch water gauge negative pressure. Inside the glove box silver impregnated charcoal filter traps connected to NaOH traps were installed to retain gaseous I-125 in case of any release due to unforeseen circumstances. Sources were subsequently sealed in titanium capsules with Nd: YAG laser installed inside a well ventilated fume hood complying with safety protocol applicable to laser operations. P-32 sources in customised shapes were prepared by incorporation of up to 74 MBq/cm2 of P-32 in adsorbent strips and their entrenchment within thin polymer sheets for treatment of skin cancer and keloids. Continuous air monitoring for air borne radioactivity was carried out at all stages.

Quality Assurance and Classification Designation: I-125 sources were assayed and subjected to classification designation tests w.r.t. temperature, pressure and impact tests. Hot water leakage and removable surface contamination tests were also carried out in accordance with the procedures promulgated in 'Testing and Classification of Sealed Radioactive Sources' (i.e., AERB/SS-3; Rev-1, 2001). P-32 patches were subjected to surface contamination and leakage tests. In all cases, released radioactivity was measured with calibrated instruments either with NaI (Tl) counter or liquid scintillation counter.

Radiation Safety in Clinical Deployment of Sources: I-125 and P-32 sources were deployed in seven cancer treatment hospitals of India. Prior to supply of these sources, it was ensured that all these hospitals have obtained valid seed procurement approvals from AERB, India in accordance with provisions of 'Safety Code for Brachytherapy Sources, Equipments and Installations' (i.e., AERB/SC/MED-3; 1998).

Transport and Management of Disused Sources: Movement of brachytherapy sources was carried out in coordination with BRIT as described in AERB Safety Code AERB/NRF-TS/SC-1(Rev-1), 2015. Ocular sources after use were received back from user. Such sources were decayed for ten half lives and were later disposed off as solid radioactive waste.

Results: Both I-125 and P-32 brachytherapy sources could be successfully fabricated and deployed for benefitting eye, prostate and skin cancer patients. The airborne radioactivity during fabrication of sources was well below 0.1 DAC. ALARA principle was followed and no overexposure was reported at any stage. In all the cases, removable surface contamination and release of radioactivity in leakage test was less than maximum permissible level of 185 Bq. I-125 seeds were classified by AERB as Class-43211 sources for brachytherapy applications. Collaborating oncology centres strictly adhered to regulatory requirements during clinical use and returning of unused/disused sources back to our Division through BRIT.

Conclusion: Radiation safety could be ensured in preparation, clinical use and safe management of I-125 and P-32 brachytherapy sources. The experience gained over the last few years is highly encouraging to offer safer radiation sources for societal benefits.

Keywords: Airborne release, brachytherapy, I-125, P-32, sealed therapeutic sources

Acknowledgement

Authors are thankful to Dr. Tapas Das, Head, RPhD and Dr. S.Kannan, Director, RC&IG, BARC for their encouragement and support.


  Reference Top


  1. Saxena SK, et al. J Radioanal Nucl Chem 2014;302:1237-43.



  Abstract - 31192: Estimation of radiological source term for an isotope production facility Top


Tanmay Sarkar1,2, Jayant Krishan1,2, S. Anand1,2, Kapil Deo Singh1, M. S. Kulkarni1,2

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

E-mail: [email protected]

Introduction: Regulatory clearance for setting up an isotope production facility requires estimation of public dose at the site boundary during normal operation. To estimate public dose source term (radionuclide, release rate, and physical form) needs to be determined either by theoretical method or measurements. In the present study, an attempt is made to theoretically estimate source term for an isotope production facility (using irradiated fission target) which will be used for carrying out radiological impact analysis.

Process Description: It is proposed that the gaseous wastes generated during the chemical process (Volatiles, noble gases and particulates) in this facility will pass through dissolver, condenser, intermediate storage vessel, hydrogen converter, water condenser, and will be stored finally in gas collection tank for six weeks. After 6-weeks of delay, the gaseous waste will be passed through charcoal filter and charcoal column for iodine trapping as well as decaying noble gases for 10 weeks. Finally the gaseous wastes are released to environment via a series of high efficiency particulate air (HEPA) filter and charcoal filter in cell ventilation room. Short-lived radionuclides in gaseous inventory will be decayed out (>99%) after 16-week in the gas storage tank and charcoal column. The movement of airborne radionuclides follow the path as mentioned in [Figure 1].
Figure 1: Schematic diagram for movement of gaseous wastes

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Table 1: Source term (Bq/y) from facility for different scenario

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Methodology: In the present study, ORIGEN2 code is used for calculating the build-up and decay of various radioactive materials in the irradiated target. Airborne release fraction, retention factor in the various filtration systems for filtering particulates and volatiles, and delay of radionuclides are also incorporated[1] in calculation. During normal operation there may be fractional releases due to valve operation, compartmental transfer of inventory etc. Three major release scenarios are postulated in this study: (1) 100% storage of airborne radionuclides in the waste collection tank, (2) 50% of airborne radionuclides are stored in the waste collection tank and while remaining 50% is released through the ventilation system after 8-days of delay, (3) Similar to the second one but 50% of airborne activity is released immediately (without 8-days delay) through the ventilation and filtration systems.

Results and Discussion: Source terms for the three scenarios are simulated using ORIGEN2 and annual releases are presented in [Table 1]. Total activity release in scenario1, 2, and 3 are 4.03E+11 Bq/y, 8.56E+14 Bq/y, and 4.31E+15 Bq/y respectively. The least source term is for scenario 1, due to decay of radionuclide having short half-life and hence corresponding public doses will be minimum. Stack monitoring of radionuclide releases during normal operation of the facility will help to verify the estimated source term. This work presents a methodology to estimate the source term by including all the filtration and decay processes during normal operation of an isotope production facility.

Keywords: Isotope, ORIGEN2, source term


  Reference Top


  1. Croff AG. ORIGEN2: A versatile computer code for calculating the nuclide compositions and characteristics of nuclear materials. Nucl Technol 1983;62:335-52.



  Abstract - 31513: Comparison of activity of fission 99Mo from FLUKA MC Simulation and ORIGEN2 code Top


Riya Dey1,2, T. Sarkar1,2, K. D. Singh1, M. S. Kulkarni1,2, S. Anand1,2

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

E-mail: [email protected]

99Mo (t1/2 = 66 h) - 99mTc (t1/2 = 6.01 h) generator is widely used in radiopharmaceutical industry. The decay of 99Mo leads to the generation of 99mTc which is referred as the workhorse isotope in nuclear medicine utilised for diagnostic imaging. This type of generator needs high specific activity of 99Mo which can be produced by the fission of 235U. In such cases, the estimation of 99Mo activity, produced by fission inside the reactor, is essential to acquire a baseline idea of activity of 99mTc that can be obtained during the usage of the generator. The activity build-up of 99Mo due to neutron induced fission depends on the duration of irradiation, neutron flux experienced by the target, chemical composition of the target. Usually, the ORIGEN2 code[1] is used for the estimation of radioactivity inventory which assumes spectrum-weighted single-group fission neutron yields per neutron-induced fission for either thermal or fast reactor. In reality, the neutron flux is a function of its energy and it changes depending on the location inside the reactor. Hence, instead of using one-group neutron interaction cross-section, energy dependent interactions need to be considered. To overcome these limitations of ORIGEN2 code, Monte Carlo (MC) based FLUKA simulations can be adopted for the realistic estimation of radioactivity. To compare the radioactivity estimated from ORIGEN2 and FLUKA code, first, the ORIGEN2 code was run in POWER mode to obtain the flux value corresponding to the burnup of the fuel in the research reactor. The flux thus obtained was then incorporated in the FLUX mode of the ORIGEN2 code to get activity build-up. This flux is different from the original flux at the activation location inside the reactor, since ORIGEN2 generates a flux corresponding to the one-group neutron profile. In FLUKA simulation, the actual flux spectra can be incorporated and hence energy dependent interaction cross-sections can be utilised instead of using a single group collapsed cross-section value. To achieve this, theoretical thermal neutron spectrum was analysed for the present research reactor which uses natural U as fuel heavy water as moderator and coolant. In the thermal region, the flux distribution of neutrons follows Maxwell-Boltzmann distribution given by



where, n0 is total thermal neutron density (neutrons/cc), v is neutron velocity (cm/s), E is neutron energy (eV). At a reactor temperature of 637K, the value of kT is (1/18.22) eV. Normalizing eq. (1) by energy-integrated measured thermal flux of 1.80E+14 cm-2 s-1 at the activation location, the final thermal flux spectrum (neutrons cm-2 s-1 eV-1) follows the following relation



A FORTRAN based code (source.f) was prepared which incorporated the above flux distribution and this code was compiled along with FLUKA. [Figure 1] depicts this thermal flux distribution. The target considered here is a UAl3 plate with an enrichment of 19.75%, the mass ratio of U and Al is 7:15. An irradiation of 14 days was considered. 99Mo will be produced as a result of fission of 235U present in the target plate as well as due to decay of other fission products (e.g., decay of 99Yr etc.). The irradiation period was introduced using IRRPROFI card, the RADDECAY card enables decay of radionuclides, and the cooling period is set as 0 in DCYTIMES card to get the output at the end of irradiation only. RESNUCLE card prints out the activity of radionuclides produced after irradiation. From the FLUKA simulation, the activity of 99Mo was found to be 1.42E+04 Ci/kg, whereas the value obtained from ORIGEN2 was 1.94E+04 Ci/kg. The discrepancy between these two values appears because ORIGEN2 uses a one-group collapsed cross-section, whereas in FLUKA, the realistic spectrum can be incorporated. Apart from this, the reactor temperature also decides the shape of the spectra as well as the most probable velocity of neutrons inside the reactor. FLUKA simulation provides an advantage over ORIGEN2 code by incorporating this information in the code which cannot be achieved if pre-defined ORIGEN2 spectrum library is used.
Figure 1: Theoretical thermal neutron spectra normalized with measured total thermal flux

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Keywords: Fission Molly, FLUKA, Monte Carlo, ORIGEN2, thermal neutron


  Reference Top


  1. Croff AG. Nucl Technol 1983;62:335-52.



  Abstract - 32119: Calculation of relative biological effectiveness of therapeutic proton beams: TOPAS Monte Carlo study Top


Arghya Chattaraj1,2, T. Palani Selvam1,2

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

E-mail: [email protected]

Introduction: Most clinical proton therapy facilities use a constant value of Relative Biological Effectiveness (RBE) as 1.1. However, recent studies have reported some variations in the RBE value along the depth.[1] The present study is aimed at estimating RBE of therapeutic protons along the on-axis depth in water using TOPAS Monte Carlo code.

Materials and Methods: The RBE of protons are calculated based on Microdosimetric Kinetic Model (MKM) and Local Effect Model (LEM). For MKM, microdosimetric distributions at 1 μm site size are calculated. The MKM-based RBE at 10% survival level can be calculated using the following equations:[2]





Where y and f(y) are lineal energy and frequency probability density of y within y and y+dy, respectively. βt = βx =0.05 Gy-2, αx (200 kVp X-rays) = 0.19 Gy-1, rd = 0.42 /μm, α0 = 0.13 Gy-1, ρ = 1 g/cm3 and αo = 150 keV/μm. Suffix 'x' and 't' are for X-rays and test radiations, respectively. The LEM-based RBE at 10% survival level is calculated for V79 Chinese Hamster Cells using the equation:



Where, D=prescribed proton dose = is a function of ; dose-weighted Linear Energy Transfer,



Results and Discussions: RBEMKM varies between 0.99 – 1.20 and 1.00 – 1.02, respectively, for 250 and 62 MeV protons [Table 1]. For a given proton energy, microdosimetric distributions shifted towards higher y values as depth increases [Figure 1].

RBELEM varies between 1.02 – 1.50 and 1.05 – 1.51, respectively, for 250 and 62 MeV protons [Figure 2]a. The values of LETd vary between 0.5 – 13 keV/μm and 1 – 13.5 keV/μm, respectively, for 250 and 62 MeV protons [Figure 2]b. The RBELEM or LETd increases gradually with depth in water and reaches maximum at the distal end. Although, there are several RBE calculation models are available in TOPAS, MKM and LEM are widely used for commercial proton treatment planning system. Considering the different biological end points of MKM and LEM, RBEMKM and RBELEM show reasonably good agreement (within 20%). RBEMKM, RBELEM and LETd are sensitive to depths and having maximum values at the distal ends.
Figure 1: TOPAS-calculated on-axis microdosimetric distributions in water at 1 μm site size for 62 and 250 MeV protons at entrance and Bragg-peak positions

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Figure 2: TOPAS-calculated on-axis (a) RBELEM and (b) LETd of 62 and 250 MeV protons as a function of depth in water

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Table 1: TOPAS-calculated Microdosimetric Kinetic Model-based on-axis relative biological effectiveness values at different depths in water

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Keywords: Microdosimetry, Monte Carlo, proton, RBE


  References Top


  1. Polster L, et al. Phys Med Biol 2015;60:5053-70.
  2. Newpower M, et al. Int J Radiat Oncol Biol Phys 2019;104:316-24.



  Abstract - 32164: Radioactive decay properties of superheavy element Livermorium Top


P. S. Damodara Gupta1,2, N. Sowmya1, H. C. manjunatha1, T. Ganesh2

1Department of Physics, Government College for Women, Kolar, Karnataka, 2Department of Physics, Rajah Serfoji Government College, Affiliated to Bharathidasan University - Tiruchirappalli, Thanjavur, Tamil Nadu, India

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

The radioactivity may be used to distinguish between rays or particles released by a very unstable system and radiation emitted spontaneously by a system with nuclear and atomic degrees of freedom near to equilibrium. An unstable nucleus will spontaneously dissolve, or decay, towards a more stable structure, but only in a few specific ways by releasing certain particles or kinds of electromagnetic radiation. Radioactive decay is a feature of numerous naturally occurring elements as well as isotopes of the elements created artificially. The rate of decay of a radioactive element is stated in terms of its half-life, which is the amount of time necessary for one-half of any given quantity of the isotope to decay. The characteristic half-lives factor are used in order to identify the radioactive decay of the superheavy element Livermorium. Coulomb and proximity potential with recent proximity function are involved in the evaluation of half-lives. The possible decay modes such as spontaneous fission, alpha-decay, cluster-decay and heavy particle radioactivity half-lives are compared. The Coulomb and proximity potential is evaluated using the theory explained in the literature.[1],[2],[3] The total potential evaluated in case of 250Lr and the plot of the same with respect to separation distance is as seen in [Figure 1]. The penetration probability is evaluated using turning points Ra and Rb. Furthermore, we have evaluated spontaneous fission half-lives, cluster and heavy particle radioactivity half-lives as explained in the reference.[1],[4] The evaluated half-lives using different decay modes were compared with each other. For an instance, the [Figure 2] shows A comparison of alpha-decay, heavy particle radioactivity (86Kr), cluster radioactivity (40Ca) and spontaneous fission half-lives as function of mass number of parent nuclei for the superheavy element Lr. From this comparison it is clear that the nuclei 257-260Lr shows smaller alpha-decay half-lives when compared to other studied decay modes. However, spontaneous fission half-lives dominate in case of below 256Lr and above 261Lr. Hence, the detail investigation of different decay modes, it is clear that the nuclei 257-260Lr undergoes alpha decay. Furthermore, below 256Lr and above 261Lr spontaneous fission half-lives dominates. Hence, this study finds an important role in the identification of new isotopes in the heavy nuclei Livermorium.
Figure 1: A plot of total potential as function of separation distance in case of 250Lr during alpha-decay

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Figure 2: A comparison of alpha-decay, heavy particle radioactivity (86Kr), cluster radioactivity (40Ca) and spontaneous fission half-lives as function of mass number of parent nuclei for the superheavy element Lr

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Keywords: Alpha-decay, beta-decay, half-lives, isotopes, stability


  References Top


  1. Manjunatha HC, Sridhar KN, Sowmya N. Phys Rev C 2018;98:024308.
  2. Nagaraja AM, Manjunatha HC, et al. Eur Phys J Plus 2020;135:814.
  3. Sridhar GR, Manjunatha HC, et al. Eur Phys J Plus 2020;135:291.
  4. Nagaraja AM, Manjunatha HC, et al. Nucl Phys A 2021;1015:122306.



  Abstract - 32200: Study for selection of aggregate and cement type to minimize induced 54Mn activity in concrete shield of particle accelerator facilities Top


A. A. Shanbhag1, Sabyasachi Paul1, S. C. Sharma2, M. S. Kulkarni1,3

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

E-mail: [email protected]

Particle accelerators now find use in basic research as well as in medical fields. The data obtained during the operation and decommissioning of such facilities worldwide proves useful in choosing materials for the facility construction to bring down its collective radiation exposure. Being a structural material and also capable in attenuation of gammas and neutrons, concrete is a preferred radiation shielding material. Past research shows that both short lived radionuclides like 56Mn, 24Na etc. and long lived radionuclides like 3H, 60Co, 152Eu, 154Eu etc. are generated in the concrete shield of accelerators due to neutron activation. The presence of 54Mn (T1/2: 312 days), that is neither very short lived nor very long lived, is commonly seen in accelerator concrete shields. In this study, we have measured the 54Mn activity in different concrete samples when exposed to accelerator produced neutrons generated from the bombarding of a thick Be target with a 200 nA beam of 20 MeV protons at the 6 M port of the BARC TIFR Pelletron Linac Facility. The average neutron flux incident on concrete samples was 3.3 X 109 n/cm2/s and duration of exposure was 40 minutes. The measured activity values of 54Mn (normalized per unit mass of concrete and per unit incident proton on the Be target) are given in the [Table 1] below. Concrete sample code GJ-01-G indicates that the aggregate used is GJ-01 and the cement used is grey portland cement and all samples are coded accordingly.

The normalized 54Mn activity values in [Table 1] for all concrete samples prepared using white Portland cement are given in italics and that for the concrete sample giving the lowest value amongst the 14 samples studied in this work is given in bold and italic. From [Table 1], the following observations can be made:

  • Concrete samples prepared using white Portland cement generated lower amount of 54Mn for all the seven aggregates studied in this work.
  • Concrete sample prepared using marble aggregate and white Portland cement generated the lowest activity of 54Mn amongst the fourteen concrete samples studied in this work.


Fe contributes towards the generation of 54Mn through the 54Fe(n,p)54Mn nuclear reaction. It has been reported that the production of white portland cement makes use of raw materials with low content of iron, manganese, chromium and titanium to ascertain white color.[1] It has also been reported that the average abundance of iron in basalt, granite (with high Ca content), granite (with low Ca content) and carbonate rocks are 8.65 wt%, 2.96 wt%, 1.42 wt% and 0.38 wt% respectively and mentioned that the data for basaltic rocks include the values for all rocks of basaltic type namely, gabbros, dolerites and basalts.[2] Gneiss, being chemically similar to granite would have iron content similar to granite. Hence marble (metamorphosed limestone) will generally have iron content lower than granite, basalt, gabbro and gneiss. This explains (i) the lower generation of 54Mn in concrete samples prepared using white portland cement as compared to concrete samples using grey portland cement for all the seven rock aggregates studied in this work and (ii) the lowest generation of 54Mn in the concrete sample prepared using marble aggregate and white portland cement amongst all the fourteen concrete samples studied in this work. The study indicates that the use of marble aggregate and white Portland cement would help in minimizing the generation of 54Mn in accelerator shield concrete.
Table 1: Normalized 54Mn activity in concrete samples

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Keywords: 54Mn, accelerators, cement, concrete shield, neutron activation


  References Top


  1. Moresova K, et al. Ceram Silikaty 2001;45:158.
  2. Turekian KK, et al. Geol Soc Am Bull 1961;72:192.



  Abstract - 32212: Experimental determination of broad beam attenuation coefficient of ordinary concrete, iron and lead for bremsstrahlung X-ray from 450 MeV electrons Top


T. K. Sahu, P. K. Sahani1, M. K. Nayak, R. Sharma, A. Mondal1, G. Haridas, M. S. Kulkarni

Health Physics Division, BARC, Mumbai, Maharashtra, 1Indus operation Division, RRCAT, Indore, Madhya Pradesh, India

E-mail: [email protected]

High energy electron accelerators are used to generate the synchrotron radiation for different scientific research. Bremsstrahlung X-ray is generated when high energy electrons beam strike any target material. Bremsstrahlung X-ray is the major radiation hazards in such high energy electron accelerator. Photo-neutron also contributes to the radiation environment but its contribution is about 3 orders less than that of bremsstrahlung X-ray. For experimentally measuring the attenuation coefficient of common materials/absorbers like ordinary concrete, iron and lead in the bremsstrahlung X-ray field generated from 450 MeV electrons, an experiment was conducted at the booster synchrotron facility, Singh et al.[1] at RRCAT, Indore, India. The work presented in the paper describes the experiment carried out and the results obtained.
Table 1: Attenuation coefficient and tenth value layer value

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The experiment is conducted to study the attenuation of high energy bremsstrahlung X-ray in ordinary concrete, iron and lead. Alignment of the absorbers w.r.t. the beam axis was carried out using fluorescent screen and radio chromic (EBT) films prior to the experiments. Electrons beam of energy 450 MeV from booster synchrotron was allowed to hit a lead (Pb) target of one CSDA range (25 mm), https://physics.nist.gov to generate the bremsstrahlung X-ray. For the depth dose measurement in the bremsstrahlung X-ray field, the absorbers with CaSO4:Dy TLD disc were set up at a distance of 1 m from the lead target. From the depth dose curve, the attenuation coefficients are deduced. TLD reader (Make: Intech, India, Model No. I-1602 DC) is used to read out the absorbed dose in CaSO4:Dy TLDs. The TLD discs used for the experiment were having sensitivity within ± 5%. The schematic view of experimental setup is given in the [Figure 1]. Based on the absorbed dose data from TLDs, attenuation curves are plotted for ordinary concrete, iron and lead. The attenuation curves of absorbers are given in the [Figure 2]. [Table 1] gives the attenuation coefficient obtained and corresponding Tenth Value Layer (TVL). The data is very useful for the shield evaluation of high energy electron accelerators.
Figure 1:. Schematic view of experimental setup

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Figure 2: Attenuation curves of the absorbers

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Keywords: Attenuation coefficient, depth dose, high energy bremsstrahlung X-ray


  Reference Top


  1. Singh et al. Proc. Conf. Indian Particle Accelerator Conference; 2011. Available from: https://physics.nist.gov.



  Abstract - 32226: Estimation of dose to cargo for indigenous portal type cargo scanner Top


Vitisha Suman, K. Biju, M. S. Kulkarni

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

E-mail: [email protected]

Pulsed X-ray LINACs are widely deployed as scanners for the interception of prohibited or restricted materials pertaining to national security as well as local regulations. The accelerator fires a collimated x-ray beam through the container and transmitted beam are collected by portal detectors. A computer program converts the collected detector data into an image. Many commercial scanners employ pulsed LINACs with interlaced dual-energy (4/6 MV) beams to allow for material discrimination. The selection of imaging and scanning systems that would allow for effective inspection of all potential threat items would require a judicious trade-off between high performance, high throughput, small exclusion zone, simple logistics with low acquisition and operational costs.[1] Indian cargo scanner Portal version (ICS-P) is indigenously developed by BARC and uses a 6 MV Linac with output of 1 Gy/min at 1 m. A fan beam of 60° and 4mm wide is produced by using primary and secondary collimators for scanning the cargo containers. [Figure 1] shows the typical fan beam irradiation geometry. American National Standard parameters[2] are to be established during the commercial operation of the system for determination of imaging performance of X-ray & γ-ray systems for cargo and vehicle security scanning. This paper presents the simulations carried out to establish the ANSI parameter, 'Dose to Cargo' and the validation by measurement for the ICS-P. This Dose to cargo parameter is important from the radiological protection point of view as it is the average dose delivered to the product inside any cargo which could be radio sensitive. As per the ANSI specifications the dose is measured in an empty cargo at the location of (h/2) and (h/4) height on the scan area centreline as shown in [Figure 1], where h is scan height of 4.2 m. The Monte Carlo simulations using FLUKA were carried out simulating the bremsstrahlung spectra from 6 MeV electron beam stopped on a tungsten target. The forward beam was then collimated to a fan beam of height 4.2 m, 4mm wide by modelling the primary and secondary collimators, as shown in [Figure 2]. The absorbed dose rates at the h/2 and h/4 heights were obtained. Measurements were also made to validate the h/2 value using thermo luminescent dosimeters (TLD).[4] [Table 1] presents the simulated and measured values of dose to cargo. The calculated value for 'Dose to Cargo' was found to be 22.1 μSv/scan and the measured value was 25 μSv/scan. The theoretical value obtained is in agreement with the measured value with in 12%. The important ANSI parameter for the indigenously developed Indian Cargo Scanner was estimated using Monte Carlo method for the facility's qualification as to be fit for commercial operations. The estimated value was in good agreement with the measured value.
Figure 1: The experimental setup showing (h/2) and (h/4) levels

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Figure 2: Cross-sectional view of simulated geometry

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Table 1: Dose to cargo measured and theoretically estimated in the stationary and dynamic modes of operation

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Keywords: Cargo scanner, dose to cargo, dual-energy


  References Top


  1. Joseph B. Phys Procedia 2017;90:242-55.
  2. ANSI N42.46-2008; 2008.
  3. Huet C, et al. Radiat Meas 2014;71:364-8.
  4. Kumar R, et al. Internal Report, RSSD/RHC/RKG/ 2018/479; 2018.



  Abstract - 32227: Experimental determination of radiation source term during the operation of 150TW laser plasma accelerator facility Top


Dilip Kumar Gupta, Aritra Mandal1, G. Haridas, M. S. Kulkarni, Tirtha Mandal2, Vipul Arora2, Anand Moorti2

Health Physics Division, BARC, Mumbai, Maharashtra, 1Indus Operation Division, RRCAT, 2Laser Plasma Division, RRCAT, Indore, Madhya Pradesh, India

E-mail: [email protected]

At RRCAT a 150 TW Ti: Sapphier Laser system is operational and being used for carrying out research in ultra-short, ultra-high intensity laser plasma interaction. Ti: Sapphier Laser (800 nm, 25 fs duration) is focused on a solid/gas target placed inside an evacuated chamber (Plasma Chamber), along with several diagnostics, which are housed inside a shielded vault. Interaction of the intense ultra-short laser with the target material leads to plasma generation and subsequent charged particle (electron/proton/ion) acceleration. Electrons could be accelerated to high energies (MeV to hundreds of MeV). These energetic electrons generate pulsed bremsstrahlung X-rays which is the main radiation hazard in such facility, Liang et al.[1] Such facilities are of unique type and very few around the world, due to which very less information is available about the radiation levels around the Chamber. In the present work, measurements were performed to find out the radiation source term by measuring the dose distribution inside the plasma camber when the intense laser beam is focused on a Cu target. Photon dose measurement was carried out using ion chamber based Direct Reading Dosimeters (DRD). Before the measurement the calibration check of the DRDs were carried out using Co-60 source. [Figure 1] shows the location of DRDs installed inside the plasma chamber in a circular arc behind the target at a distance of ~11 cm from the source i.e. laser focal point on the target. Total 25 shots of the laser pulse of beam of 25 fs duration and peak power of ~112 TW (focused at an angle of incidence of 300 and intensity > 4x1019 W/cm2) were fired on 17μm Cu target for studying the dose distribution. The dose distribution around the target with respect to the locations of DRDs is shown in [Figure 2]. From the measurements it is observed that:

  • The dose distribution around the target indicates that the maximum dose per shot is lying between the direction of laser propagation and target surface [towards DRD location 1 as in [Figure 1]].
  • Inside the chamber (inaccessible area) a maximum dose of 420 mSv for 25 shots was recorded at 11 cm from the target.
  • The source term obtained from the measurement is 203.28 μSv/shot at 1 meter from the Cu target for the given experimental conditions.
  • The photon and neutron dose outside the shielded areas obtained from radiation survey are found to be at background level.
  • The data will be helpful for assessing the radiation levels and implementing radiation safety measures in laser plasma based accelerator facilities.
Figure 1: Schematic diagram showing the DRD locations (1-18) inside plasma chamber

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Figure 2: Variation of dose around the target inside the Laser plasma chamber

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Keywords: Femto second, plasma, Ti: sapphire laser


  Reference Top


  1. Liang T, Bauer JM, Liu JC, Rokni SH. Bremsstrahlung dose yield for high-intensity short-pulse laser–solid experiments. Radiat Prot Dosimetry 2017;175:304-12.



  Abstract - 32228: Induced radioactivity in common structural material of proton accelerator in the range, 100-1000MeV Top


R. Sharma, P. K. Sahani1, G. Haridas, M. S. Kulkarni

Health Physics Division, Bhabha Atomic Research Centre, Mumbai, 1Indus Operation Division, Raja Ramanna Centre for Advanced Technology, Indore, Madhya Pradesh, India

E-mail: [email protected]

Activity induced in structural material and components of a proton accelerator is one of the major radiation hazard Thomas and Stevenson.[1] Induced activity in accelerator components is due to the interaction of primary proton beam with accelerator components and from the secondary and tertiary particle interactions. The amount of radioactivity and generated radioisotopes depend on primary beam energy and the target material. The amount of radioactivity at any given time will depend on the half-life of isotopes produced, accelerator operation time and cooling time. The estimation of radioactivity induced in accelerator structures and its auxiliary components are required to be evaluated for ensuring radiation safety during maintenance activities. The work reported in this paper presents the study of radioactivity induced in common structural materials like aluminium (Al), iron (Fe) and copper (Cu) due to bombardment of proton of different energies. Theoretical estimation using empirical method and FLUKA simulation are used to study the isotopes produced and its activity due to proton beam in the energy range, 100-1000MeV. The proton energy assumed are 100, 250, 500, 750 and 1000MeV. Activity has been estimated for various irradiation times (8 h, 40 h and 2000 h) and for a cooling time 0.1 h. The activity is estimated using the empirical relation



Where A is the activity generated in the target material in Bq/cc, N is the number of target nuclei/cc, φ is the incident flux of protons (protons/cm2-s), λ is the decay constant (s-1), t1 (h) is the irradiation time of proton with the target material and t2 (h) is the cooling period after the irradiation time and σ(barn) is the reaction cross section for producing the radioisotope. Cross-section data has been taken from published literature Th. Schiekel et al., Titarenko et al.[2],[3] In the empirical method, the radioisotopes produced from only the primary interaction of proton beam are considered to evaluate induced activity. The Monte Carlo code FLUKA is also used to estimate induced activity in these materials. A pencil beam of protons of varying energy 100 to 1000 MeV energy was allowed to incident on a cylindrical target of radius 0.564 cm and length 1 cm. RESNUCLEI and DCYSCORE scoring cards were used for activity calculation. The target was irradiated for 8 h, 40 h and 2000 h with proton beam (1 proton/sec) and the residual activity generated in the target was evaluated after a cooling time of 0.1 h. The dominant radionuclides obtained in Al, Fe and Cu with their half-lives are listed below: Aluminium: Na-24 (15 h); Iron: V-48 (15.9 d), Cr-49 (42.3 m) and Mn-54 (312.2 d); Copper: Co-58 (70.85 d), Cu-61 (3.4 h), Cu-62 (9.7 m) and Cu-64 (12.7 h). The induced activity obtained due to long term proton bombardment (2000 h) is presented in [Figure 1]. Results indicated that the activity from Fe and Cu is almost ten times higher than that of Al nearly for all the energies considered. The data helps in choosing materials in proton accelerator environment to reduce radiation hazard to workers. The activity obtained from empirical method is found be lower than the simulated data as the in empirical method only primary interaction cross- section is used whereas in simulation secondary process also contribute to the activity.
Figure 1: Activity obtained from empirical method and FLUKA simulation

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Keywords: FLUKA, induced activity, proton accelerator


  References Top


  1. Thomas RH, Stevenson GR. Radiological Safety Aspects of the Operation of Proton Accelerators. Vienna: IAEA-283; 1988.
  2. Schiekel TH, Sudbrock F, Herpers U, et al. Nucl Instrum Methods in Phys Res B 1996;114:91-119.
  3. Titarenko YE, Batyaev VF, Titarenko AY, Butko MA, Pavlov KV, Florya SN, et al. Phys Rev C 2008;78:034615.



  Abstract - 32229: Estimation of skyshine doses around a self-shielded 6 MV electron LINAC facility Top


P. Srinivasan, K. Biju, Vitisha S. Dagre, M. S. Kulkarni

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

E-mail: [email protected]

Electron linear accelerators have been developed by BARC for industrial, medical and security applications. The accelerators used for security applications are commonly deployed in an open top configuration without any ceiling wall owing to functional requirements for cargo scanning. Such LINACs are self-shielded to limit the leakage radiation level within permissible leakage levels, specified by AERB standards.[1] In addition, radiation doses from skyshine effects are important for radiation protection and occupancy in the vicinity of such LINAC installations. In a 6 MV self-shielded electron linear accelerator (LINAC) used for Indian Cargo Scanner (ICS), the conversion target is shielded locally using tungsten alloy collimator and steel. The LINAC produces a large fan beam of X-rays with a vertical beam cone angle of 60° using primary and secondary collimator, a horizontal beam width of 4 mm spanning 4.2 m of vertical dimensions at about 3 m from the source in the forward beam direction. The self-shielded LINAC is housed in an open-top concrete vault of 3.5 m height. The analytical method recommended by the National Council on Radiation Protection[2] Report No. 151 was used for the evaluation of skyshine photon dose equivalent rates. The geometry of the NCRP model is shown in [Figure 1]. Monte Carlo software monte carlo method was also used to estimate the dose rates around the target. Dry air of density 1.2905 x 10-3 g/cc and a concrete wall of density 2.3 g/cc with thickness of 70 cm were assumed along the floor and the 3.5 m tall walls. The distances from the LINAC to the walls were considered as 5 m on the sides, 2 m in the rear and 15 m in the front. Bremsstrahlung photons emitted from the LINAC below a height of 3.5 m will be shielded by the concrete walls such that the transmitted dose rates are attenuated by more than 2 orders of magnitude. But the photons emitted above 3.5 m height will be unshielded and scattered by the air above the open top LINAC leading to skyshine doses on the ground level outside the walls. The sky shine doses are negligible on contact with the outer sides of the wall and increases as we move away from the wall. After reaching a maximum at a particular distance from the wall, the skyshine doses decreases further due to distance. The average energy of skyshine and scattered radiation is found to be in the range 100–300 keV. A dose rate map of the LINAC is shown in [Figure 2]. A comparison of the skyshine doses estimated by NCRP-151 and the monte carlo method methods is shown in [Figure 3]. The NCRP-151 model overestimates the dose rates at close distances (<5m) and under predict at far-off distances between 10 – 20 m, when compared to monte carlo method. Similar trend is also reported in literature.[3] This information is important for radiation protection and shielding of such open-top electron LINACs and to establish appropriate boundary fence for protection of neighbouring occupancy areas.

Keywords: Bremsstrahlung, electron accelerator, shielding, skyshine

Acknowledgement

Authors are grateful to Shri. M. P. Kulkarni, APPD, BARC for providing the LINAC details.


  References Top


  1. Atomic Energy Regulatory Board. AERB SAFETY CODE NO. AERB/RF-MED/SC-1 (Rev. 1); 2011.
  2. NCRP. NCRP Report No. 151; 2005.
  3. McDermott PN. Photon skyshine from medical linear accelerators. J Appl Clin Med Phys 2020;21:108-14.
Figure 1: NCRP-151 model for skyshine dose estimation

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Figure 2: Simulated LINAC fan beam profile, dose rates in cGy/h

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Figure 3: Comparison of Skyshine doses by NCRP and monte carlo method

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  Abstract - 32263: Point Kernel-based shielding calculations for BLC-200 transport cask Top


Sridhar Sahoo1, T. Palani Selvam1,2, Dhiren Sahoo3, Saquib3, Prashant Dewan3

1Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, 3Board of Radiation and Isotope Technology, Navi Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Board of Radiation and Isotope Technology (BRIT) supplies equipment/radiation sources for industrial and medical applications of ionizing radiation. Depending upon the type of applications, activities of gamma sources involved are in the range of few Ci to 106 Ci. Transportation casks are used to transport these radiation sources safely in the public domain. The present study is aimed at calculating radiation levels around BLC-200 transport cask designed by BRIT that houses 200 kCi of 60Co pencil sources. The dimensions of cask are 107 cm in height and 74.5 cm in diameter. The cask can house 20 60Co sources of 10 kCi each. The cask uses lead, tungsten as shielding materials and stainless steel as casing material. Point kernel-based IGSHIELD code is used in the study and results are compared with radiometry data.

Materials and Methods: In the point kernel technique, the large volume sources are divided into many smaller volume sources, considered as point sources. From each point source, attenuation is calculated by estimating the optical path of photon from the source to the detector including interposing shielding material. The calculated un-collided dose rate is multiplied with the dose buildup factor to get total dose rate and subsequently converted to exposure rate in units of mR/h. In the simulation, 20 sources are arranged in 3 circles [Figure 1].

Results and Discussion: The exposure rates at 7 locations [Figure 2] for 200 kCi are well below the permissible limit of 200 mR/h, as stipulated in the AERB Safety Code. The study show that the transport cask has adequate shielding. As the radiometry was performed with 11 sources (total activity 47.46 kCi) arranged in the outer circle of the cage, this geometry was also simulated in IGSHIELD calculations for comparison purpose. [Table 1] compares the IGSHIELD-calculated radiation levels for 11 sources containing 47.46 kCi with the radiometry data. The agreement is good. Keywords: Point kernel method, shielding analysis.
Figure 1: Arrangement of 20 60Co source pencils in the cask

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Figure 2: Pictorial representation of the case showing locations at which radiation levels were calculated. Locations 1–6 are at the surface of the cask and location 7 is at 1 m from the surface of the cask

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Table 1: Comparison of radiation levels (mR/h) calculated using IGSHIELD with radiometry for 11 sources containing 47.46 kCi

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Keywords: Shielding analysis, Point kernel method


  References Top


  1. Subbaiah KV, Sarangapani R. Ann Nucl Energy 2008;35:2234-42.
  2. Atomic Energy Regulatory Board. Safety Code for Safe Transport of Radioactive Material, AERB/NRF-TS/SC-1 (Rev.1). Mumbai, India; Atomic Energy Regulatory Board; 2016.



  Abstract - 32271: Estimation of source term from beam loss for a self-shielded of 6 MV electron accelerator Top


K. Biju, Vitisha S. Dagre, P. Srinivasan, M. S. Kulkarni

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

E-mail: [email protected]

BARC developed Electron linear accelerators are used for many industrial and medical applications. The accelerators used for medical and industrial purposes should be self-shielded to limit the leakage radiation level (other than useful beam) within permissible leakage levels, as specified by AERB standards.[1] The source term considered for the shield design is based on the primary target source term, i.e. the bremsstrahlung dose and angular emission profile from the photon conversion target.[2],[3] In a 6 MV self-shielded electron linear accelerator (LINAC) used for Indian Cargo Scanner (ICS), the conversion target is shielded locally using tungsten alloy collimator and steel whereas the accelerating cavity was shielded with 3.2 cm thick steel such that the predicted leakage dose rates is 0.05% of that of the primary useful beam.[4] However in a practical scenario, the radiation levels measured outside this self-shielded LINAC were observed to be higher than the predicted levels by the calculations using the primary source term. The investigations showed that elevated dose rates were due to the bremsstrahlung produced by the electron beam loss in the accelerating cavity. Generally these beam losses in the medium energy electron LINAC is about 50-70%. This paper presents the study of estimation of the dose rates from a typical beam loss estimated by beam dynamics, which is significantly high and needs to be considered as a secondary source term for the shielding evaluations. The electrons from the electron gun are accelerated through the RF cavity. The electrons that are lost have an energy distribution depending on the location of loss in the cavity. The 6 MV LINAC studied here has a total length of 90 cm from electron gun to the conversion target. The [Figure 1] presents energy of the electrons and electron current lost as a function of distance in the cavity from the electron gun. The FLUKA based Monte Carlo simulations were made using the above mentioned inputs to calculate the resultant bremsstrahlung dose rates from the electrons lost in the cavity, secondary source term to be used for the shielding purpose. The [Figure 1] also presents the longitudinal distribution of the bremsstrahlung dose rate (secondary source term) computed using the FLUKA code at the surface of the LINAC with no shielding. It is found that the dose rate decreases at the end of the cavity where the beam loss is lesser by two orders. The maximum secondary source term due to the beam loss is about 30 Gy/h. The estimated dose levels at the same location after accounting for the self-shielding of LINAC is 9 Gy/h. Subsequently the total ambient dose levels on the surface of the self-shielded ICS LINAC is estimated as 10.5 Gy/h from both primary and secondary source terms . This is in good agreement with the measured ambient dose rate (within 20%), 8.6 Gy/h on the self shielded ICS LINAC. It is concluded that the present approach of radiation shield designs and ambient dose rate computations for LINAC considers only the primary beam target for estimating the photon source terms. However recent experimental observations and the case study presented in this paper clearly show that the 'photons due to beam losses in the first few accelerating cavities' also need to be taken into account for ambient dose rate calculations in addition to the primary target source. This is especially important for self-shielded LINACs deployed for industrial and security applications to ensure compliance with the leakage dose rate limits as per AERB guidelines.
Figure 1: Longitudinal distribution of energy gained by electrons in MeV, electron beam loss in μA and the computed source term from electron beam loss

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Keywords: Beam loss, bremsstrahlung, electron accelerator, shielding

Acknowledgement

Authors are grateful to Dr. Jayanta Mondal, APPD, BARC for providing the beam loss data.


  References Top


  1. Atomic Energy Regulatory Board (AERB). AERB SAFETY CODE NO. AERB/RF-MED/SC-1 (Rev. 1); March 2011.
  2. Joshi DS, Sarkar PK. Indian J Pure Appl Phys 2010;42:771-7.
  3. NCRP Report No. 144 National Council on Radiation Protection and Measurements; 2003.
  4. Atomic Energy Regulatory Board AERB/IMS/L-111/RSD/06 (Level Ill); May 2019.



  Abstract - 32273: Shield effectiveness assessment of different materials for electron cyclotron resonance ion source Top


Brij Kumar1, Pradeep Bhargava1, K. D. Singh1, M. S. Kulkarni1,2

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

E-mail: [email protected]

ECR (electron cyclotron resonance) ion sources are widely used to produce high quality multiple charged ion beams for accelerators, atomic physics research and industrial applications. Many general characteristics of ECR sources explain their widespread use in the accelerator community. One of the abilities is to produce CW beams from any element at useful intensities for nuclear and atomic physics research. Another characteristic of ECR sources is that the discharge is produced without cathodes. Therefore, only the source material injected into an ECR source is consumed. Therefore, ECR sources can be operated continuously for long periods without interruption. Nuclear Physics Division had procured one such ECR ion source based accelerator system from Pantechnik (PK-ISIS) which had been installed and operational at FOTIA.[1] It has now been proposed to shift the PK-ISIS, presently installed at FOTIA, to High Bay area in Hall-9. There it is required to estimate shielding thicknesses for different materials such that the acceptable dose rate criterion of 1 μSvh-1 is achieved. The resonance cavity which is source of X-rays, is cylindrical in shape with 100 mm diameter and 500 mm height. This paper studies the shielding effectiveness of different materials (Concrete, MS and Lead) for shielding thicknesses to achieve acceptable dose rate criterion of 1 μSvh-1 at the required locations as per the AERB regulatory stipulations.[2] The design schematics of the facility are as shown in [Figure 1]. Since the PK-ISIS is a source of X-rays. The radiation shielding adequacy study has been carried out taking into consideration the most conservative estimates of 4000 μSvh-1 at 1m in the axial direction and 2000 μSvh-1 at 1m in the radial direction (as reported by M/s Pantechnik), as the initial dose rates (source term). These values of axial and radial dose rates were measured at an operating power of 1.8 kW. Estimation of shielding thicknesses using different materials for requirement of 1 μSvh-1 as per regulatory stipulation of AERB for the proposed facility has been performed using QADCGPIC a point kernel based method code.[3] The fluence to dose conversion coefficients are taken from ANSI report.[4] All the X-rays produced are assumed to be of maximum energy corresponding to the respective operating voltage.

Results and Discussion: The thicknesses of concrete, MS and lead shields for different operating voltages have been estimated and given in [Table 1] above. It has been observed that at the maximum operating voltage of 300 kV, the estimated thickness of 40 cm for concrete is in agreement with the concrete thickness reported by the Pantechnik safety guideline document.
Figure 1: Design Schematics, layout of the facility and location of the detector positions of the ECR Ion source. ECR: Electron cyclotron resonance

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Table 1: Thickness of shielding materials for <1 μSvh.1 at 3 m axially or 2 m radially from the electron cyclotron resonance ion source for different operating voltages

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Keywords: Dose rate, electron cyclotron resonance ion source, shielding


  References Top


  1. Goubert G. Safety Guideline for PK-ISIS on 300 kV Platform, Pantechnik Document PK11-0301-IN02, Dated 27 September, 2011.
  2. AERB Guide. AERB/NPP-PHWR/SG/D-12; 2005.
  3. QAD-CGGP A Point Kernel Code System for Neutron and Gamma-Ray Calculations Shielding Using the GP Buildup Factor. RSICC, ORNL, AECL Research; 1995.
  4. American Nuclear Society. Calculation and Measurement of Direct and Scattered Gamma Radiation from LWR Nuclear Power Plants. ANSI/ANS-6.6.1-1987; 1987.



  Abstract - 32282: Radiation dose assessment for postulated initiating events of white beam line on Indus-2 Top


M. K.Nayak, G. Haridas, M. S. Kulkarni

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

E-mail: [email protected]

Indus-2 synchrotron radiation source (SRS) is in regular operation at RRCAT, Indore with a stored current of 200 mA at 2.5 GeV. The electron beam at 550 MeV from booster synchrotron is injected into the ring, ramped up to 2.5 GeV and stored for generating the synchrotron radiation (SR) in the hard x-ray region. Seventeen beamlines are under regular operation with the approval of Atomic Energy Regulatory Board (AERB) and 9 beamlines are either under trial operation/ installation/design stage. Among the operational beamlines 3 are white others are pink & monochromatic. SR beam from Indus-2 is transported to the beamline through a front end section in which many safety devices are installed. At the end of the front end, a safety shutter followed by a Be-window is installed (~ 15 m from the storage ring). The safety shutter allows/dis-allow the SR beam in the beamline through various interlocks whereas the Be-window isolates the vacuum between the beamline and the storage ring. The beamlines are installed inside specially designed shielded hutches for minimizing radiation level to accessible areas. The radiation environment of SR beam line consists of primary bremsstrahlung (BR) [solid and gas], synchrotron radiation (SR) and photo-neutrons. These radiation gets scattered from optical components of the beamline in addition to primary radiation. Due to low cross section, photo-neutrons are not considered in the present calculations. For detailed radiation dose analysis for the beamline, six initiating events are identified. The initiating events identified are:

  1. Vacuum degradation due to sputter ion pump failure
  2. Inadvertent entry to exp. Hutch
  3. Spurious opening of safety shutter
  4. Power supply failure in Personnel Safety


    5. Interlock System (PSIS)

  5. Trapping of person inside experimental hutch
  6. Safety shutter fails to close during sample


    7. Changing

    For the dose evaluation, the assumptions used are:

  7. Stored beam of 200 mA @ 2.5 GeV in Indus-2 ring.
  8. Accidental beam loss takes place for 1 s.
  9. Accidental channelling of electrons to beam


    10. lines: 2 bunches loss (2.47x109 e/bunch at 2.5 GeV)

  10. Point loss is considered.
  11. Dose assessment for primary electrons, Gas BR, Solid BR and SR.
  12. Dose evaluation is done for sample location in experimental station.


For the dose evaluation of various radiation components, source term/methodology used are from: BR,[1] electron (using flux -dose conversion factor, IAEA TR 188) and SR (using energy absorption coefficient, http://nist.gov).[4] The dose values obtained for the initiating events are listed in [Table 1] and [Table 2]. The dose/event and dose rate values obtained are listed in [Table 1] and [Table 2]. The highest dose rate at the experimental station (inaccessible during operation) is found to be as high as ~109 Sv/h due to primary SR whereas the scattered SR dose rate is ~1 Sv/h. Direct gas BR dose rate is found to be ~10-3 Sv/h in experimental station and the scattered gas BR & SR dose rate outside the experimental hutch (accessible area) is found to be negligible. Electron dose due to accidental channelling is found to be as high as ~102 Sv/event. Hence rigorous exposure control measures (shielding, radiation interlocks and emergency trip systems) are implemented in Indus-2 SR beamlines to prevent accidental exposures to SR users within the experimental hutch.
Table 1: Dose rate in experimental hutch of white beamline for initiating event number 1

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Table 2: Dose rate in experimental hutch for initiating events number 2-6

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  References Top


  1. Khan S, Nayak MK, et al. Shielding Report for Medical & Imaging Beamline, Indus-2 SRS; 2011.
  2. Nayak MK, Haridas G, et al. Radiat Prot Dosimetry 2015;164:187-93.
  3. Swanson WP. IAEA Technical Report Series No. 188; 1979.
  4. Available from: http://nist.gov.



  Abstract - 32295: Assessment of radiation shielding properties of composite polymer material Top


Amar D. Pant, Santosh K. Suman1, Amit K. Verma, K. A. Dubey2, Anilkumar S. Pillai, A. Vinod Kumar

Environmental Monitoring and Assessment Division, 1Health Physics Division, 2Radiation Technology Development Division

E-mail: [email protected]

The ionizing radiation (X-rays and gamma rays) is widely used in medical and industrial application that resulted to the demand of better radiation protection of worker and environment and hence encouraging the researcher to develop new shielding materials.[1] The photoelectric cross section of X-rays and gamma rays is high for high Z and high-density material hence are preferred for the shielding of X-rays and gamma rays. Recently, polymer composites with high Z material are widely used for radiation shielding.[2] In this work Ethylene propylene diene monomer (EPDM) based composites were developed with Strontium Ferrite (500 phr) and Tungsten Powder (300 phr). Sample sheets (10 cm x 10 cm) of varied thickness were made with compressed with moulding system. The formulations were crosslinked by gamma radiation. The crosslinking of EPDM chain by gamma radiation increased its mechanical properties significantly while retaining complete flexibility. With the loading of Tungsten Powder and Strontium Ferrite powder, the density increased from 0.9 g/cm3 of EPDM to 7.0 g/cm3. An experimental set up used for attenuation studies of the developed material is shown in [Figure 1]. HPGe based high-resolution gamma spectrometry technique was used for studying the attenuation of various gamma photons in the material. Certified standard point sources of 241Am, 133Ba and 152Eu were used for this measurement. The detector was shielded with 3“inch thick Pb to reduce natural background. The sample (composite polymer) with a thickness of 2.4 mm was placed between the two Pb collimators with a thickness of 1.4 cm and source was placed behind the collimators to avoid the scattering component. The peak intensity of gamma energy was measured with and without the composite material as a peak area of respective energy in the spectrum. Mass attenuation coefficients () of the composite material was experimentally calculated using Beer–Lambert formula given in Eq 1:



where I0 and I are the incident and the attenuated peak intensities of gamma rays, respectively. t, (g/cm2), is the mass thickness of the material. μρ (cm2 /g), is the mass attenuation coefficient. Further half value layer (HVL) thickness was calculated using Eq. 2:



μ (cm-1), is the linear attenuation coefficient

The calculated mass attenuation coefficients and HVL of the material are tabulated in [Table 1]. The mass attenuation coefficient values are higher at low energies because at low energy photoelectric interaction is dominant compared to other interactions. HVL values evident that this composite polymer material could be used as a shielding material for X-rays and low energy gamma rays that used in nuclear medical application.
Figure 1: Schematic diagram of experimental setup

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Table 1: Experimental calculated values of composite polymer material

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Keywords: Attenuation coefficient, gamma radiation, polymer, shielding


  References Top


  1. IAEA. Safety Reports Series No. 47 – Radiation Protection in the Design of Radiotherapy Facilities. Vol. 9. IAEA; 2006.
  2. Verdipoor K, Alemi A, Mesbahi A. Photon mass attenuation coefficients of a silicon resin loaded with WO3, PbO, and Bi2O3 micro and nano-particles for radiation shielding. Radiat Phys Chem 2018;147:85-90.



  Abstract - 32373: Pre-equilibrium neutron multiplicity estimates from heavy ion reactions at the energies between 10-50 MeV/nucleon: A theoretical approach Top


Maitreyee Nandy1,2 and Sabyasachi Paul3

1Saha Institute of Nuclear Physics, 1/AF, Bidhannagar, Kolkata, West Bengal, 2Homi Bhabha National Institute, 3Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

At intermediate and high energy accelerators, the emission neutron spectra can typically extend from MeV to few tens of MeV depending on the projectile energy and the interacting target material. At energies > 10 MeV/A, the non-equilibrium emission processes viz., direct and pre-equilibrium (PEQ) starts contributing significantly along with evaporation neutrons.[2] Multiple particle emissions pathways, directional anisotropy and use of thermal neutron detectors with moderation spheres make both experimental measurements and theoretical estimations equally challenging for neutrons beyond energies above 20 MeV. The direct neutrons have the highest possible discrete energies with forward angle emissions compared to the PEQ emissions, where the emission neutron energies are intermediate between direct and evaporation neutrons with forward anisotropy. The initiation or end point energy quantification is difficult and strongly dependent on the target and projectile combinations. However, the quantification of direct and PEQ emission fraction plays an important role in source terms estimations for determination of the shielding adequacy estimates at the design stages for upcoming high energy accelerator installations. In the present work, the PEQ emission neutron multiplicities were theoretically estimated for different combination of projectile and targets. The nucleon-nucleon (N-N) scattering kinematics based formalism (Nandy, 1999) with modified emission probability calculations based on realistic non-uniform spatial nucleon density dependent collision rates were used for the PEQ multiplicity estimates.[2] At higher energies, simultaneous multiple PEQ formalism from a single exciton hierarchy was considered. It was observed that the neutron energy fraction for En >20 MeV can be as high as 20% of the evaporation yields. The estimated PEQ doses at variable composite masses were evaluated to be almost 3-20% of the evaporation neutron doses depending on the target/projectile combinations. Present work estimates the PEQ neutron multiplicities in the 4π-directions with 7Li and 16O particle bombardment on natural Cu, 181Ta and 56Fe targets at energies between 10-50 MeV/A. The Cu and Ta are used as the beam dumps and Fe as the dominant structural material. The multiplicity variation shown in [Figure 1], indicates that PEQ fraction can be significantly high. For both projectiles, 181Ta yielded maximum multiplicities. However, for 7Li projectile interaction on natCu yielded higher multiplicities than 56Fe but a reverse trend was observed with 16O. Evaporation contributes up to 15-20 MeV neutron energy. The multiplicity variations are strongly dependent on the target and projectile combinations and can be explained on the basis of spatial nucleon density distributions and N-N collisions at the composite systems. During radiation monitoring around high energy particle accelerators using conventional neutron dose equivalent (NDE) meters, quantification of this PEQ fraction may significantly be underestimated due to the non-linear response of detectors for neutrons >20 MeV. These theoretical estimates can complement the underestimations in radiation dosimetry and can be accounted for a more realistic dose estimates for workplace monitoring.
Figure 1: Estimated PEQ neutron multiplicities with 7Li and 16O projectiles on different targets

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Keywords: Heavy ion interactions, neutron yields, pre-equilibrium emission, radiation monitoring


  References Top


  1. Nandy M, Ghosh S, Sarkar PK. Phys Rev C 1999;60:044607.
  2. Paul S, et al. Phys Rev C 2017;96:044607.



  Abstract - 32401: Simulation of thick target neutron yields using 14N, 16O, 20Ne beams incident on various high Z materials of dosimetric interest Top


Arup Singha Roy1,2, K. S. Varshitha1, S. K. Mishra1, R. Ravishankar1, M. S. Kulkarni1,2

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

E-mail: [email protected]

In heavy Ions accelerator, neutrons are produced when the energetic projectile interacts with various accelerating components. These neutrons are not only harmful for the occupational worker but are detrimental for sensitive instruments also. In a class IV accelerator like K130, VECC the radiation level can be as high as it can produce fatal dose to the working personnel in case of accidental trap in the vault or experimental caves. In K130 cyclotron, VECC Nitrogen (14N), Oxygen (16O) and Neon (20Ne) beams are also accelerated along with light ion beams such as proton and alpha. Higher energy heavy ion beams have also been extracted from the K500 superconducting cyclotron recently. Beam loss occurs during the transport and steering of ion beams when they hit beam line components such as dee, deflector, deflector septum, Faraday cup, beam viewer and beam transport line etc. These components are made of Cupper (Cu) Tungsten (W), Tantalum (Ta), Aluminum (Al) and SS etc. Subsequently prompt neutrons and gammas are produced by nuclear reaction during their interactions. Neutrons produced by proton and alpha beams on high Z material had been reported extensively however heavy ions induced secondary radiation data are relatively limited. Estimation of these neutron yields for heavy ion projectiles of energy 7-12 MeV/u is important for radiological protection purposes in K-130 cyclotron. So in this work simulations were carried out using Monte Carlo based codes like Geant4 as well as FLUKA to find out the neutron yield using thick target of the above mentioned material and projectile combinations. For simulations a pencil beam of monoenergetic projectile were bombarded to a thick target of the above material. Neutrons produced in the nuclear reaction are emitted from the target and detected by a sensitive detector placed outside the target. Estimated neutron yield is shown in [Table 1]. Good agreement is obtained between the results from two codes. It is observed that in the energy range considered (i.e., 7-12 MeV/u), neutron yield is roughly independent of Z of target. It is also evident from [Figure 1] that with increasing beam energy, neutron yield increases. These yield data thus obtained are found to be vital for accidental dosimetry purpose in K130 as well as K500 cyclotron facilities in VECC.
Figure 1: Neutron yields by oxygen beam on Tantalum target

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Table 1: Thick target neutron yields for 10MeV/u particle energy

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Keywords: FLUKA, Geant4, heavy ion, high Z, neutron yield


  References Top


  1. Guo ZY, Allen PT, et al. NIMB 1987;29:500-7.
  2. Trinh ND, Fadil ME, et al. EPJ Web Conf 2017;153:01018.



  Abstract - 32402: A radiologically safer production route of 48V from 16O induced reaction on chloride target: comparison of experimental and theoretical excitation function Top


Kousiki Ghosh Jana, Susanta Lahiri1, R. Ravishankar, M. S. Kulkarni

Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, 1Diamond Harbour Women's University, Sarisha, West Bengal, India

E-mail: [email protected]

In literature only few studies on heavy ion induced reactions on halide targets are available. Systematic studies using different combinations of heavy and light ion induced reactions on halide target have been carried out.[1],[2],[3] In this study evaporation residue obtained from 16O induced reaction on chloride target have been predicted and identified. 48V was produced from BaCl2 targets via natCl(16O,2pxn)48V reaction. This is an alternate production route for 48V, the common production routes are from proton or alpha induced reactions on natTi targets.[4],[5] The experiment was performed at BARC-TIFR Pelletron facility, Mumbai. The steps are as follows:

  • Six BaCl2 targets of thickness 3.5 mg/cm2 were prepared in VECC target laboratory by centrifugation technique on 1.8 mg/cm2 thick Al backing.
  • Post irradiation gamma spectrometry have been carried out using p-type HPGe detector of 2.06 keV resolution at 1.33 MeV in combination with a digital spectrum analyzer (DSA 1000, CANBERRA) and Genie 2K software (CANBERRA).
  • The energy and efficiency calibration of the detector were done using different standard sources such as 152Eu (T1/2 = 13.53 a), 60Co (T1/2 = 5.27 a), 137Cs (T1/2 = 30 a) and 133Ba (T1/2 = 10.51 a).
  • Theoretical calculations have been carried out using nuclear model codes PACE4 and EMPIRE3.2.2 in the present study for 30-80 MeV projectile energy range.


The details of irradiation parameters, yield at EOB (End of Bombardment) and cumulative cross section have been presented in [Table 1]. The experimental yields of the radionuclides were compared with the nuclear model codes PACE4 and EMPIRE. The excitation function of 48V (theoretical and experimental) obtained from BaCl2 targets have been shown in [Figure 1]. It is observed that in some experimental points, the theoretical estimations over predict the experimental data, however in few points the experimental data and theoretical predictions are close to each other. In this experiment a new heavy ion induced production route for 48V has been established. Comparison of theoretically predicted cross section and experimentally obtained cross section have been shown. The maximum yield is 1.63 kBq/μAh at 61 MeV projectile energy. No evaporation residue is produced from Ba in these projectile energies. The present production route does not produce any long lived radioisotopes which is advantageous in terms of radiation safety.
Figure 1: Comparison of theoretical values (codes:PACE4 and EMPIRE) and experimental values of excitation function for 48V produced from BaCl2 target

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Table 1: Yield at end of Bombardment and cross section of 48V at different projectile energies

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Keywords: Chloride target, heavy ion, 48V, PACE, EMPIRE.


  References Top


  1. Ghosh K, Choudhury D, Lahiri S. J Radioanal Nucl Chem 2019;321:91-5.
  2. Ghosh K, Choudhury D, Lahiri S. Appl Radiat Isot 2021;178:109966-1-5.
  3. Ghosh K, Naskar N, Lahiri S. J Radioanal Nucl Chem 2021;331:483-90.
  4. Szelecsényi F, Tárkányi F, Takács S, Hermanne A, Sonck M, Shubin Y, et al. Nucl Instrum Methods Phys Res Sect B 2001;174:47-64.
  5. Lahiri S, Banerjee S, Das NR. Appl Radiat Isot 1996;47:1-6.



  Abstract - 32444: Study of induced activity on beam diagnostic probes of super conducting cyclotron, VECC Top


Satish K Mishra, Kousiki Ghosh Jana, Arup Singha Roy, Mausumi Sengupta Mitra, R. Ravishankar

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

E-mail: [email protected]

Super Conducting Cyclotron (SCC) is undergoing beam extraction trials of various heavy ion beams like N4+(already extracted), Ne6+ etc. in first harmonic mode. Proper radiation protection surveillance is being required during maintenances. Beam losses on various components of SCC during acceleration/transportation leads to activation of the components. Beam is diagnosed at various distances inside accelerating region, hence probes (two different mechanical arrangements for diagnostic of beam) are exposed to different energy and current of beam. During radiation surveillance, it is being observed that main probe and bore probe have maximum radiation levels owing to deliberate loss of beam on them as they are utilized for beam profile diagnosis. Measurement of radiation levels was carried out during shut down on both the probes (with Cu and Al as probe head) using Radiation Survey meter (Automess-6150AD). Measured dose rate have been observed to be in the range 10 μSv/h to 3 mSv/h depending upon the beam energy, current and duration of beam tuning; and also the time lapse between last operation of machine and current period of maintenance. Monte Carlo Simulation Code FLUKA was used for identifying the isotopes contributing to the doses on probes. For this purpose, 18 MeV/A narrow beam of N4+ with irradiation profile of seven-days beam and one-day cooling and again seven-days beam (with beam intensity 105 particles/s) impinging on Cu and Al target, and RESNUCLEi card was used to identify isotopes produced and their contribution to total activity in Cu and Al material. Other isotopes whose contribution are less than 5% in total activity after 1 hour of cooling are not shown in [Table 1].61Cu Isotope have good contribution to total activity after 1 hour of cooling but after 10 hours of cooling its contribution is diminished. Other isotopes have relatively longer half-lives and contribute to total activity even after 10 hours of cooling. Measured contact dose rates on irradiated probe heads of Cu and Al block are mentioned in [Table 2] and [Table 3]. In one of the earlier study, Bore-probe with Al head was taken for gamma spectroscopic analysis and also it was observed that main contributor to the total activity was 24Na.This analysis has been validated with Monte Carlo Simulation Code FLUKA. More elaborate irradiation profile and longer cooling study using FLUKA and gamma spectroscopic study need to be done to quantify specific activity due to each isotope, so as to have better understanding of residual activity in SCC components. As the SCC is approaching towards regular operation/commissioning, this study will be helpful for better radiation surveillance in SCC facility.
Table 1: Induced radioisotopes (on Cu block) estimated through FLUKA and their contribution to the total activity

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Table 2: Measurement of contact dose rate on bore probe (Cu block)

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Table 3: Measurement of contact dose rate on bore probe (Al block)

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Keywords: FLUKA, harmonic mode, induced activity, probe


  References Top


  1. VECC Safety Committee. Super Conducting Cyclotron Operational Manual. Kolkata: VECC; 2008.
  2. Ferrari, Sala PR, Fasso' A, Ranft J. CERN-2005-10, INFN/TC_05/11, SLAC-R-773; 2005.
  3. Jana KG, Mishra SK, Mitra MS, Ravishankar R. Proceedings of Indian Particle Accelerator Conference (InPAC-2022), DAE-BRNS; 2022.



  Abstract - 32445: Dose uniformity evaluations for gamma radiation processing facility using Monte Carlo simulations and its validation  Top


Raksha Rajput, Jyoti Garg, J. R. George, P. K. Jothish, Kalpana Khedkar, N. Jayachandran, Pradip Mukherjee

Department of Atomic Energy, Board of Radiation and Isotope Technology, Navi Mumbai, Maharashtra, India

E-mail: [email protected]

Radiation processing is a fast-growing industry worldwide and so is the case in India, with a number of new processing plants commissioned over last few years. The radiation dosimetry is an essential part of the whole process of radiation processing. Though there is always a commissioning dosimetry associated with source loading in the GRAPF, there is also a need to work out the dosimetric parameters theoretically to design the best source loading pattern. The dose uniformity ratio (DUR) is a measure of the maximum and minimum dose received within the product. It is an important parameter to be realised for uniformly irradiating the products. It is thus required to determine the best combination of DUR and dose rate in the product to achieve the maximum source utilization. In the work presented here, the dose evaluations are carried out using Monte Carlo simulations based FLUKA code.[1] The simulations are run for different source loading designs by dividing the box in 9 x 7 bins for top, middle and bottom planes. The GRAPF belonging to Microtrol Sterilization Services Pvt. Ltd, Bawal, Haryana, India, is selected. The 2-Tier, split-type rectangular source frame is designed to hold a total of 400 W-91 type source pencils, supplied by BRIT, India. The medical products, of density 0.15 g/cc, are filled in the product box dimensions of 61cm x 45cm x 115cm in the tote of 2mm thick aluminium. Each box takes 8 positions per pass in a 2+2 pass system around the radiation source. The side and top view of the box movement is shown in [Figure 1]. The box also shifts the shelves to take a total of 64 positions to complete the irradiation cycle. The average dose rates are calculated for all the bins for the box taking all the positions.At first, 149.999 kCi/21.02.2020 is loaded in 6 source pencils in the Plant on the basis of source loading design provided by BRIT. Later, the source strength is enhanced to 520.607 kCi/28.10.2020 in 44 pencils. The model of source positions is shown in [Figure 2]. This paper also presents the validation of the theoretically estimated results obtained by Monte Carlo simulations with the Plant Commissioning dosimetry[2] which is an experimental result. Results: The results are given in [Table 1]. The theoretical results are in good agreement with experimental values. The theoretical estimations are also carried out for AV Gamma Tech, Ambernath which has a different source-product geometry and Hi Media Labs., Ambernath which is loaded with another type of 60Co-source pencils (BRIT made BC-188). The results are given in [Table 2]. Both of these data are also validated with the actual dosimetry performed after source loading in the Plant.[3]
Figure 1:. Side and top view of the box occupying different shelves and positions

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Figure 2:. Source positions in 2-Tier source frame for second loading (Microtrol, Bawal)

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Table 1: Comparison between theoretical estimations and experimental measurements of M/s. Microtrol, Bawal

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Table 2: Comparison between theoretical estimations and experimental measurements of other plants

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Keywords: Dose uniformity, dosimetry, source-product geometry


  References Top


  1. Ferrari A, Sala P, Fasso A, Ranft J. FLUKA: A Multiparticle Transport Code, CERN-2005-010, INFN TC 05/11, SLAC-R-773; 2008.
  2. BRIT's Plant Commissioning Dosimetry Report of AV Gamma Tech (Medical Products); 2020.
  3. BRIT's Plant Commissioning Dosimetry Reports of AV Gamma Tech (Medical Products)/2021 & Hi Media Lab. (Medical Products); 2021.



  Abstract - 32446: Experimental assessment of radiation level during commissioning of insertion device synchrotron beamline at Indus-2 Top


P. K. Sahani, B. S. Thakur, D. Waghmare, G. Haridas1, A. Agrawal2, A. Dwivedi2, R. K. Sahu

Indus Operation Division, Raja Ramanna Centre for Advanced Technology, Indore, Madhya Pradesh, 1Health Physics Division, Bhabha Atomic Research Centre, 2Beamline Development and Application Section, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Indus-2 is a 2.5 GeV, 200mA third generation synchrotron light source (electron storage ring), operational at Raja Ramanna Centre for Advanced Technology, Indore. The storage ring is housed in a circular tunnel with an outer shield thickness of 1.5 m ordinary concrete. Synchrotron radiation (SR) emitted from the insertion device (Undulator-U1 of the storage ring) is transported to the experimental station outside the concrete shield through specially designed beamline. The beamline typically of length few tens of meters is housed in shielded hutches. The radiation environment of an insertion device beamline consists of gas bremsstrahlung radiation (GBR), generated due to the interaction of electron beam with residual gas molecules within the storage ring. GBR has a broad energy spectrum extending up to energy of the electron in the ring and is transported to the beamline along with SR, Liu et al.[1] The GBR spectrum generated using FLUKA Monte Carlo code for the 2.5 GeV storage ring, Indus-2 is shown in [Figure 1], Sahani et al.[2] GBR on interaction with optical components (like x-ray mirror) produces scattered GBR. Since the SR photons from the undulator beamlines are limited to few hundreds of eV they are contained in the vacuum chamber. Thus essentially the radiation environment of an undulator beamline consists mainly of scattered GBR. The paper reports the experimental assessment of GBR in the pre-mirror hutch of the undulator beamline (BL-5, Indus-2) during commissioning trial of the beamline. During the beam transport trials at BL-5, the synchrotron beam was focussed (using a pre-mirror) and dumped on a water cooled beam dump in the pre-mirror hutch. Integrated GBR dose mapping inside the hutch was carried out using semiconductor detector based digital pocket dosimeter (PDM-112, Aloka make) and Ion chamber type direct reading dosimeters (DRD, Arrowtech make). Schematic layout of BL-5 pre-mirror hutch showing the dose mapping locations (1-10) is given in [Figure 2]. The integrated dose measurement was carried out at Indus-2 beam current in the range of 88 to 110 mA at electron energy of 2.5 GeV. The measurement indicated scattered GBR dose rate within the pre-mirror hutch in the range 0.19 to 10.75 μSv/h whereas outside up to 0.71 μSv/h was observed. The data from both PDM and DRD were found to be in very good agreement. From the data, it is understood that the radiation level towards the adjacent beamline BL-6 was higher in comparison with other locations. Hence shield augmentation and rearrangement of local shielding within the hutch were carried out and as a result the radiation field reduced to background level. Once regular operation of the beamline starts, more detailed measurements along with Monte Carlo simulation are planned.
Figure 1: Gas bremsstrahlung energy spectrum generated in undulator section of Indus-2 storage ring

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Figure 2: Schematic diagram of BL-5 pre-mirror hutch showing measurement locations

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Keywords: Gas bremsstrahlung radiation, insertion device, synchrotron radiation source


  References Top


  1. Liu JC, Rokni SH, Asano Y, Casey WR, Donahue RJ, Job PK. Radiat Meas 2007;41:S206-20.
  2. Sahani PK, Das AK, Haridas G, Sinha AK, Rajasekhar BN, Puntambekar TA, et al. Proc. 9th International Workshop on Radiation Safety at Synchrotron Radiation Sources – RadSynch. Taiwan: NSRRC; 2017.



  Abstract - 32529: Radiation protection in positive ion beam accelerator facilities Top


R. Ravishankar, M. S. Kulkarni

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

E-mail: [email protected]

Radiation protection is more complex and significant in particle accelerator facilities which have neutron field environment in addition to high gamma background. Radiation protection includes techniques involving estimation and measurement of various radiation fields prevailing in particle accelerator facilities. It also includes the instrumentation and the methods used for the thorough analysis of various strictures contributing to such mixed radiation fields, the practices used to minimise the radiation exposure prevailing in various bunkers and other active and non-active zones. The protective measures and other control procedures established to minimise the radiation exposure are also form a part of it. At Particle accelerator facilities both prompt and residual radiation are present. Prompt radiations are the radiations emitted during the beam-on conditions. Knowledge of type, energy and quantity of radiation sources in an accelerator facility is of prime importance in evaluating the complete radiation environment. Radiations are produced by interactions of accelerated beam particles with the material that it strikes (accelerator components: Dee, deflector, beam focusing & steering components, beam enclosure tubes and other beam loss components). Higher the kinetic energy of the beam particles, the greater the intensity and the number of types of secondary radiations produced. Beam particles which are intentionally produced are called primary radiations and all other radiations (no matter how they are produced) are termed as secondary radiations. In positive ion/proton accelerator facilities like VECC, Kolkata neutron dominates the radiation field followed by photons. At high energy accelerator facilities however shielding and other radiological safety considerations should be based on both neutron and photon field environments prevailing. In this work, the different radiation environment prevailing at positive ion accelerators, the possible locations of high radiation fields during operation and maintenance, the possible radiation hazards and mitigation methods, the critical radiation safety issues and challenges in radiation protection are discussed. The Peak radiation levels (after a few days of continuous irradiation) observed for 50 MeV alpha beam or above with maximum possible beam intensity and along with transferrable contamination levels observed in primary components of K-130 Cyclotron are given in the [Table 1]. The deflector system of k-130 cyclotron, VECC [Figure 1] is the most critical radioactive component in terms of exposure and possibility of personnel contamination during maintenance work. The Septum which is part of deflector system contains more loose contamination as it is porous & grainy in nature. The scenario of hazardous environment, adherence to rigorous radiation protection procedures and stringent safety practices to avoid unplanned significant radiation and avoiding the spread of radioactive contamination in the non-active areas are all critical in practice.
Figure 1: The deflector System of K-130 Cyclotron, VECC

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Table 1: Relative peak radiation levels and transferrable contamination levels observed in K-130 Cyclotron

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Keywords: Induced activity, shielding, ventilation


  References Top


  1. Ravishankar R, Bhaumik TK, Bandyopadhyay T, et al. Appl Radiat Isot 2013;80:103-8.
  2. Ravishankar R. Ph.D. Thesis University of Mumbai; 2013.



  Abstract - 33113: Alternate measurement technique for alanine dosimeter Top


Sandip Mondal1,2, S. H. Shinde1, V. Sathian1, Probal Chaudhury1, S. Adhikari2,3

1Radiation Safety Systems Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, 3Scientific Information Resource Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Amino acids such as alanine when irradiated in solid form; produces stable free radicals due to zwitterion structure. This property of alanine is used for dosimetry in radiation processing of food and allied products. The most widely used measurement technique for alanine dosimeter is ESR spectrometry (ISO/ASTM 51607 2013), which is an easy, accurate and non-destructive technique for dose measurement. But, the cost of ESR instrument is restricting its wide acceptance. Another well-established cost-effective technique is the absorption spectroscopy read-out method which is a destructive technique and involves precise weighing of irradiated and un-irradiated alanine powder and then dissolution of that powder in solution containing recommended concentration of xylenol orange, ferrous sulphate and sulphuric acid in single distilled water i.e. FX solution.[2] Absorbance measurement involves delay of about 1 hour required for the reactions to be completed after dissolution of amino acid powder in the FX solution. Thus, the read-out method is time consuming and also tedious as precise weighing is involved; which limits its use for routine dosimetry. Diffuse reflectance spectro-photometry (DRS) is another read-out for dosimetry, which was reported earlier.[3] Hence this read-out method was used in the current work as an alternative read-out for alanine powder dosimeter using spectrophotometer. All reagents viz.: L-Alanine and barium sulphate were of analytical reagent grade obtained from Merck, Germany. Irradiations were carried out in the calibrated position of Gamma Chamber-1200 using a specially designed Perspex jig for providing reproducible geometry and electronic equilibrium during irradiation. UV-Vis double beam spectrophotometer [UV 3600 Plus, Shimadzu, Japan] along with ISR-603 integrating sphere attachment was used for measuring the diffuse reflectance spectra. The DR spectra of irradiated and un-irradiated alanine powder were measured against barium sulphate. DR spectrum of irradiated alanine powder was subtracted from the average of spectra for five un-irradiated alanine powder samples in-order to obtain the net response of the alanine for a particular known dose. [Figure 1] shows the change in response against the respective wavelengths for different dose values from 1-10 kGy. The maximum response was found to be at 372 nm. These response values were found to be proportional to absorbed dose in the range of 1-10 kGy, following a 3rd order polynomial. The limitation in lower measurable dose of 1kGy is due to the sensitivity of the read-out system. The reproducibility of the DRS response was found to be within ±5%. The response is stable for at least six month duration with negligible fading. [Table 1] provides the list of the various read-out methods available for alanine dosimetry system along with the respective dose ranges. It is clear from the table that, using DRS method even spectrophotometer can be used as a read-out for alanine dosimeter up to 10 kGy yielding cost-effective and fast dose evaluations with ease, usually preferred for routine dosimetry. These suggest potential of DRS as an alternate read-out technique for alanine dosimeter for routine dosimetry in radiation processing applications.
Figure 1. : Response of alanine for different doses measured using DRS. DRS: Diffuse reflectance spectro-photometry

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Table 1: Comparison between different read-out methods for alanine dosimetry system

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Keywords: Absorbed dose, alanine dosimeter, diffuse reflectance, radiation processing


  References Top


  1. ASTM International. ISO/ASTM 51607. West Conshohocken, PA 19428, USA; 2015.
  2. Gupta BL, Bhat RM, Narayan GR, Nilekani SR. Radiat Phys Chem 1985;26:647-56.
  3. Zagórski ZP, Rafalski A. J Radioanaly Nucl Chem 1995;196:97-105.



  Abstract - 33114: Polystyrene-methyl red film: Preliminary study for routine dosimetry in radiation processing Top


S. H. Shinde, Sandip Mondal, V. Sathian and P. Chaudhury

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

E-mail: [email protected]

Radiochromic film dosimeters are suitable for routine dosimetry in radiation processing because of the ease of use. These films are available commercially but their cost restricts the widespread use in our country. There are number of other film dosimeters which are reported[1],[2] but are not available commercially. Hence an attempt was made based on the reported research hypothesis to develop a cost-effective film dosimeter involving methyl red in polystyrene matrix. The film was prepared by mixing 0.5 mM methyl red [MR] and 20% w/v polystyrene [PS] in chloroform. Mixture was stirred continuously in the dark at room temperature using a digital stirrer till a uniformly colored PS solution was obtained. All reagents and solvents were obtained from Merck, Germany and used without further purification. All pre-cleaned glasswares were used. In order to cast the film from this solution, 10 cm diameter glass dish with a flat base & a height of 2 cm was taken. Mercury was then poured in this glass dish till about a height of 1 cm. A metal ring having a diameter of about 8 cm and thickness of about 1.5 cm was carefully placed in the mercury. Required volume of the solution was then carefully added using graduated pipette onto the mercury in the ring and left in the fume hood for about five days. Solution is dried under dark conditions for normal laboratory temperatures without using heat or blower. This is necessary to obtain a clear transparent film without wrinkles. The dried film was then pulled by tweezers carefully and smaller films having dimensions 2 cm (width) x 4 cm (length) were cut. 50 randomly selected films were used for the thickness measurement using high precession digital thickness gauge [Elcometer; UK]. As compared to the conventional film development method, the film casting method used during these studies produced uniformly thick films, each having an average thickness of 0.122 ± 0.017 mm. Gamma Chamber-1200 used for irradiation, was calibrated at the center position of its irradiation volume as per the recommended procedure (ISO/ASTM 52116 2008); using Fricke dosimeter - a reference standard (ISO/ASTM 51026 2015). Specially designed perspex jig was used for providing reproducible geometry and electronic equilibrium during irradiation of the films. Irradiation temperatures encountered during irradiation were around 30oC. The absorbance spectra of un-irradiated and irradiated films were measured in the wavelength of 445 to 545 nm using a double beam UV/Vis spectrophotometer [UV3600 plus, Shimadzu; Japan]. Net absorbance spectra for each of the doses were obtained by subtracting spectrum of irradiated from the average spectrum of un-irradiated films and are as represented in [Figure 1]. It is clear from figure, that net absorbance increases with increase in dose. This is because radiation-induced discoloration of PS-MR film increases with absorbed dose; decreasing the absorbance of the irradiated film and thus increases the difference between the absorbance of un-irradiated and irradiated film. Maximum net absorbance peak for all the respective spectra was found to be at 506 nm. The response of this film may be affected by external factors such dose-rate, dose fractionation, irradiation temperature and relative humidity, hence detailed dosimetric studies needs to carried out and the response of the films has to be calibrated under actual processing conditions. However, preliminary dose response of this film suggests that it has potential use for routine dosimetry in radiation processing in the dose range of 5 –150 kGy.
Figure 1: Net absorbance spectra of PS-MR film dosimeter. PS-MR: Polystyrene-methyl red

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Keywords: Chemical dosimetry, film dosimeter, methyl red, radiation processing, spectrophotometry


  References Top


  1. Bhat NV, Nate MM, Bhat RM, Bhatt BC. Ind J Pure Appl Phys 2007;45:545-8.
  2. Chen YP, Liu SY, Yu HQ, Yin H, Li QR. Chemosphere 2008;72:532-6.
  3. ASTM International. ISO/ASTM 52116. West Conshohocken, PA; 2013.
  4. ASTM International. ISO/ASTM 51026. West Conshohocken, PA 19428, USA; 2015.



  Abstract - 33115: Testing of leuco crystal violet visual indicator in commercial radiation processing facility Top


S. H. Shinde, Sandip Mondal, V. Sathian, P. Chaudhury

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

E-mail: [email protected]

Radiation processing is one of the important technologies for food preservation and processing. In India, growth of radiation processing industry depends mainly on cost-effective technologies supporting the industry. There is increasing demand for visual indicators for providing an easy identification and segregation of irradiated products. Use of films containing leuco crystal violet (LCV) in poly vinyl butyral for high dose dosimetry in the range of 1 – 100 kGy has been reported.[1] Thus, cost-effective visual indicator based on Leuco Crystal Violet [LCV] for doses ≥ 10 kGy was developed. This visual indicator changes color from white to purple on irradiation. The change in color is measurable from 500 Gy onwards upto 30 kGy but is visually distinguishable for doses >= 10 kGy only. The mechanism of the radiation-induced color change of the LCV can be attributed to formation of the highly colored quinoid chromophore as a part of resonant carbonium cation.[2] Current works deals with preparation of visual indicators in laboratory and testing in a commercial radiation processing facility viz.: M/s Akshar Gamma Steriles LLP. All reagents and solvents were obtained from Merck, Germany and used without further purification. All glasswares were cleaned before use. 3 % v/v of LCV was added to solvent mixture of trichloroethanol and toluene (1/4 v/v) containing 20% polystyrene. All the reagents were mixed thoroughly using a digital stirrer for a period of 2 hrs under dark conditions. Strips each having dimensions of 30 mm width and 60 mm length, were cut from a single A4 size self-adhesive commercially available label paper. Protective covering from the adhesive side of paper was not removed. Length of 50 mm of these strips were dipped into the indicator solution and kept for drying in dark place for 3 hrs. Uncoated part of the strip was cut off after drying. Coating procedure followed was able to produce uniformly thick indicators, each having an average coating thickness of 0.0249 mm, measured by high precision digital thickness gauge (Elcometer, UK). Each of these strips were cut into size of around 20 mm width and 20 mm length. Coated side of strips were protected with commercially available self-adhesive one-way transparent silicon tape which also acts as UV protector. A batch of 100 such strips were prepared and stored in aluminium foils under normal laboratory conditions. M/s Akshar Gamma Steriles LLP is a commercial radiation processing facility situated at Ambernath (E) in Maharashtra. This facility is designed for maximum source strength of 37PBq of Cobalt – 60. The product movement consists of an automatic conveyor system which conveys the product boxes into the cell room through labyrinth. Product box has dimensions as 120 cm length, 170 cm height and 57.5 cm width. This facility is mainly used for sterilization of medical products and radiation processing of food and allied products. Visual indicators were provided to this facility with the instructions that the indicator strips were not to be exposed to direct sunlight before irradiation and to be stored in normal temperature conditions. Protective covering on the adhesive side of the strip was removed and strips were pasted onto the outer sides of the product boxes. Each product box was filled with turmeric powder leading to bulk density of 0.34 g/cc. Before loading all these boxes on the conveyor system, ceric-cerous and glutamine dosimeters were placed inside one of the boxes in-order to measure the dose delivered to the product and measured after irradiation respectively by facility and our division [RSSD] respectively. Average dose of about 10.5 kGy was evaluated using both the dosimetry system. Average irradiation temperature of 31oC was reported by the facility. It is evident from [Figure 1] that the colour of irradiated indicator strip has changed to purple from the initial white colour which indicates that the product box is irradiated to dose of >10 kGy. Further it was observed that there was no significant fading in colour of the irradiated indicator strips that were stored exposed to normal light and temperature conditions for period of six months.
Figure 1. : Irradiated and unirradiated LCV indicators. LCV: Leuco crystal violet

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Keywords: Leuco crystal violet, radiation indicator, radiation processing, visual indicator


  References Top


  1. Soliman YS, Abdel-Fattah AA. Radiat Meas 2013;49:1-6.
  2. Shinde SH, Mondal S, Sathian V. IPTEK J Tech Sci 2019;30:2088-33.



  Abstract - 33121: Development and characterisation of standard 241Am sources Top


Ritu Sharma, D. B. Kulkarni, R. Anuradha, V. Sathian, Probal Chaudhury

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

E-mail: [email protected]

Numerous methods for the preparation of alpha emitters have been well established and effective since last few decades. Nevertheless, electrodeposition method has become a preferred technique for obtaining thin, homogenous sources for high resolution alpha spectrometric measurements of actinides.[1] Actinides are a serious concern in environmental monitoring, nuclear power plants and industrial activities. To detect their presence in routine samples, accurate and reliable estimation of alpha activity is required. This estimation is highly dependent on the performance of detector for the radionuclide in question. Calibration of the detector with standard sources is thus essential to accurately measure the activity of samples and check the performance of detector. Radiation Standards Section, RSSD develops alpha standard sources to fulfil these requirements. Prior to the preparation of sources by electrodeposition, standardisation of 241Am solution is required. 4πβ(PS)-γ and 4πβ(LS)-γ coincidence counting systems were used for determining the activity concentration (Bq/g) whose international equivalence has been established by participating in international intercomparison programmes. Using the activity concentration from primary methods, known amount of activity weighing 10-40 mg was added in the electroplating cell and electrodeposition was performed on stainless steel (SS) planchette (cathode) in Ammonium Sulphate medium at pH 2.2 with a steady current of 300 mA at 6 V for 2.5 hrs.[2] For uniform plating of Americium on SS planchette, spirally wound Platinum was employed as anode.[3] These electrodeposited sources were analysed for activity measurements, source efficiency and homogenous deposition. The deposited activity on the planchette was determined by measuring the activity of remaining solution along with washings after electroplating. This solution was transferred to standard injection glass vial and was counted using HPGe, a secondary standard maintained by the laboratory which is calibrated against the primary standards. The efficiency of HPGe detector for 241Am radionuclide was determined by preparing a standard source in geometry similar to that used for measuring remaining solution. The source efficiency of planchette was determined by counting it in 2π geometry of proportional counter which has an intrinsic efficiency of 100 %. To investigate the homogeneity of deposition in electroplated sources, autoradiography and radial activity distribution plots were studied. GAFCHROMIC EBT3 film was used for autoradiography where active material was sandwiched between two 125 μm polyethylene terephthalate (PET) polymer sheets. As the alpha radiations were attenuated by these PET sheets, one side of the PET was peeled off and the electroplated source was kept inverted for an exposure time of ~30 days to expose the active layer. The exposed film was read by EPSON 11000XL Flatbed Scanner and data was acquired from the software FILMQA PRO. Activity concentration of 241Am solution was obtained as 34.06 Bq/mg as on 16/12/2021. The uniformity of deposition by autoradiographic studies is shown in [Figure 1]. The strength of exposure is depicted with colours. Highest exposure is represented by the colours black and brown, low exposure by the colour red, and no exposure by the colours blue and green. The homogeneous distribution of 241Am on the electroplated source in both the X and Y directions is further supported by the plateau obtained in radial activity distribution graphs [Figure 2]. Seven homogeneously electroplated sources with activities 30 to 450 Bq were successfully prepared with consistent yields of 86% to 89%. These sources can be used as alpha standards and are ready for dissemination to users.
Figure 1: Autoradiograph of electroplated 241Am on source 9889

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Figure 2: Radial activity distribution plots of source 9889 in X axis (left) and in Y axis (right)

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Keywords: Actinides, activity, electrodeposition, homogeneity, standardisation


  References Top


  1. Talvitie NA. Anal Chem 1972;44:280-3.
  2. Prabhu SP, Sawant PD, Bhati S. Radiat Prot Environ 2010;33:137-9.
  3. Klemencic H, Benedik L. Alpha-spectrometric thin source preparation with emphasis on homogeneity. Appl Radiat Isot 2010;68:1247-51.



  Abstract - 33186: Establishment of reference diagnostic X-ray beam qualities using in-house developed free air ionization chamber Top


Rahul Kumar Chaudhary, Sudhir Kumar, S. D. Sharma, B. K. Sapra

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

E-mail: [email protected]

Introduction: The standardization of medical diagnostic x-ray beam is required for the calibration of the dosimeters and quality assurance tools used in the diagnostic radiology. As per IAEA TRS-457 guidelines,[1] standardization of the diagnostic x-ray beam involves two steps: (1) determination of the additional filtration, and (2) measurement of the first half-value layer (HVL) and homogeneity coefficient (HC) of the hardened beam. This paper describes the process for establishment of the reference diagnostic x-ray beam qualities.

Materials and Methods: Siemens Polydoros LX diagnostic x-ray machine was used to generate the x-ray beams and 0.65 cc Farmer type ionization chamber was employed to monitor the fluctuations of the x-ray beam output. Aluminium (Al) filters (99.5% purity) of 100 mm diameter and 0.05, 0.1, 0.2, 0.4, 0.8, 1.6, 3.2, 6.4, 12.8 and 25.6 mm thicknesses were used in different combinations to generate the attenuation curves for 40, 50, 60, 70, 81, 90, 100 kVp diagnostic x-ray beams. Free air ionization chamber (FAIC) was placed along the central axis of the beam at the distance of 100 cm from the focal spot of x-ray tube. Field size of 20 x 20 cm2 was opened. High voltage (-2600 V) was applied to operate the FAIC in the saturation region. The attenuated air kerma (Katt) and the unattenuated air kerma (Kunatt) was measured using the FAIC. Thereafter the ratio of Katt to Kunatt (i.e. Katt/Kunatt) was determined. The attenuation curves were plotted between the logarithm of (Katt/Kunatt) and Al filter thickness. Additional filter thicknesses were determined for 40, 50, 60, 70, 81, 90, 100 KVp x-ray beams using the attenuation curves following the procedure mentioned in IAEA TRS-457. The additional filtration determined for each diagnostic x-ray beam was placed in the beam and the attenuation curves were plotted again. The first HVL and HC (ratio of first to second HVL) values were obtained from the attenuation curves generated with additional filtration. The measured values of first HVL and HC were compared with the values given in IAEA TRS-457.

Results: [Table 1] shows the measured values of first HVL taken at corresponding value of (Katt/Kunatt) listed in the table. IAEA TRS-457 specified values of first HVL for the listed beam qualities are also shown. [Table 2] shows the measured and IAEA TRS-457 specified values of HC. It can be observed from [Table 1] that the values of (Katt/Kunatt) used for determination of first HVLs lie within IAEA TRS-457 specified values (0.485 - 0.515). Further, it can be observed from [Table 2] that the measured values of HC are within 0.03 of IAEA TRS-457 specified values. As applied voltage of 80 kVp was not available in the x-ray machine used in this work, nearest applied voltage of 81 KVp was considered for standardization and comparison. Based on the data presented in these tables, it can be inferred that the beam qualities have been established as per IAEA TRS-457 specifications.

Conclusion: The agreement between measured and IAEA TRS-457 standard values establishes the diagnostic x-ray beam qualities as reference beam which can be used for the calibration of dosimeters used in diagnostic radiology.
Table 1: Measured and technical report series number 457specified first half-value layer values along with (Katt/Kunatt) measured values

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Table 2: Measured and technical report series number 457specified homogeneity coefficient values

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Keywords: Diagnostic X-ray beam, free air ionization chamber, homogeneity coefficient, standardization


  Reference Top


  1. International Atomic Energy Agency, IAEA, 2007. Dosimetry in Diagnostic Radiology: An International Code of Practice, TRS-457. IAEA, Vienna, Austria.



  Abstract - 33328: Evaluation of energy dependence of EBT3 films response in different beta energies for use in dosimetry and comparison exercise in beta radiation field  Top


S. Rakshit1,2, M. S. Kulkarni2,3, S. P. Vinatha1, V. Sathian1

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

E-mail: [email protected]

Dosimetry in beta radiation fields is an important aspect in the field of radiation protection for personnel working in an environment where beta radiation field is present. Standardization of beta radiation fields and calibration of personnel and area monitors used in beta dosimetry for this purpose is important. Various dosimetric techniques for measurement of beta dose to tissue are discussed in ISO 6980.[1] A suitability study of EBT films for dosimetry in beta radiation was reported by Benavente et al.[2] Another type EBT3 radio chromic films (Ashland Specialty Ingredients, USA) are widely used in radiation therapy dosimetry. Advantages of using EBT3 films are their high spatial resolution and usability in a wide dose range up to orders of tens of Gy which is common in case of radiation therapy. In the present work, suitability of EBT3 films in dosimetry in beta radiation field and their suitability as a transfer standard and for dose comparison exercises are studied for 85Kr, 90Sr-90Y and 106Ru-106Rh beta energies. The reference beta radiation fields from 85Kr, 90Sr-90Y and 106Ru-106Rh radionuclide were standardized using extrapolation ionization chamber as per ISO 6980[1] guidelines. EBT3 films were cut into rectangular strips for irradiation in these beta radiation fields for dose of 0.5 Gy to 10 Gy. The irradiated films were digitized using EPSON 11000XL flatbed scanner in transmission mode. The films images were analysed using ImageJ software for red, green and blue colour channel. Changes in optical density (△OD) of the irradiated films were evaluated for different beta doses from different beta sources. The △OD of each film strips were evaluated using the following equation
Figure 1: Typical Response of EBT3 film in different colour channel (for 106Ru-106Rh beta source)

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Table 1: Sensitivity of EBT3 film at different beta energies

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Here, PVtransparent, PVirradiated and PVunirradiated are mean pixel values measured from the scanner for transparent medium, irradiated film strip and unirradiated film strip. Sensitivity of EBT3 films were evaluated in term of △OD/Gy for different beta energies. Sensitivities for different colour channel used for the analysis are mentioned in [Table 1]. The sensitivity of similar EBT3 film in 60Co beam is also mentioned in the [Table 1]. [Figure 1] shows response of EBT3 film for different doses in beta radiation field from 106Ru-106Rh radionuclide. It also depicts variation in response of EBT3 films for three different colour channels used for analysis. Response of EBT3 films were studied in 85Kr, 90Sr-90Y and 106Ru-106Rh beta energies and sensitivity of EBT3 films found to be energy dependent within 20% for the said beta radionuclides. Response of EBT3 films in beta radiation fields suggests that EBT3 films can be used as transfer standard and for dose comparison exercises with an associated uncertainty of 4% or less for dosimetry in beta radiation fields considering batch wise calibration of the films is required in beta radiation fields of different energies.

Keywords: Beta dose, EBT3, film response, radiochromic film, skin dose


  References Top


  1. ISO 6980-Part 2, 2004, Nuclear Energy Reference Beta-Particle Radiation Part 2: Calibration Fundamentals Related to Basic Quantities Characterizing the Radiation Field; 2004.
  2. Benavente JA, Meira-Belo LC, Reynaldo SR, DaSilva TA. Feasibility of EBT Gafchromic films for comparison exercises among standard beta radiation fields. Appl Radiat Isot 2012;71:52-6.



  Abstract - 33329: Automation of new 4π-γ ion chamber a secondary standard used for establishing traceability of activity measurements in the country Top


Anuradha Ravindra, A. P. Das1, D. B. Kulkarni, Ritu Sharma, V. Sathian

Radiation Safety Systems Division, BARC, 1Division of Remote Handling and Robotics, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Calibrated well type re-entrant ionization chambers are maintained as secondary standards of activity for gamma emitting radionuclides by national metrology institutes to establish traceability of activity measurements as well as to assure the quality of standards disseminated to users in the country. Radiation Standards Section, RSSD maintains a high-pressure 4π gamma ion chamber (GIC) as secondary standard calibrated from primary standardizations. The system is used for calibration γ sources and for national audit of 131I radioactivity measurements at Nuclear Medicine Centers (NMCs) as a quality audit program to deliver quality and accurate diagnosis or treatment, the patients receive at NMCs. 19 such audits were conducted till date. With growing demand for nuclear medicine procedures number of participants has escalated over the years. Despite the pandemic about 200 NMCs participated in the recent 19th national audit held in 2020, and 200 sources were standardized as each NMC is provided with a standard 131I source of nominal activity 110 MBq. Each source was measured 6 times to improve the uncertainty associated with activity measurements. Thus, total number of measurements carried out during a single audit was 200*6 =1200. As 131I has a short half-life of 8 days all measurements must be carried out within the short available time, which is very challenging. Hence, a new GIC system is in the development for automation. This new system makes use of a commercial digital electrometer to measure the ionization current and a custom designed robotic sample changer. The new system consists of a storage bay, the transport section and the measurement zone. At a time, the storage bay can store 40 samples and 3 reference sources. The sources are kept inside lead cask on table, shielded by lead bricks and SS plate, [Figure 1]. To cater to different source geometries and obtain a common interface, each source to be measured is placed in a low-density plastic jig as shown in [Figure 2]. The thickness of the jig is optimized for minimal attenuation of the γ's while counting. A XY2Z robotic arm is equipped with two Z axes each, having a gripper provided for handling the source jig. The transport section transfers the source from storage bay to the measurement zone. The XY2Z robotic arm picks the source with the jig from the storage bay, loads it in the transfer shuttle lead cask which moves linearly at speed 0.5 m/s to transfer the source to measurement zone i.e., GIC. One transfer which includes extracting the vial from storage bay and loading it into the GIC counter takes 50 sec A large 6 m distance is deliberately maintained between the storage bay and GIC to minimize the background of GIC due to sources stored in the bay area as well as its contribution during the source current measurements. The Y-Z robotic arm which is in line with the Y-Z plane picks the source jig from the shuttle cask, clamps it on the chamber loading unit which lowers the source into the well of GIC. The current data is acquired through the serial port available in the electrometer and the data acquisition software shall be integrated with the automation software so that the completion of data acquisition is indicated by the automation software. Once data acquisition is completed the source is picked from the GIC and placed back into the lead transfer shuttle which moves linearly to the storage bay. This entire process is repeated for all the selected sources in the storage bay. The complete process is automated and the motions are pre-programmed. It can be remotely computer controlled with a user-friendly software integrated with data acquisition software which facilitates the movement of source back and forth and its measurements. This new automated GIC will be established as a secondary standard for radioactivity measurements. Automation caters to the national demanding requirement of nuclear medicine centers where large number of sources can be standardized in the available short half-life. With minimum human intervention, large batches of sources and frequency of samples measured is increased leading to sizeable data resulting in better accuracy, additionally remote computer controlled minimizes radiation exposure of the personnel.
Figure 1: Schematic of GIC automation

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Figure 2: Source holder

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Keywords: Ionization chamber, radioactivity, standardization, traceability


  Abstract - 33334: Studies for the sensitivity evaluation for newly developed reference ionization chamber using Fluka simulation Top


Liji Shaiju, Sunil K. Singh, Shobha S. Ghodke, S. M. Tripathi, V. Sathian

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

E-mail: [email protected]

Radiation Standards Section is in the process of developing a new reference standard, ionization chamber for air kerma measurements for protection level application. A volume of 100 cc was chosen for the ionization chamber as it will cover a range from ~0.1 mGy/h to 10 Gy/h. The absorbed dose can be determined in any material by applying the Burlin cavity theory to measurements made by a cavity ionization chamber with walls constructed from an air equivalent material. Graphite, being an air equivalent material, was chosen as the wall material for the ionization chamber. The code FLUKA2021 Version 2.2 Sep-21[1] was used to simulate the radiation transport in the ionization chamber to determine the sensitivity for the various energies from 15 keV to 2 MeV. The cylindrical ionisation chamber as in [Figure 1] has a height to diameter ratio 1.32, which ensures minimum correction due to photon intensity variations over the chamber dimensions.[2] Using simulation, the thickness required for charged particle equilibrium (CPE) in graphite was determined as 4mm for energies up to 2 MeV monoenergetic photon energy, thus graphite wall of the chamber 4 mm thick is sufficient. For simulation, the chamber was housed inside a room of dimensions 5 m × 5 m × 5 m to take into account the effect of Compton scattered photons in air. Photon tracks were initiated from a point source having an isotropic distribution and distance between source and detector center was 100cm. The room was filled with air of density 1.205 × 10-3 gcm-3, and graphite of density 1.7 gcm-3 was used as wall and anode of ionization chamber. Usrbin card was used to determine the photon energy deposition in the detector per source particle. Simulation was carried out with 109 primaries in a cycle within 1% uncertainty. To determine the total energy transfer, the energy threshold for the production and transport was set at slightly higher than the energy of the secondary electron produced. Air Kerma per source particle was calculated as follows,



To determine the energy deposited in the detection volume, energy threshold for electron and positron production was set at 1 keV and for photon production set at 100 eV. The energy deposited per primary in the sensitive volume of the chamber was converted to charge by the following relation,



The sensitivity of the ionization chamber was determined for energies from 15 keV to 2 MeV using simulation. The sensitivity from 150 keV to 2 MeV is found to be constant within ±2.6% at 20°C, 1atm with average value 3.43E-6 C/Gy.

Conclusion: The chamber is fabricated and preliminary measurements were carried out using the ionization chamber with 137Cs and 60Co sources to verify the reference radiation field which is within ±1% of the calculated sensitivity. Other studies like long term, short-term stability studies, etc. are to be performed to establish the chamber as the reference standard.
Figure 1: Cross section of simulated cylindrical ion chamber

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Figure 2: Relative response of chamber w.r.t 137Cs from 15 keV to 2 MeV using simulation

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Keywords: Air Kerma, ionization chamber, reference standard, simulation


  References Top


  1. Ferrari A, et al. Fluka: A Multi Particle Transport Code, FLUKA Website. Available from: http://www.fluka.org.
  2. Kondo S, et al. Effect of finite size of Ionization chambers on measurements of small photon sources. Radiat Res 1960;13:37-60.



  Abstract - 33335: Establishment of a facility for calibration of low level gamma radiation monitors: A preliminary study Top


Sunil K. Singh, Liji Shaiju, V. Sathian, Probal Chaudhury

Radiation Standards section, RSSD, HS&EG, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

Radiation levels encountered at work place, in various applications of ionising radiation, are well below 1 μGy/h (~ 100 μR/h). Therefore, to check such low level radiation fields, suitable low range gamma radiation monitors are frequently used for routine health physics applications, environmental monitoring and other regulatory purposes. These radiation monitors need periodic calibrations to ensure that they are still working properly and are suitable for intended use. Calibration of these radiation monitor (hand held or fixed) are also carried out after they undergo repairs or if any doubt arises on the accuracy of routine measurements. Calibration ensures reliability in measurements and provides measurements traceable to national standards. However, calibration at radiation levels below 2 μGy/h becomes a challenging task for most of the calibration laboratories which generally use high activity radionuclide sources to cater calibration of all types of radiation monitors covering the range till 10 Gy/h. Hence, gamma background radiation levels in these calibration laboratories stays ~ 0.3 – 1.0 μGy/h which varies with different locations in the laboratory. Therefore, calibration of radiation monitors below 2 μGy/h requires a dedicated facility with low activity gamma radionuclide source/s (137Cs). A dedicated calibration facility consisting of low (AL) and high (AH) activity panoramic, point 137Cs sources along with distancing / positioning system is established [Figure 1] for the calibration of low range gamma radiation monitors at BARC, Trombay. A distance of ~ 70 – 580 cm is used to get lower radiation levels starting from ~ 0.2 μGy/h (over and above background using 1.5 mCi 137Cs source) using low activity 137Cs source (AL). During establishment of facility, a high activity 137Cs source (AH) source is positioned in a source holder (at a height of 135 cm). Air kerma rate from this source was measured (say,Km,d ,AH) at various distances, d (70 - 580 cm from source centre), using a standard ionization chamber[2] coupled with electrometer.[1] Thereafter, source AH is removed and is placed in source storage pit (in other lab) to bring background radiation (to natural background < 0.1 μGy/h) and then source AL is placed in the same source holder (both sources have same physical dimensions). Radiation field from the source AL is measured at 70 cm (Km,70,AL) using the same standard PTW instruments. Due to low activity, signal becomes too low for measurement at other distances, hence measurement could not be done. Considering that scattering contribution pattern will vary exactly in the same manner as it varies for high activity source (as geometry remains same), air kerma rate at other distances is extrapolated / calculated (Kcal ,d ,AL) based on a scaling factor (f), as per equation 1. where,





A value of scaling factor (f), with distance, is shown in [Table 1] below. From the table, a significant deviation of air kerma rate from inverse square law can be seen. Thus, in panoramic irradiation geometry, only standardizing radiation field at one point and applying inverse square law for other distances may not be adequate for calculating air kerma rate at other distances. However, as per presented method, measuring radiation field at one point and calculating air kerma at other distance will give accurate value. Thus, evaluation of dose rate inclusive of scattering contribution is must before using such values. The established facility is dedicatedly used for calibration and testing of low-level gamma radiation monitors. Future work: The reported work is based on the use of panoramic radionuclide source however, in future the facility will be upgraded with a collimated source.
Figure 1: Sketch of calibration facility at BARC, Trombay

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Table 1: Variation of scaling factor with distance

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Keywords: Air kerma, calibration, inverse square law, low range radiation monitors


  References Top


  1. PTW UNIDOS E. Available from: http://www.ptw.de/unidos_e_dosemeter_ad0.html.
  2. 1L PTW Ionization Chamber. Available from: https://www.ptwdosimetry.com/fileadmin/user_upload/Online_Catalog/DETECTORS_Cat_en_16522900_13/blaetterkatalog/index.html#page_54.



  Abstract - 33336: Measurement of thermal neutron fluence rate at D9 location of critical facility, BARC using foil activation technique Top


Meghnath Sen1,2, S. S. Ghodke1, Y. Singh1, R. B. Rakesh1, V. Sathian1, S. K. Verma3, P. Chaudhury1

1Radiation Safety Systems Division, Bhabha Atomic Research Centre, 3Reactor Operation Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Anushakti Nagar, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Critical Facility (CF) is a low power research reactor which is mainly used for the validation of computer codes for reactor physics of AHWR reactors and studying reference core of different types of reactors. However, it is also extensively used for the applications such as, neutron activation analysis, characterization of archeological samples, measurement of neutron sensitivity of various thermal neutron detectors used in power reactors, response study of neutron sensitive solid-state detectors etc. using specific irradiation locations (D9 and D10).[1] Majority of these applications require accurate estimation of the thermal neutron fluence rate present at the irradiation location which varies depending upon the detector dimension. Keeping different requirements in view the thermal neutron fluence rate was determined over the useful length of the D9 irradiation channel using standard gold foil activation technique. The paper describes the methodology adopted and presents the measurement results. Materials and Methods: Standard gold (197Au) foils (purity 99.99%, diameter: 10 mm, make: Advent) were irradiated at the D9 location over the length of 90 cm from the starting point. A total of ten gold foils were used out of which nine were bare and one was kept under Cadmium cover. The bare foils were kept 45 cm and 2.5 cm apart in the horizontal and vertical planes, respectively with respect to the axis as pictorially depicted in [Figure 1]. The Cadmium covered foil was kept to evaluate the Cadmium ratio in close proximity of core. All the foils were placed on an Aluminium stand. The exposure was carried at reactor power of 30 Watt for 212.13 minutes.

The thermal neutron fluence rate was determined using the equation:



where,Φ: thermal neutron fluence rate, A: measured bare foil Activity, N: Number of 197Au atoms, σ: Activation cross section for thermal neutron (98.6±0.3), tirr: Irradiation time, td: Decay time, λ: Decay constant of 198Au (0.000179 min-1), and RCd: Cadmium ratio. The induced activity of the 198Au foils were measured using 4πβγ coincidence system, the absolute standard for activity measurement.

Results and Discussion: The measured thermal neutron fluence rate at different location is shown in [Table 1]. The overall uncertainty in the measurement is ±5% having the components viz. A, N, σ, tirr, td, λ and RCd. The Cadmium ratio (RCd) was found to be 46.1 which suggest that about 97.8% induced activity was due to thermal neutrons. In horizontal plane the fluence rate at location 'G' was 0.54 times the fluence rate at location 'A' which may be due to the higher distance of 'G' from the center of reactor core. Whereas in vertical plane the fluence rate at location 'I' is 0.88 time the fluence rate at location 'Indicating downward decrease in the fluence rate. Apart from detector sensitivity measurement, the measured data is highly useful for the dosimetric response study of various thermal neutron sensitive solid-state dosimeters.
Figure 1: Arrangement of 197Au foils in D9 location of Critical Facility, BARC

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Table 1

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Acknowledgements

The authors are grateful to Dr. D. K. Aswal, Group Director, HS&EG for his encouragement and support in carrying out the work. Thanks to all the CF members for their help during the foil irradiation's studies.


  Reference Top


  1. De SK, et al. Operation and Utilisation of Low Power Research Reactor Critical Facility for Advanced Heavy Water Reactor, IGORR; 2014.



  Abstract - 33344: Evaluation of microdosimetric quantities of 252Cf source using inhouse developed cylindrical tissue equivalent proportional counters  Top


Shobha Ghodke, R. B. Rakesh, Meghnath Sen, Sunil Singh, Yashoda Singh, V. Sathian

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

E-mail: [email protected]

Introduction: Tissue Equivalent Proportional Counters (TEPC) are widely used in radiation biology studies and radiation protection. TEPC filled with tissue equivalent gas at low-pressure works on Bragg-Gray cavity principle and records ionising radiation energy deposition per event in a microscopic volume simulating a human tissue. It measures the stochastic quantities lineal energy transfer, y, and specific energy, z of the radiation, analogues to macroscopic quantities Linear Energy Transfer (LET) and absorbed dose respectively.[1]The mean quality factor and the absorbed dose estimated in terms of the frequency-mean specific energy are used to estimate dose equivalent.

Materials and Methods: A right circular cylindrical TEPC is developed with specification given in [Table 1] as reference standard for absorbed dose and dose equivalent. This paper describes the estimation of the micro dosimetric quantities and dose equivalent for 252Cf neutron source using it and the validation of measurement with fluka simulation. The pressure of the gas is chosen to simulate the 2μm diameter (d) human tissue. The TEPC is placed inside an aluminium vacuum chamber. The simulation of TEPC was done for 252Cf neutron source using FLUKA2011 Version 2x.2 May-18. Distance between the source and center of TEPC was 100 cm. The source was assumed to be in conical form and defined using user subroutine SOURCE.f to reduce the simulation time and uncertainty. The input geometry of TEPC and source configuration is shown below. Detect card was used to score energy deposition inside TEPC on an event-by-event basis. Measurement was done at distance of 20 cm from source to centre of TEPC using 252Cf source having strength 1.22 x 106 n/s. The spectrum was acquired for 7200 sec. The lineal energy (y) was calculated as the ratio of the energy imparted, ε to the matter from a single energy deposition event by mean chord length, [INSIDE:]. The absorbed dose (D) is given by
Figure 1: Left: Input geometry of TEPC (side and top view), Right: source configuration of fluka simulation. TEPC: Tissue equivalent proportional counters

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Where, N = Total no of ionizing events, ρg and Vg is Density and volume of gas.



is the frequency-mean lineal energy.

Determination of Dose Equivalent using TEPC:



Quality factor Q(L) was determined as per ICRP 103 [1] by assuming the approximation y ≈ L. Results and Discussion: The simulated spectrum for 252Cf neutron source after corrected for distance, time and solid angle was compared with measured one. The results are compared and found to be in agreement as shown in [Table 2]. The overall uncertainty is ±5% measured values of absorbed dose and ambient dose equivalent.

Acknowledgements

The authors are grateful to Dr. D. K. Aswal, Director, HS&EG and Shri Probal Chaudhury, Head RSSD, for their encouragement and support in carrying out the work.
Figure 2: Comparison of simulated & measured spectra of (a) energy deposition inside TEPC (b) dose distribution vs lineal energy spectrum. TEPC: Tissue equivalent proportional counters

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  Reference Top


  1. Chang SY, Kim BH. Understanding of the microdosimetric quantities obtained by a TEPC. J Nucl Sci Technol 2008;213-6.
Table 1: Construction details of tissue equivalent proportional counters

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Table 2: Microdosimetric quantities for 252Cf neutron obtained using tissue equivalent proportional counters

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  Abstract - 33535: Uranyl based metal organic framework as OSL dosimeter Top


Shailesh Joshi1, Madhusmita Panda1, O. Annalaskhmi1,2, C. Venkata Srinivas1,2, B. Venkaraman

1Environmental Assessment Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India

E-mail: [email protected]

Low dose irradiators are used in various industries including food processing for inhibiting sprouting, pest management for reproductive sterilization, and blood irradiation for preventing TA-GVHD. Precise estimation of delivered dose is important for quality control purpose and various reference dosimetry systems viz. Alanine/EPR dosimetry, Fricke dosimetry, and Dichromates are used for dose estimation. Recently Metal organic frameworks (MOFs) are introduced in the field of dosimetry.[1] MOFs are a branch of coordination polymer where metal atoms or a cluster of metal atoms are linked by the organic ligand to form a rigid and robust network. The periodic arrangement of inorganic nodes and bridging organic ligands results in crystalline framework materials with ultra-high porosity, tuneable structure and enormous internal surface area. In this work, an OSL sensitive Uranyl MOF was synthesized and its dosimetry properties were investigated. A mixture of UO2(NO3)2·6H2O, oxalic acid, succinic acid, tetramethyl ammonium bromide, deionized water, and dilute HNO3 were loaded into a 500 mL Teflon-lined stainless steel autoclave. The autoclave was sealed and heated at 150 °C for 36 hrs and then cooled to room temperature after 48 hrs. Yellow crystals of synthesized MOF were washed with deionized water followed by methanol.[2] FTIR analysis confirmed the formation of Uranyl MOF. The IR peaks at 1726 and 1680 cm-1 are due to C=O stretching of ester and amide group of Uranyl MOF. The intense band at 930 cm–1 originated from antisymmetric stretching vibrations of the Uranyl ion, and a very weak band appears at around 830 cm–1, which can be attributed to the symmetric stretching vibration of the Uranyl ion. In order to check the possibility of synthesized MOFs as radiation dosimeter, OSL response of irradiated and unirradiated MOF was recorded at room temperature for a stimulation time of 40 s. It is evident that irradiation has resulted in a significant increase in the OSL signal of the MOF [Figure 1]. The repeatability of the OSL output of the MOF has been tested for 10 cycles of irradiation and OSL readout. The variation in OSL output is within ± 9 %, indicating a good repeatability. The minimum detectable dose was estimated to be 1.1 Gy. Dose response of MOF was checked by irradiating three, 5 mg aliquot for each dose in the dose range of 1 to 200 Gy. The dose response is linear from 1.1 Gy to 100 Gy, after which the signal intensity start decreasing [Figure 2].
Figure 1: OSL response of MOF

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Figure 2: Dose response of MOF

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Keywords: Dosimeter, metal organic framework, OSL


  References Top


  1. Liu H, Qin H, Shen N, Yan S, Wang Y, Yin X, et al. Angew Chem Int Ed 2020;59:15209-14.
  2. Xie J, Wang Y, Liu W, Yin X, Chen L, Zou Y, et al. Angew Chem Int Ed 2017;129:7608-12.



  Abstract - 33566: Ratio of 238U and 226Ra in surface soil in an arid climate and its association with particle size distribution Top


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

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

E-mail: [email protected]

Assumption of unity ratio of 238U to 226Ra in soil matrix in gamma spectrometry is very common for estimation of activity concentration of these radionuclides. However, depending on the climatic condition, soil-water interaction, site specific Kd values, physico-chemical characteristics of soil, etc. influence the activity concentration ratio of the parent-daughter duo. Deviation of the ratio of 238U to 226Ra from unity is very common in high rainfall areas but efforts are made to evaluate in an arid climatic condition in the State of Rajasthan where the annual average rainfall reported less than 800 mm and its correlation with the particle size distribution of the soil in the study area. Ten surface soil samples were collected around Rawatbhata in the Chittaugarh district of Rajasthan as per the IAEA methodology. Soil samples was dried at 110oC for 24 hours to remove the moisture, sealed in standard geometry and kept in sealed condition for one month to ensure secular equilibrium between 226Ra and 222Rn along with its daughters. Soil samples were counted (84000 sec) for identification and quantification of gamma emitters using 50% relative efficiency p-type high-resolution photonic emission spectrometry based on high-purity germanium (HPGe) detector. The MDA for 226Ra was found to be 0.24 Bq (t=300000 sec). An aliquot of the soil samples was analyzed for its particle size distribution using Laser Diffraction Particle Size Analyser based on Fraunhofer Diffraction Theory.[1] Samples were analysed in the range of 0.04 – 2500 μm in liquid dispersion mode using 830 nm and 635 nm wave length of laser lights. Activity ratio of 0.6 to 1.2 was observed for 226Ra/238U in the present study which indicates clear deviation of the assumption of unity in activity estimation. Activity concentration of 238U is measured using 92 KeV while the weighted mean 226Ra is reported from four gamma energies of its daughters (295, 351, 609 and 1764 KeV).[2] Measured activity concentration of 238U and 226Ra was observed in the range of 20.7 – 59.6 Bq/Kg and 16.2 – 53.2 Bq/kg, respectively. The uncertainty in the activity concentration of the measured radionuclides is observed to be in the range of 6- 10% while the counting uncertainty is less than 2%. On the basis of particle size distribution, the surface soils samples were found to be Loamy sand, Silt loam and Sandy loam type as per standard soil classification. It was observed that the activity ratio values of 0.6 – 0.7, 0.7 – 0.8 and > 0.8 correlated with loamy sand, sandy loam and silt loam which is greatly influenced by the sand percentage in soil. Scientific correlation observed in the present study may be corroborated with a greater number of samples so that an empirical relation could be developed for consistent measurement of natural radionuclides in geological matrix.
Figure 1: (a and b) Particle size distribution of two soil samples collected around the study site

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Keywords: 226Ra, 238U, activity ratio, HPGe detector, particle size


  References Top


  1. Miller BA, Schaetzl RJ. Precision of soil particle size analysis Using Laser diffractometry. Soil Sci Soc Am J 2011;76:1719-27.
  2. Knoll GF. Radiation Detection and Measurement. 3rd ed. Wiley; 2009.



  Abstract - 34301: 177mLu as an impurity in a direct-route 177Lu: Implications for patients' dosimetry in peptide receptor radionuclide therapy Top


Kamaldeep1,2, Gaurav Wanage3, M. K. Sureshkumar1, Tapas Das2,4, Sandip Basu2,3, Sharmila Banerjee2,3

1Health Physics Division, BARC, 2Homi Bhabha National Institute, 3Radiation Medicine Centre, BARC, 4Radiopharmaceutical Division, BARC, Mumbai, Maharashtra, India

E-mail: [email protected]

177Lu based radiopharmaceuticals has demonstrated spectacular growth in the field therapeutic nuclear medicine in recent years for the treatment of neuroendocrine tumors (NET). The growth of 177Lu as a therapeutic radionuclide is attributed to its favorable nuclear decay characteristics and its chemical properties. Another important factor which has enhance the availability & popularity of 177Lu radiopharmaceuticals has been the feasibility and ease of production even using medium flux reactors with high yield and adequate specific activity of 177Lu for Peptide Receptor Radionuclide Therapy (PRRT). In India, 177Lu is usually produced by employing 176Lu (n, γ) 177Lu direct method. However, 177Lu-produced by this method invariably contains small amount of 177mLu (half-life of 160.1 days) as a long-lived radionuclidic impurity. This is due to the nuclear reaction 176Lu (n,γ) 177mLu, which is inevitable during production of 177Lu by direct (n,γ) method. The amount of 177mLu coproduced depends on the duration of target irradiation and available neutron flux. The amount of 177mLu radioactivity present in 177Lu produced in medium flux research reactor is experimentally determined to be <0.02%.[1] The aim of the present study was to determine the percentage of 177mLu present in 177Lu-DOTA-TATE at the time of its administration to the NET patients and its impact on radiation dosimetry. For this study, 5 samples of 177Lu-DOTA-TATE were taken, which were formulated using 5 different batches of 177LuCl3. From each sample of 177Lu-DOTA-TATE, 0.1 mL of aliquot, was withdrawn and counted in HPGe detector (P-type coaxial detector GCD-50 190, BSI), after diluting with 5 mL of water. Energy and efficiency calibrations of the detector were done by employing 152Eu standard source, using fixed geometry at 20 cm. From the analyses of HPGe spectrum, the percentage of 177mLu present in 177Lu-DOTA-TATE at the time of dose administration to the patients was calculated. In order to estimate the absorbed dose in kidneys, spleen, liver, bone marrow (BM) and whole- body (WB), 72 hours biokinetic data (4 time points) of 10 patients were used, who underwent treatment with 177Lu-DOTA-TATE for metastatic NET. These biokinetic data were fitted in OLINDA 2.1 software and the absorbed dose per unit activity was estimated for 177Lu-DOTA-TATE and 177mLu-DOTA-TATE. From the data obtained for the percentage of 177mLu impurity present in 177Lu, absorbed dose per unit activity from 177Lu/177mLu -DOTA-TATE and percentage enhancement of absorbed dose by 177Lu due to 177mLu as impurity was calculated. From the 5 samples of 177Lu-DOTA-TATE, which were used, it was found that the average percentage impurity of 177mLu present in 177Lu-DOTA-TATE at the time of dose administration to the patients was 0.011±0.001% (0.814±0.077 MBq of 177mLu per 7400 MBq of 177Lu). The absorbed dose per unit activity due to 177mLu/177Lu-DOTA-TATE, and percentage enhancement of absorbed dose by 177Lu due to 177mLu as impurity are shown in [Table 1]. From this study it can be concluded that the impact of presence of 177mLu as impurity in 177Lu is negligible in patients' radiation absorbed dose in comparison to the benefit received by the patient from this treatment, which is easily available at an affordable cost in India.
Table 1: Absorbed dose estimates for 177Lu-DOTA-TATE and 177mLu-DOTA-TATE (as impurity) and percentage enhancement of the absorbed dose of 177Lu due to the presence of 177mLu as impurity

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Keywords: 177Lu-DOTA-TATE, 177mLu impurity, absorbed dose, neuroendocrine tumor


  Reference Top


  1. Pillai MR, Chakraborty S, Das T, Venkatesh M, Ramamoorthy N. Production logistics of 177Lu for radionuclide therapy. Appl Radiat Isot 2003;59:109-18.



  Abstract - 34443: Dosimetry of indigenously developed round type 106Ru eye plaque  Top


Rajesh Kumar, Ankit Srivastava, Nitin Kakade, S. D. Sharma, Daya Banerjee1, B. K. Sapra

Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 1Process Development Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Plaque brachytherapy is the most widely used treatment modality in the management of uveal melanoma. Currently the majority of eye plaques consist either 125I or 106Ru/106Rh as a radioactive source. However, 106Ru/106Rh based eye plaques are very popular for treatment of tumour with apical height upto 5 mm due to its limited range, steep dose fall off and relative long half-life. As per best of our knowledge, 106Ru/106Rh eye plaque worldwide available from only one manufacture. Considering importance of 106Ru in the management of ocular cancer, BARC has also developed 106Ru/106Rh eye plaque for the treatment of choroidal melanoma and retinoblastoma. This works present the dosimetry of BARC developed circular type 106Ru/106Rh based eye plaques using a specially designed phantom and radiochromic films for its clinical application.

Materials and Methods: 106Ru used in BARC developed eye plaques is recovered fission product. The purified solution of Ru was used for electro-deposition of 106Ru on a silver disk. Final plaque consist of three layers, a 0.9 mm thick silver backing plate of 15.8 mm dia, a 0.2 mm silver substrate/disk on which 106Ru is deposited electrochemically and a 0.1 mm thick silver window. These are arranged such that 106Ru coated silver substrate is sandwiched between silver backing and silver window. The eye plaques resemble to a section of spherical shell with a diameter of 16 mm and internal height of 2.3 mm. The final eye plaque was subjected to regular QA process as recommended in the National/ International standard. Uniformity of activity distribution was estimated using diode dosimeter. Sources with uniformity better than 20% were considered for dosimetry. Dosimetry of eye plaque was carried out using a dedicated eye phantom and EBT3 films. Films were cut with help of the template and filled in the concave portion of the plaque. Total 14 round shape films of varying radii were cut. Out of these 8 films were stacked in concave shape of the plaque and 6 films were stacked above it. Thereafter, 4 films of rectangular shape of size 4 cm by 3 cm were stacked. After it, another 3 films were stacked at an interval of 1mm PMMA separation. Thus the films were placed up to a depth of about 10 mm. After irradiation, the films were read using flatbed scanner. Digitized images of the films were imported into the indigenously developed film dosimetry system, where images were converted into dose map based on preloaded calibration function. Dose to each films were read. These doses were assigned to dose at different water equivalent depth along the central axis of the plaque according to the location of the film the apex of eye plaque. The water equivalent depth was arrived by taking density and scaling factor in to account. All depth dose values along the central axis have been normalized to the depth of 1 mm. Dose rates at clinical reference points of 1 mm and 2 mm were determined using interpolation technique.

Results and Discussions:[Figure 1]a shows the scanned image of the EBT3 films which were irradiated with eye plaque. This scanned image is converted into the dose map image using the previously generated calibration curve [Figure 1]b. Results of the measured depth dose curves using radiochromic film along the central axis have been shown in the [Figure 1]c. Reference absorbed dose rate to water were determined at the depth of 1 mm and 2 mm from the source surface in accordance with ICRU 72 and ISO 21439 and found to be 195.68 mGy/min and 153.75 mGy/min respectively at date of measurement. Film based measurement are simple and Data can be further verified by a medical physicist at Hospital before its clinical implementation.

Conclusions: The clinical dosimetry parameters for the indigenously developed 106Ru eye plaque were generated using radiochromic film. The dosimetry results were used for its regularity approval and further, by hospitals for its clinical application.
Figure 1: (a) Exposed EBT3 films with 106Ru eye plaque. (b) Corresponding dose map (c) measured depth dose curves

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Keywords: 106Ru eye plaque, Gafchromic film Dosimetry, relative depth dose, tissue equivalent phantom


  Reference Top


  1. Heilemanna G, Nesvacil N, Blaickner M, Kostiukhina N, Georg D. Multidimensional dosimetry of 106Ru eye plaques using EBT3 film and its impact on treatment planning. Med Phys 2015;42:5798-808.



  Abstract - 34483: Prospect and radiation safety of helium and oxygen in ion therapy Top


Maitreyee Nandy1,2, Sabyasachi Paul3

1Saha Institute of Nuclear Physics, Kolkata, West Bengal, 2Homi Bhabha National Institute, 3Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India

E-mail: [email protected]

Treatment of some specific types of malignancies is preferably done with energetic positive ions over high energy photons due to the nature of interaction of the former in matter. Energy deposition characteristics of protons and heavy ions help to ensure low toxicity in the normal cells preceding the tumour while depositing a large dose in the target volume. Though heavy ions are more effective than light ions treatment of for hypoxic tumours, but in the case of the entrance channel radiotoxicity is high for very heavy ions. The physics of interaction of the positive ions required that different ions be investigated for their suitability in ion therapy. While the benefits of proton and carbon ion therapy are well established, there are some disadvantages also. These facts led to the study of effectiveness of other ions for therapy. Two such beams studied are 3He and 16O. In this work we have carried out simulation studies of dose distribution in tissue and of the several parameters that influence the radiotherapy using these two types of ions. The Bragg peak is the most important entity to be considered in ion therapy. Linear energy transfer (LET) is proportional to the square of the atomic number of the incident ion for equivalent beam energies. But some nuclear reaction characteristics introduce important differences in the dose deposition profile. These are (i) scattering of the incident beam and (ii) target and projectile fragmentation.[1] Longitudinal straggling of the ion beam at a range R is given by σR∝M-1/2 where M is the mass of the projectile. The lateral scattering σθ is responsible for the dose halo and is proportional to the atomic number Zp of the incident beam and to that of the target (ZT). Another phenomenon which influences the dose distribution in ion therapy is nuclear fragmentation σf. Target fragmentation occurs both in the case when protons as well as heavy ions interact in the tissue. But only heavy ion beams undergo projectile fragmentation. In a given tissue the nuclear fragmentation cross section increases roughly as ~(AP)2/3 and is more pronounced as for heavier ions. These projectile fragments contribute to an effective dose beyond the Bragg peak. We have calculated the longitudinal scattering (σR), lateral scattering (σθ) and projectile fragmentation (σf) for 170 MeV/A 3He and 300 MeV/A 16O beams in water. The calculated values are compared with those obtained for 148 MeV proton and 270 MeV/A 12C beam and are given in [Table 1]. The energies of the beams are chosen so that they have a range of 15 cm in water. Lateral broadening of the 3He beam at the end of range is ~ 3.3 mm. For equivalent energy of the incident beams, longitudinal and lateral scattering of 3He is less than that for hydrogen ions while it is higher than that for 12C. Nuclear fragmentation of 3He is lower than those for 12C. In the case of 16O ions an opposite trend is observed. This study also showed that for the different ions having same range in the tissue dose tailing of helium ions is more than that for protons but is less than the value for carbon ions. Dose distribution in 8 cm x 8 cm x 8 cm volume of water from 170 MeV/A 3He and 300 MeV/A 16O ions has been simulated using FLUKA particle transport code[2] with 106 histories. The results showed that for a target dose of ~2 Gy the ions deliver a dose ~0.4 Gy in the entrance channel. The low LET 3He beam in the entrance channel deliver a low dose to the normal tissues proximal to the Bragg peak. The study showed that oxygen ions, though heavier than carbon, are still in the low LET regime in the incident channel resulting in low toxicity to the normal tissues. The simulation studies carried out in this work showed that radiotherapy with 3He beams can be an optimum choice with respect to low scattering and fragmentation. In dosimetric studies by other workers[3] it was found that Helium ions are effective for therapy of both radiosensitive and radioresistant tumours. This may also be a cost-effective alternative to the heavy ion beams. 16O with a sharp dose fall off is effective for radioresistant tumours.
Table 1: Longitudinal, lateral scattering and fragmentation of 3He, 16O ions compared to 12C and H

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Keywords: 16O ions, 3He ions, fragmentation, ion therapy, longitudinal and lateral scattering


  References Top


  1. Durante M, Debus J, Loeffler JS. Nat Rev Phys 2021;3:777.
  2. Battistoni G, et al. Front Oncol Rad Oncol Sect 2016;00116.
  3. Tessonnier T, et al. Phys Med Biol 2017;62:6784.



  Abstract - 34525: Study of surface dose in breast radiotherapy using unlaminated radiochromic film Top


Nitin Kakade, Rajesh Kumar, Shaju Pilakkal1, Ankit Srivastava, S. D. Sharma, B. K. Sapra

Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 1Department of Radiation Oncology, Reliance Hospital, Mumbai, Maharashtra, India

E-mail: [email protected]

Introduction: Breast cancer is the most common malignancy among Indian females. The advanced radiotherapy techniques are being used in the treatment of breast cancer. Breast radiotherapy is associated with acute and late radiation effects most importantly of the skin, heart and lung. The surface dose is of great relevance in breast radiotherapy because of its impact on acute/late skin toxicity, cosmetic outcome, and local control. The assessment of accurate surface dose is necessary to assure that the skin dose is below the tolerance level to minimize the radiation-induced side effects onto the skin and is sufficient to prevent tumour recurrence. Radiochromic film dosimeter possesses the required dosimetric properties such as, high spatial resolution, nearly energy independent, near tissue equivalence and small thickness. Hence, newly introduced unlaminated Gafchromic EBT3 film (153 μm thick) can be an appropriate dosimeter for surface dose measurement. This work presents the measurement of breast surface dose with unlaminated EBT3 Gafchromic film for 3-Dimensional Conformal Radiotherapy (3D-CRT) and Intensity Modulated Radiation Therapy (IMRT).

Materials and Methods: Unlaminated Gafchromic EBT3 film (International Specialty Product, NJ, USA) having single active layer of 28 μm thickness on single 125-μm transparent polyester substrate was utilized in this study. Before measurement, the calibration curve was generated by exposing EBT3 films for dose up to 4 Gy using 6 MV x-rays. For surface dose measurement, CIRS thorax phantom with breast insert was used. The CT images of the phantom were acquired and 3D-CRT and IMRT treatment plans were generated using the Treatment Planning System. The breast phantom was marked into 4 quadrants and pieces of unlaminated films (3x3 cm2) were placed at each quadrant. The phantom along with the film samples was irradiated using 3D-CRT and IMRT treatment plans. The irradiated films were digitized using EPSON 10000XL scanner and analysed to determine the surface dose using the calibration curve.

Results and Discussion: Measured surface doses with respect to prescribed dose for 3D-CRT and IMRT were 46.97% and 37.92%, respectively. The surface dose is found to be lower in IMRT compared to 3D-CRT. The results of this study are in good agreement with the data reported by Rudat et al.

Conclusion: The breast surface dose was measured for 3D-CRT and IMRT treatments using unlaminated EBT3 Gafchromic film. The dosimetry results suggest that new unlaminated EBT3 Gafchromic film can be a good choice for estimation of surface dose.
Figure 1: (a) CIRS IMRT Thorax Phantom during CT scan, (b) Dose distribution of IMRT treatment plan. IMRT: Intensity modulated radiation therapy, CT: Computed tomography

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Keywords: Breast radiotherapy, intensity modulated radiation therapy phantom, surface dose, unlaminated EBT3 film


  References Top


  1. Rudat V, Aziz Alaradi A, Mohamed A, Kahled A, Saleh A. Radiat Oncol 2014;6:1-7.
  2. Almberg SS, Lindmo T, Frengen J. Radiother Oncol 2011;100:259-64.



  Abstract - 35275: Beneficial applications of radiological imaging technologies during COVID-19 pandemic Top


M. Kumaresan, Ajay Choubey, Surita Kantharia, Shubhra Gupta, Pratishruti Hui

Department of Radiology, Medical Division, BARC Hospital, Mumbai, Maharashtra, India

E-mail: [email protected]

Corona virus disease 2019 (COVID-19) is caused by severe acute respiratory coronavirus 2 (SARS-CoV-2) which causes pneumonia like symptoms in the patient. On 11 March 2020 this disease is labelled as pandemic by WHO. [1] The COVID-19 infection is confirmed by identification of viral RNA by the reserve-transcriptase polymerase chain reaction (RT-PCR) and swab taken from nasal or oropharyngeal area. Medical radiological imaging plays an important role in supporting clinical decision making in the diagnosis, and treatment of various illness. Rapid imaging modalities such as chest X-ray (CXR) and high resolution computed tomography (HRCT) scan and are commonly used in case of general respiratory ailments. Hence these techniques have been deployed from the beginning of COVID-19 pandemic as an additional diagnostic investigation and management tool. This paper discusses the observations on the applications of CXR and HRCT chest during the COVID-19 pandemic. Image acquisition and evaluation CXR imaging of suspected or confirmed Covid-19 cases were performed with portable equipment within specifically designated isolated rooms created near the emergency department for eliminating the risks of cross-infection point of central services. CXRs were performed in the anterior-posterio (AP) and posterior-anterio (PA) views with full inspiration using a portable radiography unit (Konica simple 100 portable DR) in separate dedicated X - ray unit. All images were stored in a computerised radiography database. The advantage of chest radiography (CXR) lies in treatment planning, detecting complications, and assessing progress in radiographically positive cases. Important CXR findings were found to be Bilateral, Peripheral, Lower Zone, Consolidation and/or ground glass opacities. Distribution and discovering subtle changes that are often not visible on chest radiography, hence high resolution computed tomography (HRCT) scan was used. It also supports the diagnosis in suspected cases, determine severity, guide treatment, manage complications and assess treatment response. Multi-detector (MDCT Philips 128) CT scanners was used; the examination is carried out during the end-inspiration phase, when patients can follow breathing instructions. Reconstruction to 1.25 mm slice thickness and multi-planar reconstruction was carried out. The radiological probability of pulmonary manifestations of COVID-19 was reported based on the “CO-RADS classification” [2], a standardised reporting system for patients with suspected COVID-19, ranging from 1 (very unlikely) to 5 (very likely) CO-RADS scores of 1–2 were considered as negative, scores of 4–5 were positive, and a score of 3 was indeterminate. CORADS 6 – Known case of Covid-19 (RTPCR proved). 16620 CXR and HRCT examinations were carried out during 2020-2021 done on patients referred from special COVID-19 OPD started in our hospital. Out of which 8978 cases were conformed covid-19 cases as shown [Figure 1]. These examinations were performed for patients presenting to emergency services at our hospital with clinical chest pain/ loose motion, loss of smell and loss of taste. Common presentation including body ache, cough, fever and breathlessness. We have performed examinations for all suspect or confirmed cases of Covid-19. Of these approximately 50% were Covid-19 proven on RTPCR positive reported present admission. The others were non Covid manifestation of pulmonary or cardiac disease. Medical imaging emerged as a key clinical decision tool in the diagnosis, triaging for appropriate treatment pathways and repeat evaluation of the severely ill patients. Radiation protection principles (justification and optimization) and radiation safety standards have been followed in the radiology department.
Figure 1: Radiological examinations carried during COVID-19 pandemic (2020-2021)

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Keywords: Chest X-ray, COVID-19, HRCT


  References Top


  1. World Health Organization. Use of Chest Imaging in COVID-19: A Rapid Advice. Geneva: World Health Organization; 2020.
  2. Yang W, Sirajuddin A, Zhang X, Liu G, Teng Z, Zhao S, et al. The role of imaging in 2019 novel coronavirus pneumonia (COVID-19). Eur Radiol 2020;30:4874-82.



  Abstract - 35299: A review on sterilization of various types of facemasks by ionizing radiation Top


Amit Kumar, V. Subramanian, B. Venkatraman

Radiological and Environmental Safety Division, Safety, Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, A CI of Homi Bhabha National Institute, Kalpakkam, Tamil Nadu, India

E-mail: [email protected]

Healthcare institutions were forced to reuse respirators after applying different decontamination due to the shortages during the COVID-19 pandemic. Consequently, the usability of the non-woven and textile masks increased significantly. The reuse of various masks also brings unquestionable benefits, both economic and environmental. Although most respirators are disposable and for single use, they can be reused effectively and safely if appropriately sterilized and perform adequately. Ionizing radiation deactivates microorganisms such as bacteria, fungi, viruses, and spores. It is attractive for health care products because it can sterilize large quantities of material through a hermetically sealed package, leaving no residual toxicity and providing safety and logistic benefits. Recently much work has been executed simultaneously for mask sterilization with ionizing radiation Kumar et al. (2020). This paper aims to evaluate the usability of ionizing radiation to sterilize the various masks with different textile materials and compositions used in the present pandemic. Further, providing information on the effectiveness of sterilization by ionizing radiation is viable or necessity for further research. Here, we present a critical review of the available literature on the effect of ionizing radiation on the de-activation of viruses and bacteria Feldmann et al.,[1] the impact on filtration effectiveness for various kinds of masks (certified, non-woven and textile) and possible physio-chemical deterioration of integrity of mask material Kumar et al. (2021).[2] Literature showed different ionizing radiation sources used for mask sterilization, viz. gamma source, high-energy X-rays, linear accelerator, and electron beam irradiation with variable dose ranges from 0.5-50 kGy. Masks effectiveness like filtration efficiency (FE), breathing resistance (BR), penetration (P), quality factor (q) and fit factor were studied with various exposure dose rates (0.51-2.2 kGy/h), different face velocities (0.17-2.6 m/s), range of particles sizes (0.01-10 μm). Further, the changes in morphology, colour, wettability, liquid permeability, mechanical resistance, and physical & chemical alterations of the mask materials were evaluated after sterilization and narrated to modifications in their chemical structure and physical degradation.

The comparison between various studies on the certified mask (N95) is summarized in [Table 1]. The change in FE (ΔFE) of certified masks ranged from 2-77% and was highest for the most penetrating particle size (MPPS: 0.1-0.3 μm) Kumar et al. (2020). An important factor contributing to the FE of certified masks is the presence of charged electret fibres that trap particles through electrostatic or electrophoretic effects. Electrostatic potential (V)/ surface charge density (σ) on the filtration layer is the crucial indicator for predicting FE loss Kumar et al. (2022).[4] The Δσ/ ΔV is altered significantly after sterilization, which is the primary root cause for FE deterioration. As far as BR and fit factors are concerned, no significant change was observed after sterilization. The surgical mask can be sterilized with ionizing radiation without suffering a considerable loss in FE, BR, Q, morphological or structural modifications for a few cycles only. In contrast, the textile mask can be sterilized for at least 20 cycles without considerable suffering in FE or morphological or structural alterations. The textile mask has limited protection (FE: 10-40% for MPPS); however, their uses will protect the inhalation of larger aerosol (> 3.0 μm) which may be generated due to coughing and sneezing. Few recent works advocates recharging the electret masks after sterilization, which recovers FE. Though the ionizing radiation exhibited adverse effects on the FE of certified masks in particle size-dependent manners, its use after sterilization is not medically acceptable. Nonetheless, textile or non-woven mask sterilization tests proved that the materials could be efficiently sterilized by ionizing radiation.
Table 1: Comparison of various works on certified mask

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Keywords: Filtration effectiveness, ionizing radiation, personal protective equipment, sterilization


  References Top


  1. Feldmann F, Shupert WL, Haddock E, Twardoski B, Feldmann H. Gamma irradiation as an effective method for inactivation of emerging viral pathogens. Am J Trop Med Hyg 2019;100:1275-7.
  2. Kumar A, Sangeetha DN, Yuvaraj R, Menaka M, Subramanian V, Venkatraman B. Aerosol Air Qual Res 2021;21:200349.
  3. Kumar A, Bhattacharjee B, Sangeetha DN, Subramanian V, Venkatraman B. J Ind Text 2021;51:3430S-65S.
  4. Kumar A, Joshi S, Subramanian V, Venkatraman B. J Ind Text 2022;51:378S-405S.





    Figures

  [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], [Figure 40], [Figure 41], [Figure 42], [Figure 43], [Figure 44], [Figure 45], [Figure 46], [Figure 47], [Figure 48], [Figure 49]
 
 
    Tables

  [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], [Table 21], [Table 22], [Table 23], [Table 24], [Table 25], [Table 26], [Table 27], [Table 28], [Table 29]



 
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Online since 25th June, 2011